RADIATION PROTECTION
    STANDARDS FOR SCRAP METAL:
 PRELIMINARY COST-BENEFIT ANALYSIS
             Prepared for:

       Radiation Protection Division
        Office of Air and Radiation
   U.S. Environmental Protection Agency
            Prepared under:

Contract Numbers 68-D4-0102 and 68-D2-0155
               June 1997

-------
Page Intentionally Blank

-------
                                                             Preliminary Draft: June 13, 1997
EXECUTIVE SUMMARY
INTRODUCTION

       The U.S.  Environmental Protection Agency (EPA) prepared this cost-benefit analysis to
support the Agency's development of preliminary draft regulations on release standards for scrap
metal from nuclear facilities. Upon their completion, EPA plans to release the preliminary draft
regulations for public comment. This solicitation of comments will not constitute proposed or final
Agency action or a proposed or final EPA rule. Rather, with this solicitation the Agency will begin
a two-year, publicly accessible process that will culminate in the publishing of final regulations. Once
final, these regulations  would  replace  existing release  limits, (e.g.,  the  Nuclear Regulatory
Commission's  Regulatory Guide 1.86) and would likely provide clearance standards for scrap metal
exhibiting either  surface or volumetric contamination; current guidance exists only for surface
contamination.

       EPA anticipates that establishing new standards will alter the management of scrap metal
from Department of Energy (DOE) facilities and facilities licensed by the Nuclear Regulatory
Commission (NRC), with resulting implications for scrap metal management costs and human health
risks. Our preliminary assessment indicates that these impacts vary considerably across three analytic
options: a 0.1  mrem standard, a  1.0 mrem standard, and a 15.0 mrem standard.1 For example, the
analysis suggests  that scrap metal management costs under a 1.0 mrem standard are likely to be
similar to those under current standards; the estimated cost impact of a 1.0 mrem standard ranges
from zero to a savings of $20 million (1997 dollars, present value). In addition, our preliminary
analysis suggests  that a 1.0 mrem standard would be somewhat more protective of human health
than current standards, reducing cancer incidence (i.e., the number of total cancer cases predicted
to occur over 1,000 years) by six to 10 cases relative to baseline conditions. In contrast, we estimate
that a  0.1 mrem  standard would increase costs relative to the baseline by $200 million to $500
million and reduce cancer incidence by eight to 14 cases, while a 15.0 mrem standard would save
$1.4 billion to $1.7 billion but increase  cancer incidence by 19 to 29 cases.

       We  note  that  these results are preliminary  and based  on  a  number of  simplifying
assumptions. As a result, they should be interpreted with caution. We  believe, however, that the
results provide a good indication of the relationship of each of the three analytic options to current
standards.  In  upcoming months, EPA intends  to conduct further  research  to strengthen this
preliminary analysis. s    ,  -  i    „  ,     •       ,  ,	/. ,   ,,„..»-  \; »••  '
    1 These options have been developed to illustrate potential impacts across dramatically different
release standards. They do not reflect specific regulatory options under consideration by EPA.

                                           ES-1

-------
                                                             Preliminary Draft: June 13, 1997


       This summaiy contains three sections. The first outlines the analytic approach employed to
generate the preliminary results. The second section presents the results of the analysis for each of
the three options analyzed, discussing results for scrap metal from both DOE facilities and NRC-
licensed commercial nuclear power reactors. The third section discusses the potential implications
of these results, as well as the limitations of the analysis.

                                                                                  s
OVERVIEW OF ANALYTIC APPROACH

       The  cost-benefit  analysis  described in this report compares  estimated scrap metal
management costs and human cancer risks under three analytic scenarios  to the costs and risks
associated with the standards that currently govern the release of scrap metal from DOE and NRC-
licensed nuclear facilities. Exhibit ES-1 illustrates the approach employed to assess these effects. As
shown in this exhibit, the analysis requires first predicting likely current and  future practices under
existing standards, then comparing these "baseline"  practices to likely practices under alternate
clearance standards.  The disposition practices analyzed include: (1) disposing of the scrap metal in
burial facilities; (2)  fabricating the metal into products  for reuse within the nuclear complex
(generally referred to as "restricted recycling"); and (3) releasing scrap metal for unconditional use.
EPA's preliminary draft regulations would affect practices  relating to option three, releasing scrap
metal for unconditional use.2

       In this phase  of our assessment of the  potential effects of EPA's preliminary draft
regulations, we identified the major sources of scrap metal potentially affected by the rulemaking.
These sources include 11 large DOE facilities and 123 NRC-licensed commercial  nuclear power
reactors.3 The analysis  considers 936 thousand metric tons  of scrap metal likely to be generated by
the DOE facilities and  641 thousand tons likely to be generated by the power reactors. We collected
information on the physical and radiological characteristics that affect decisions to decontaminate
and release scrap metal for unconditional use (i.e., use outside DOE or NRC regulatory control),
such as the source of the scrap metal item, the type of metal it contains, its physical form, initial
radioactivity  levels,  and whether contamination  is  limited to  the metal's  surface  or  extends
significantly beyond  the surface (i.e., whether the  item is  "volumetrically" contaminated). Due to
considerable  uncertainty concerning the radiological  profiles of the metal in this inventory, we
characterized potential  ranges of  activity levels for each scrap metal  item, based  on  our
understanding of the operations of a particular facility.4
    2 Throughout this report, we use the term unconditional to refer to the determination that
residual levels of radioactivity hi scrap metal from Federal or NRC-licensed nuclear facilities are low
enough that the metal need not be managed as radioactive material, but instead can be released
from the institutional control of the nuclear facility with no limitations.

    3 We may address additional Federal and nonfederal sources of scrap metal in future analyses.
                                                                  i
    4 Separate radiological profiles were developed for each of the 11 DOE facilities included in the
analysis.  The characterization of scrap  metal from NRC-licensed commercial nuclear power plants
was based upon radiological profiles for a reference boiling water reactor (BWR) and pressurized
water reactor (PWR).

                                           ES-2

-------
                                                    Exhibit ES-1
                           ANALYTIC APPROACH FOR COST-BENEFIT ANALYSIS
Identify and character-
ize potentially affected
scrap metal volumes.
• 11 major DOE
 facilities
•123NRC-licensed
 commercial nuclear
, power plants
                              Predict baseline disposition
                              practices and related costs.
                              • Assume maximum release at
                               Reg. Guide 1.86 and DOB
                               order 5400,5 standards
                              Predict post-regulatory
                              practices and related costs for
                              each analytic option.
                              * Assume maximum release at
                               0.1 mrem, 1.0 mrem, and
                               15.0 mrem levels
 Predict baseline cancer
      incidence.
 Predict post-regulatory
cancer incidence for each
    analytic option.
                                   Estimate impacts of
                                   each analytic option.
                                   » Change in scrap
                                     management costs
                                   • Change in cancer
                                     incidence ^
                                   « Other impacts

-------
                                                            Preliminary Draft: June 13, 1997


       The analysis considers likely scrap metal management practices under three analytic options
 and compares them to current and future practices under current, or "baseline"  standards.  The
 baseline standards used for this phase of  the analysis include the current surface contamination
 release guidelines established under NRC's Regulatory Guide 1.86 and DOE Order 5400.5.  The
 three analytic options would establish release standards designed to limit the annual radiation dose
 to the reasonably maximally exposed (RME) individual to less than 0.1 mrem, 1.0 mrem, or  15.0
 mrem. We assume that these standards would apply to scrap metal exhibiting either surface or
 volumetric contamination.5

       We developed detailed cost estimates for two disposition options: (1) permanent disposal as
 low level radioactive waste; and (2) unconditional clearance, with or without prior decontamination.6
 Due to uncertainties concerning disposal  costs, we considered both high-end and  low-end  cost
 estimates for scrap metal 9iiginating from both DOE facilities  and NRC-licensed commercial
 nuclear power reactors.

       To predict how scrap metal  would be managed under alternate release standards, we
 developed an economic model. For each analytic option, the model employs data on scrap metal
 characteristics and the costs of alternate management practices to estimate the costs associated with
 disposal  or unconditional clearance.  The model then compares  these costs to determine the
 approach that would be selected under each set of standards, assuming that decision-makers will
 always choose the lowest cost option.

       The methodology  employed to characterize the  effect of alternate standards  on human
 cancer risks  is  based on  a  method used  in recent analyses by EPA.7 The basic approach for
 evaluating individual risk consists of estimating the dose to individuals exposed to scrap metal at
various stages of the scrap recovery process, including transport, processing, and disposal workers,
 as well as consumers of products containing scrap metal; For example, we estimate the dose that
would be received by a truck driver who transports scrap metal to a processing facility, a worker who
cuts the metal prior to processing, workers exposed to the metal at various stages hi the  production
process,  workers who use tools or  machinery  manufactured from the  recovered metal,  and
consumers who use products (e.g., a kitchen range or frying pan) made from  recovered metal. We
 also estimate the dose that would be received by an individual who consumes  groundwater
 contaminated by slag leachate from a metal recovery facility,  and the dose  that would be received
 by a subsistence farmer whose crops are contaminated by air emissions from a metal recovery plant.
These doses can then be scaled to develop release standards that are protective of the most exposed
    5  Current standards do not  provide  generic activity guidelines for releasing scrap metal
containing volumetric contamination.

    6 We qualitatively discuss  the impact  of restricted recycling options on our findings; future
analyses may incorporate a quantitative assessment of these impacts.

    7 The Technical Support Document (TSD), which accompanies this report, explains the method
in detail.

                                           ES-4

-------
                                                            Preliminary Draft:  June 13, 1997


individual (eig., release standards that reduce doses to less than 1.0 mrem per year for the most
exposed individual). Using models that incorporate the release standards developed in the individual
risk analysis, the collective impacts analysis estimates the number of cancer cases that would occur
under baseline requirements and under each of the analytic options.

       The analysis described above -yields three categories of impacts potentially attributable to
EPA's rulemaking:

       '•      Cost impacts.  EPA's new standards may increase or decrease management
              costs for scrap  metal that exhibits surface contamination, .depending upon
              whether the standards impose release limits that are lower or higher than
              current limits. These impacts are the subject of Chapter 4. If EPA's release
              limits are  lower, allowing less residual surface contamination than current
              standards, then scrap metal management costs are likely to increase, since
              more metal is likely to require decontamination prior to release. In some
              cases, the value of  the metal to be  recovered is unlikely to justify the
              additional decontamination costs, resulting in  a  decrease in the quantity of
              metal released  and an equivalent increase in the quantity disposed. The
              opposite is likely to hold true if EPA's standards allow more residual surface
              contamination than current guidelines; more metal will qualify for release
              without  decontamination,  and more metal will be recovered rather than
              disposed.

              EPA's  new standards  are unlikely to  increase  management  costs  for
              volumetrically-contaminated scrap metal, since the current lack of generic
              guidelines for the release of such metal would likely result in most or all of
   /           it being disposed. To the extent  that decontamination and release of such
              metal under EPA's standards would be more cost-effective than disposal,
              EPA's  rulemaking  would reduce  management costs for volumetrically-
              contaminated metal.

       •      Predicted  changes in cancer risks.  EPA's  new standards may increase or
              decrease the number of cancer cases predicted  to occur in the population
              over the next  1,000 years. The  direction of  the change depends  on the
              amount of scrap metal released and on whether the release limits that EPA
              establishes are higher or lower than existing standards. These changes are the
              subject of Chapter 5.

       •      Other impacts.  EPA's rulemaking may have  a  variety of other impacts  in
              addition to those cited above. These impacts include changes in the markets
              for scrap metal and waste disposal capacity, as well as effects on non-cancer
              human health risks, ecological impacts, and demand for virgin materials. In
              this phase of the analysis, we address these issues qualitatively. These impacts
              are the subject  of Chapter 6.

We discuss these impacts in detail below.
                                           ES-5

-------
                                                            Preliminary Draft: June 13, 1997
COMPARISON OF COSTS AND BENEFITS
       Exhibit ES-2 summarizes the results of the preliminary analysis for each of the three analytic
options considered. We include the results for both the low and high disposal cost scenarios,
assuming initial activity levels at the midpoint of the ranges estimated for the scrap metal of interest.
Future disposal  costs have been discounted to 1997 dollars using  a  real discount rate of seven
percent. The results of the human health risk analysis indicate the change in number of cancer cases
predicted to occur over 1,000 years.
0.1 Mrem Option

       As illustrated in the exhibit, we estimate that the 0.1 mrem option would increase scrap
metal management costs by $0.2 to $0.5 billion and decrease cancer incidence by eight to 14 cases.
This-result is expected, since a 0.1 mrem standard would set lower release limits than current
standards, making it more costly to decontaminate scrap metal to meet the limits and lowering the
residual activity in released metal. A change hi the disposition of scrap metal from NRC-licensed
commercial nuclear  power reactors accounts for nearly all of the cost increase, as only 11 to 31
percent of available scrap metal from these facilities would be released under  the 0.1 mrem
standard, compared  to 62 percent to 74 percent under current standards. In contrast, the analysis
suggests that changes in the management of scrap metal ffrom the 11 major DOE facilities would
have little impact on costs; we estimate that only six to nine percent of the DOE facilities' scrap
metal would  be released under current standards, and that none  of this metal would be released
under the 0.1 mrem  standard.

       With  more scrap metal flowing to burial under the 0.1 mrem option and lower activity levels
in the scrap metal that is released, the 0.1 mrem standard yields an estimated reduction in cancer
risks of eight to 14 cases. Again, a change hi the management of scrap metal from NRC-licensed
commercial nuclear power reactors accounts  for most of the predicted impact, as metal that would
be released under the current standards would  instead flow to burial.8 The estimated change in
cancer incidence associated with management of scrap metal from DOE facilities is minimal, since
most of the affected scrap metal would be buried under both current  standards and the 0.1 mrem
option.
   8 In this preliminary analysis, we assume that burial has zero cancer risk.

                                           ES-6

-------
Preliminary Draft: June 13,1997
Exhibit ES-2
POTENTIAL IMPACTS OF EPA CLEARANCE STANDARDS
Impact1
Change in Costs2
Change in
Cancer Incidence3
Other Impacts4
0.1 mrem Option
Tetal
$0.20;- $0.47
(8.2) -<14.3)
• .
DOE
Facilities5
$0.03 - $0.06
negligible
uncertain
NRC
Facilities*
$0.17 - $0.41
(8.2) -(14.3)


Total
$0.0 - ($0.02)
(6.3) -(10.0)

1.0 mrem Option
DOE
Facilities5
($0.04) - ($0.04)
negligible
uncertain

NRC
Facilities'
$0.04 - $0.03
(6.3) -(10.0)

15.0 mrem Option
DOE NRC
Total Facilities5 Facilities'
($1.40) - (1.65) ($1.36) - ($1.58) ($0.04) - ($0.08)
19.2 - 28.8 1.2 - 1.2 17.9 - 27.6
uncertain
Notes: • .
1 Low and high values represent results under low and high disposal cost scenarios, respectively, relative to current release guidelines (Regulatory Guide 1.86). The values shown
are based upon initial levels of radioactivity at the logarithmic midpoint of the range reported for each scrap metal item.
2 Expressed in billions tif J997 dollars; costs discounted to their present value using a real discount rate of seven percent.
3 Total cases (fatal and flqff-fatal) predicted to occur over 1,000 years.
4 Includes other economic impacts potentially attributable to the rulemaking such as effects on scrap metal markets and non-cancer human health and environmental effects.
These impacts are likely to be small, and insignificant relative to impacts on scrap metal management costs and cancer incidence.
5 Includes scrap metal generated by 11 large DOE facilities.
6 Includes scrap metal generated by 123 NRC-licensed commercial nuclear power reactors.

-------
                                                            Preliminary Draft: June 13, 1997
 1.0 Mrem Option
        Our analysis of the 1.0 mrem option shows minimal cost impacts relative to the current
standards (savings of zero to $20 million) and a decrease in predicted cancer incidence of six to 10
cases over the 1,000-year modeling period. The estimated cost savings, which are attributable entirely
to the management of scrap metal from DOE facilities, largely stem from the availability of a
volumetric release standard. Under current guidelines, no generic standards exist for the release of
volumetrically contaminated metal; this not only limits the recovery of volumetrically contaminated
items, but generally rules out melting as a scrap metal decontamination and recovery option. As a
result, disposal is  the only practical management method  for volumetrically contaminated scrap
metal or for items for which melting is the only feasible or cost-effective recovery technology. The
establishment of  a volumetric  clearance  standard under EPA's  rulemaking  would  create
opportunities to recover volumetrically contaminated metal, and would also make melting a more
viable metal recovery practice, thus allowing material that otherwise would be buried to be released
instead.

        In contrast to the change in scrap management costs, the change hi predicted cancer
incidence under the 1.0 mrem option is primarily attributable to changes in the management of scrap
metal from NRC-licensed commercial nuclear power reactors. We estimate that the same volume
of scrap metal from these reactors (61 to 85 percent of the total) would be released under both
current standards and the 1.0 mrem option. The 1.0 mrem standard, however, would establish lower
release limits for the indicator radionuclides at NRC facilities, thereby increasing the extent to which
it would be necessary to decontaminate such metal prior to release. The resulting assumed reduction
in residual activity levels accounts for the predicted reduction in cancer incidence.9
15.0 Mrem Option

        A 15.0 mrem standard would allow higher release limits than the current standards, yielding
estimated cost savings of $1.4 billion to $1.7 billion but prompting an increase in predicted cancer
incidence of 19 to 29 cases. Again, a change in the management of scrap metal from DOE facilities
accounts for most of the estimated cost savings, as 98 percent of the DOE facilities' total scrap metal
inventory would be released  for unconditional use (compared to only six to nine percent  under
current standards). Most of this scrap metal (94 percent) would not require prior decontamination.
We estimate that NRC-licensed commercial nuclear power reactors would release up to 84 percent
of their total scrap metal inventory under this option, which is slightly more than under current
standards, and would also realize some savings due to lower decontamination costs. Unlike the DOE
facilities, approximately 72 percent of the scrap metal released from the commercial nuclear power
reactors would  require decontamination prior to release.
    9 In this preliminaxy analysis, the radionuclide primarily responsible'for the health impacts
attributed to scrap metal from NRC-licensed commercial nuclear power plants is Cobalt-60 (Co-60).
The surface  activity limit  (dpm/lOOcm2)  corresponding to a  1.0  mrem  dose for  Co-60 is
approximately one-fifth of the limit currently prescribed in Regulatory Guide 1.86.

                                           ES-8

-------
                                                             Preliminary Draft: June 13, 1997


        Despite the large increase in the quantity of DOE scrap metal likely to be released, this
scrap metal accounts for only a small percentage of the projected increase in cancer incidence under
the 15.0 mrem standard. This result is largely due to the higher activity levels that are assumed to
be associated with scrap metal from the commercial nuclear power reactors. Our analysis assumes
that final activity levels in scrap metal items decontaminated prior to release are at the maximum
levels allowed under each option, while final activity levels in scrap metal released directly from
facilities with  no prior decontamination  are equal to starting activity levels. As noted  above, we
estimate that 94 percent of the DOE facilities' scrap metal could be released under the 15.0 mrem
option without prior decontamination, compared to only 28 percent  of scrap metal from NRC-
licensed commercial nuclear power reactors. As a result, the analysis treats final activity levels for
scrap metal released from DOE facilities as lower, on average, than final activity levels for the scrap
metal generated by the commercial power reactors. In addition, the radionuclides and exposure
pathways that  drive the cancer risk analysis also  differ for the two source categories. In the analysis
of scrap metal from DOE facilities, U-238 is responsible for virtually all of the change in predicted
cancer incidence; the key exposure pathway is exposure to workers handling slag. In contrast, Co-60
accounts for the majority of predicted cancer cases related to exposure to scrap metal from NRC-
licensed commercial nuclear power reactors; in this case, the dominant exposure pathway is through
consumer products. These differences indicate that our assumptions concerning the types of nuclides
present in the scrap metal and related activity levels can significantly affect the results of our risk
assessment.
Other Impacts

        EPA's rulemaking may have additional impacts on a variety of factors, including: (1) scrap
metal market prices and the demand for low level radioactive waste disposal capacity; (2)  non-
carcinogenic human health risks; (3) ecological impacts; and (4) demand for virgin materials  (e.g.,
iron ore). Based on our preliminary review of these issues, it is likely that these impacts are small
and insignificant compared to direct cost effects and impacts on cancer risks. As a result, we have
not attempted to quantify these impacts or to differentiate their magnitude across the three analytic
options.10 We may revisit these issues in subsequent phases of this analysis.
    10 This preliminary analysis does not  address several other potentially significant impacts,
including environmental justice issues, effects on small businesses, and the relationship of EPA's
standards to other governmental programs.   Assessments of these potential  impacts  may be
conducted over the next several months.

                                           ES-9

-------
                                                             Prelimmary Draft:  June 13, 1997
CONCLUSIONS
       Although these results are preliminary and subject to considerable uncertainty, we can draw
the following initial conclusions from the analysis.

       •     The analysis provides useful information on the relative impacts of the three
              analytic options.  The results provide rough measures of the relationship of
              each option to the current standards; however, they should not be interpreted
              as providing precise absolute estimates of either disposition costs or potential
              cancer incidence.

       •     Cost impacts and predicted cancer risks vary considerably across the three
              analytic options.  The predicted change in scrap metal  management costs
              ranges from an increase of up to $0.5 billion under the 0.1 mrem standard to
              savings of up to $1.7 billion under the 15.0 mrem standard.  In addition, the
              0.1 mrem standard yields an estimated decrease in cancer incidence of up to
              14 cases over 1,000 years, while the 15.0 mrem standard yields an estimated
              increase of up to 28 cases. These results, however, are highfy dependent on
              a range of assumptions,  including those concerning baseline practices and
              numerous others embedded in our risk modeling.

       *     The relationship between predicted changes in costs and predicted changes
              in cancer  risks varies  across  the three analytic options.  Under the 15.0
              mrem standard, costs are predicted to  decrease relative to  costs under the
              current standards, while cancer risks are predicted to increase. The 0.1 mrem
              Standard yields the opposite result, with a predicted increase in costs but a
              predicted decrease in cancer risks. In contrast, the analysis suggests that scrap
              metal management costs would remain relatively unchanged under the 1.0
              mrem option, while cancer risks are predicted to decline.

       *     The results  of  the  analysis  are highly dependent upon the  assumed
              radiological profiles of affected scrap metal. For example, the differences in
              predicted cost and cancer impacts for scrap metal from  DOE facilities and
              scrap metal from NEC-licensed commercial nuclear power reactors are largely
              attributable to differences hi the assumed mix of dominant radionuclides and
              related activity  levels  in the scrap  metal generated  by the  respective
              complexes. The results of the analysis are extremely sensitive to variation in
              radiological profiles.
                                                -I?
                                          ES-10

-------
                                                             Preliminary Draft: June 13, 1997
KEY UNCERTAINTIES AND NEXT STEPS
        Each component of the analysis, including identifying and characterizing affected volumes
of scrap metal, estimating baseline and post-regulatory scrap metal management practices and costs,
and  estimating  changes in human cancer  risks, has  associated limitations. The key areas of
uncertainty include:

        •     Scrap metal characteristics data. Available information on the year in which
              scrap metal is likely to become available for recycling, the metal's radiological
              characteristics, and its physical  form is limited and highly uncertain. These
              uncertainties may lead us to either under- or overstate the effects of alternate
              release standards.11

        •     Future scrap metal disposition practices and related costs. The analysis does
              not consider restricted recycling, which may lead us to overstate total scrap
              metal management costs and  the quantities of metal likely to be disposed or
              released for unconditional use. Moreover, decontamination costs are likely to
              change as the industry evolves,  and disposal options are difficult to predict,
              creating  additional analytic uncertainty. Finally, the analysis assumes  that
              generators will select the least-cost disposition option, ignoring the effects of
              non-economic factors (e.g., public opinion) that may  discourage release of
              scrap metal. To the extent that non-economic factors influence decision--
              making, we likely understate scrap metal management costs and overstate the
              quantity of metal likely to be released for  unconditional use.

        •     Predicted cancer risks.  The  risk model employs a number of conservative
              assumptions that may lead  it to overestimate doses under various exposure
            - scenarios. In addition,  our cost model estimates the maximum quantity of
              scrap metal that could be released for unconditional use under the proposed
              standards and assumes that decontamination efforts reduce activity levels only
              to the maximum permitted under each release standard, leading to potential
              overstatement of collective cancer impacts. We are uncertain, however, how
              these limitations may affect our assessment of the relative effects of each of
              the analytic  options.
                             ,    .
  .- " In addition, the analysis probably  understates the total amount of scrap metal potentially
affected by EPA's rule, since it focuses only on metal from 11  major DOE facilities and NRC-
licensed commercial nuclear power plants; other Federal and non-federal  facilities will also be
affected by the rulemaking.

                                           ES-11

-------
Page Intentionally Blank

-------
                              TABLE OF CONTENTS
OVERVIEW OF ANALYTIC APPROACH 	 CHAPTER 1

       Introduction	1-1
       Regulatory Framework	1-3
       Analytic Options	-	1-12
       Analytic Approach  	1-15


SCRAP CHARACTERISTICS  	 CHAPTER 2

       Introduction and Summary	2-1
       Analytic Approach  	2-4
       Scrap Metal from Major DOE Facilities  	.*..,..	2-6
       Scrap Metal from Commercial Nuclear Power Reactors  	2-11
    •   Key Uncertainties and Plans for Future Analysis	2-15
                 i
DISPOSAL AND RECYCLING COSTS	 CHAPTER 3

      Introduction and Summary	3-1
      Analytic Approach 	3-3
      Disposal Costs	3-4
      Unconditional Clearance Costs ~.	: 3-14
      Restricted Recycling	3-21
      Key Uncertainties and Plans for Future Analysis	3-23

                                                                •»
CHANGES IN COSTS AND QUANTITIES RECYCLED 	CHAPTER 4

      Introduction and Summary  .	4-1
      Analytic Approach	4-5
      Scrap from Major DOE Facilities	\	4-8
      Scrap from Commercial Nuclear Power Reactors 	4-11
      Implications and Plans for Future Analysis	4-13


CHANGES IN HEALTH EFFECTS ATTRIBUTABLE TO THE REGULATIONS .. CHAPTER 5

      Introduction and Sumrriaiy  '	*}. . .••; ...•*	-	•.': ..:..* .5-1
      Analytic Approach	t :„...•„	.,... . . . t,	- .•". . tip., -\	.5-6
      Findings	5-20
      Uncertainties and Next Steps  . .T:	<.	';.-..-	5-26

-------
                           TABLE OF CONTENTS
                               , (continued)
OTHER IMPACTS	  CHAPTER 6

      Introduction and Summary	6-1
      Other Economic Impacts	6-2
      Other Impacts on Human Health and the Environment	6-8
      Environmental Impacts of Reducing Demand for Virgin Materials  	6-17


SUMMARY AND CONCLUSIONS	  CHAPTER 7

      Introduction and Summary	7-1
      Comparison of Costs and Benefits  	7-1
      Implications and Next Steps	7-5


REFERENCES
APPENDICES

     Appendix A:  SURFACE AND VOLUMETRIC  RELEASE LIMITS
                 UNDER CURRENT STANDARDS AND THE THREE
                 ANALYTIC OPTIONS

     Appendix B:  DEFINITIONS OF PHYSICAL FORM CATEGORIES

     Appendix C:  DETAILED DATA ON DOE SCRAP CHARACTERISTICS

     Appendix D:  DETAILED DATA ON NEC SCRAP CHARACTERISTICS

     Appendix E:  REJECT RATES FOR DECONTAMINATED SCRAP METAL
                 BY PHYSICAL FORM

-------
                                                             Preliminary Draft: June 12, 1997
OVERVIEW OF ANALYTIC APPROACH                                        CHAPTER 1
INTRODUCTION

       The U.S. Environmental Protection Agency (EPA)  is currently  developing radiation
protection standards for the release of scrap metal from nuclear facilities.  Management of the
potentially large scrap metal inventoiy at nuclear facilities has attracted heightened attention  as
more facilities enter the decontamination and decommissioning phase. The purpose of this report
is to analyze the extent to which EPA's rulemaking may change the scrap metal disposition decisions
made by nuclear facilities, and to estimate the economic, health,  and environmental impacts of these
changes.

       As Exhibit 1-1 illustrates, nuclear facilities generally can  manage scrap metal in one of three
ways: (1) they can dispose the scrap metal hi a facility designed to accept nuclear materials; (2) they
can fabricate the metal into products for re-use hi a nuclear setting (e.g., containers for waste
disposal), a practice generally referred to as "restricted recycling;" and (3) they can release materials
for unconditional use (e.g., through sale to a private scrap metal dealer).1 EPA's mlemaking will
affect release standards for the third option - the unconditional clearance of scrap metal.  Changes
in the release standards for unconditional clearance may in turn affect the quantity of metal that is
disposed or released, with potentially significant implications for scrap metal management costs and
associated health or environmental risks.

       This introductory chapter provides information on the regulations that currently govern the
release of scrap metal from nuclear facilities, describes the focus of EPA's rulemaking effort, and
outlines the analytic  approach employed in this preliminary  cost-benefit analysis.   Subsequent
chapters discuss the methodology and preliminary results for each component of the analysis. These
chapters also provide information on the limitations of the analysis and plans for future research.
  1  Scrap metal may also be placed in storage prior to final disposition.

                                            1-1

-------
                                               Exhibit 1-1

                            OVERVIEW OF SCRAP DISPOSITION OPTIONS
                       -Nuclear Regulatory Controls
                         Option 1: Disposal
Scrap Metal from
Nuclear Facility
                         Option 2: Restricted Recycle
                         Option 3: Unconditional Clearance
Manufacture
Products for
Nuclear Use
                                                           Decontaminate,
                                                            if necessary
                                                                                            Sell Clean Metal to
                                                                                              Scrap Dealer

-------
                                                             Preliminary Draft:  June 11, 1997
 REGULATORY FRAMEWORK
       Nuclear facilities operated by the Federal government and private industry can currently
 release scrap metal for unconditional use under guidance developed by the U.S. Department of
 Energy (DOE) and the U.S. Nuclear Regulatory Commission (NRC). This section discusses the
 existing guidance as well as the potential changes that may occur under EPA's regulations.  We
 focus on requirements that affect  DOE facilities and commercial nuclear power plants because, as
 discussed,in more detail in Chapter 2, most of the scrap metal potentially affected by EPA's rule
 is likely to be generated by such facilities. Smaller quantities of scrap metal potentially affected by
 the rule may be generated by the activities of other Federal agencies, such as the U.S. Department
 of Defense, and by other NRC licensees, such as hospitals.

       For both DOE sites and NRC licensees, releases are governed by guidance documents that
 describe methods for determining whether unconditional clearance is appropriate and that require
 that radiation levels be "as low as reasonably achievable" (ALARA).  The process for deciding
 whether  materials  can be  released  includes  determining whether  the  material  is potentially
 contaminated, identifying relevant release limits, surveying the material to determine whether it is
 below the limits, then selecting the appropriate disposition options. An overview of these steps is
 provided in Exhibit 1-2; related requirements are described in more detail below.

       It is important to note  that the 'guidance discussed below specifies maximum activity levels
 for the release  of scrap metal.   Other factors, such as concerns about adverse public reaction, can
 significantly affect decisions to release scrap metal.  Our research suggests that these other factors,
 in combination with ALARA requirements, may limit the release of scrap metal to a quantity less
 than that which would otherwise be eligible for release, although the extent of this effect is difficult
 to quantify.
Current DOE Requirements

       To release scrap metal for unconditional use (e.g., by selling it to a commercial scrap metal
dealer), DOE site managers must comply with a variety of regulations and guidelines. The major
source of relevant guidance is DOE Order 5400.5,  "Radiation Protection of the Public and the
Environment."2 This order describes the procedural and analytical requirements for releasing scrap
metal as well as other materials from DOE control, and provides guidance on the surface activity
levels allowable at the point of release.
  2 U.S. Department of Energy.  DOE Order 5400.5:  Radiation Protection of the Public and the
Environment.    Washington,  DC,  1990;  and  "Response  to  Questions  and Clarifications  of
Requirements  and Processes:  ' DOE  5400.5, Section n.5  and  Chapter IV  Implementation
(Requirements  Relating  to Residual  Radioactive  Material)," DOE Assistant Secretary for
Environment, Safety and Health, Office of Environment (EH-41), November 17,1995.

                                            1-3

-------
                                        Exhibit 1-2
                           OVERVIEW OF RELEASE PROCESS
    Requirements
                                                               Release Process
  DOB Order 5400.5,
NRC Regulatory Guide
   l.8f>;ind Related
    Requirements
  ALARA Analysis
 NRC Agreement State
     Requirements
 Is matenal
 potentially
ontaminated
                                                                     Yes
 Determine
 Applicable
  Limits
No
                                                         or dispose
                                                                       Unconditional
                                                                        . clearance

-------
                                                            Preliminary Draft: June 11, 1997


       DOE Order 5400.5 establishes limits on the amount of radiation exposure an individual and
the public can receive.  The Order sets a primary dose limit for the public from all exposures of 100
mrem per year and requires that any single release of material from DOE must account for only a
fraction of this total (e.g., one-third or less of the primary public dose limit).  DOE Order 5400.5
prohibits the release of contaminated  material unless sufficient analyses have been completed  to
ensure that the release will not result in harmful exposure.  The analyses must document that the
level "of radioactivity is "as low as reasonably achievable" (ALARA). The process for  determining
whether materials  meet the ALARA goal is formally documented in DOE guidance and must be
followed to minimize worker and general population exposure to radiation from all DOE activities,
not only the  unconditional clearance of scrap metal. A full scale ALARA  assessment for scrap
metal involves specifying alternate disposition  options and completing a radiological risk and
economic  assessment of these alternatives.  The level of detail required to complete  an ALARA
assessment varies according to the potential level of risk.  For example, releases involving a large
volume  of metals  or occurring over extended periods of time would require  a more extensive
analysis.

       To release material from a DOE facility, site managers must complete the following general
steps, as previously illustrated hi Exhibit 1-2.

       •      Determine whether material is contaminated. Material must be  treated as
              contaminated if it has been  stored or used in areas where radioactivity is
              present.  In general, material stored or used outside of radiation control
              areas is not subject to DOE Order 5400.5  requirements.

       •      Develop authorized limits. Authorized limits are the maximum amount of
              radioactivity that may be present for material to be released.  The ALARA
              process  is used to establish these limits and to ensure that release will not
              exceed the basic dose limits under "worst case" or "plausible use" scenarios.

       •      Employ generic surface contamination guidelines  where applicable.   To
              facilitate the release process, DOE Order 5400.5 includes  generic surface
              contamination guidelines that site managers can apply rather than developing
              a more detailed, case-specific ALARA analysis. The ALARA principles still
              apply; however, only a semi-quantitative or qualitative assessment is generally
              required.  DOE has not established similar guidelines for  material that is
              volumetricalfy-contaminated, which generally cannot be released without a
              detailed, quantitative ALARA analysis. We present the DOE Order 5400.5
              surface contamination release limits in Exhibit 1-3. Note that these limits are
              based primarily on survey technology capabilities.
                      ' ^
                                            1-5

-------
                                                                           Preliminary Draft:  June 11, 1997
                                                   Exhibit 1-3

                   DOE ORDER 5400.5 SURFACE CONTAMINATION GUIDELINES3
Radionuclides5
Group 1 - Transuranics, 1-125, 1-129, Ac-227, Ra-226,
Ra-228, Th-228, Th-230, Pa-231
Group 2 - Th-natural, Sr-90, 1-126, 1-131, 1-133, Ra-223,
Ra-224, U-232, Th-232
Group 3 - U-natural, U-235, U-238, and associated decay
products, alpha emitters
Group 4 - Beta-gamma emitters (radionuclides with
decay modes other than alpha emission or spontaneous10
fission) except Sr-90 and others noted above
Tritium (applicable to surface and subsurface)11
Allowable Total Residual Surface Activity
(dpm/100 cm2)4
Average*7
100
1000
5000
5000
N/A
Maximum7'*
300
3000
15000
15000
N/A
Removable*
20
200
1000
1000
10000
        4  As used in this table, dpm (disintegrations per minute) means  the rate of emission by radioactive material  as
determined by counts per minute measured by an appropriate detector for background, efficiency, and geometric factors associated
with the instrumentation.
        5 Where surface contamination by both alpha- and beta-gamma-emitting radionuclides exists, the limits established for
alpha- and beta-gamma-emitting radionuclides should apply independently.
        6  Measurements of average contamination should not be averaged over an area of more than 1  m2.  For objects of
smaller surface area, the average should be derived for each such object.
        7 The average and maximum dose rates associated with surface contamination resulting from beta-gamma emitters should
not exceed 0.2 mrad/h and 1.0 mrad/h,  respectively, at 1 cm.
        8 The maximum contamination  level applies to an area of not more than 100 cm2.
        9 The amount of removable material per 100 cm2 of surface  area should be determined by wiping  an area of that size
with dry filter or soft absorbent paper, applying moderate pressure, and measuring the amount of radioactive  material on the
wiping with an appropriate instrument of known efficiency. When removable contamination on objects of surface area less than
100 cm2 is determined, the activity per unit area should be based on the actual area and the entire surface should be wiped.  It
is not necessary to use wiping techniques to measure removable contamination levels if direct scan surveys indicate that the total
residual surface contamination levels are  within the limits for removable contamination.
        10  This category of radionuclides includes mixed fission products, including the Sr-90 which is present in them.  It does
not apply to Sr-90 which has been separated  from the other fission products or mixtures where the Sr-90 has been enriched.
        11  Property recently exposed or decontaminated should have measurements (smears) at regular time intervals to ensure
that  there is not  a build-up of contamination over time.  Because tritium typically penetrates  material it contacts, the  surface
guidelines in group 4 are not applicable to tritium. The Department has reviewed the analysis conducted by the DOE Tritium
Surface Contamination Limits Committee  ("Recommended Tritium Surface Contamination. Release Guides," February 1991), and
has assessed potential doses associated  with the release of property containing residual tritium. The Department recommends
the use of the stated guideline as.an interim value for removable tritium.  Measurements demonstrating  compliance of the
removable fraction of tritium on surfaces  with this guideline are acceptable to  ensure that non-removable fractions and residual
tritium La mass will not cause exposures that exceed DOE dose limits and constraints.
      3 Replication of U.S. Department of Energy, November 17, 1995, page 9.
                                                       n
                                                       1-6

-------
                                                            Preliminary Draft: June 11, 1997


       •      Apply for DOE approval. DOE field offices or headquarters may review and
              approve authorized limits and survey protocols. In general, DOE field offices
              apply the surface activity guidelines as the authorized limits for the release
              of material  that exhibits surface contamination; DOE headquarters must
              review and approve the authorized limits for the release of volumetrically-
              contaminated material.4 DOE facilities must also comply with any relevant
              NRC requirements (including those imposed  by  Agreement States).   In
              addition, DOE recommends that sites establish public participation programs,
              and requires inclusion of authorized limits in the public record.

       •      Survey materials. Once the survey protocol is approved, site managers must
              implement  the protocol to ensure that radiation  levels fall below  the
              authorized levels.

       •      Make disposition decision.  If the  activity levels fall below the authorized
              limits, the material can be cleared for unconditional use.  If not, the  site
              manager must  consider the viability of other disposition options (e.g.,
              decontamination to release limits, restricted recycle, or burial).

       DOE recently promulgated a  policy encouraging recycling initiatives within the DOE
complex (commonly referred to as "Recycle 2000").5 The policy encourages sites to decontaminate
and release material  for unrestricted use if it is. economically feasible to do so. In instances where
unconditional clearance is not economical, the  policy suggests that the metal should be used in
restricted recycling initiatives to fabricate low-level waste containers.

       In addition, DOE is promulgating new regulations under 10 CFR 834 to formalize and clarify
the standards and release guidance established by DOE Order 5400.5.6 DOE is also developing a
set of guidance documents that will detail the  procedures site managers must follow to release
material from nuclear facilities.
  4 The November 1995 DOE guidance indicates that any release of volumetrically-contaminated
materials that may lead to doses in excess of 1 mrem (individual)  or 10 person-rem (collective)
annually must be approved by DOE headquarters.  Materials that may lead to doses below these
levels may  be released with approval of the relevant DOE field office, in coordination with the
program offices.

  5 Memorandum from Alvin Aim, Assistant Secretary for Environmental Management. "Policy on
Recycling Radioactivety Contaminated Carbon Steel." U.S. Department of Energy. September 20,
1996.

  6 Notices for 10 CFR 834 can be found in the Federal Register (March 25,1993,58 FR16268 and
August 31, 1995, 60 FR 45381).

                                            1-7

-------
                                                            Preliminary Draft: June 11, 1997
 Current NRC Requirements
       NRC-licensed facilities must also comply with a variety of requirements to release scrap
metal for unconditional use. Regulatory Guide 1.86, which provides guidelines for nuclear material
licensees or reactors interested in terminating their operating licenses and releasing the site for
unrestricted use, is often applied to decisions to release scrap metal.7 Regulatory Guide 1.86 forms
the basis for DOE Order 5400.5 and is similarly focused on protecting public health and safety by
limiting exposure to radioactivity.

       NRC applies somewhat different release criteria to commercial nuclear power plants than
to decontamination and waste management firms. These latter firms also may be subject to state
level  requirements.  In addition to specific release criteria, both decontamination and waste
management firms and power plants are required to use "procedures and engineering controls based
upon sound radiation control principles to achieve occupational doses and doses to members of the
public that are as low as reasonably achievable (ALARA)."8


Guidelines for Nuclear Power Plants

       Nuclear power plants generalty do not release materials with detectable levels of radiation.
Regulatory Guide 1.86, published in 1974, includes acceptable surface contamination limits for
decommissioning based on the detection limits of the technology available at the time, and indicates
that licensees must demonstrate  that "reasonable effort has been  made to  reduce residual
contamination to as low as practicable  levels."9  (Activity levels for volumetric contamination  are
not explicitly addressed.) According to NRC personnel, this guidance is generally interpreted as
meaning that material  cannot be  released if it contains detectable levels of radiation. Residual
activity levels in released materials may be lower than those contained in Regulatory Guide 1.86 if
the licensee employs more sensitive surveying techniques than those applied when the guidance was
developed,

       NRC clarified its guidelines for releasing material hi IE Circular 81-07, which it developed
to "establish operational detection levels below which the probability of any remaining, undetected
contamination is negligible  and can be disregarded, when considering the practicality  of detecting
  7 U.S. Atomic Energy Commission, Regulatory Guide 1.86: Terminations of Operating Licenses
for Nuclear Reactors. Washington, D.C., June 1974. The guidance is also used to release portions
of sites without terminating a license.

  g JQ £2^ 2Q^ "standards for protection Against Radiation," page 291.

  9 U.S. Atomic Energy Commission, Regulatory Guide 1.86, June 1974, page 3-5.

                                           1-8

-------
                                                            Preliminary Draft:  June 12, 1997


and controlling such potential contamination and associated negligible radiation doses to the
public."10  This guidance generally mirrors the guidelines established by Regulatory Guide 1.86.


Guidelines for Waste Management and Decontamination Firms"

       NRC release requirements for waste management and decontamination firms differ from
those for nuclear power plants. While power plants are prohibited from releasing materials with any
detectable levels of radioactivity, waste management and decontamination firms can release material
at the limits specified in their individual licenses even if that limit is detectable. In general,  these
limits are similar to the Regulatory Guide 1.86 guidelines.

       In addition, twenty-nine states -have formed agreements with NRC to assume regulatory
responsibility for materials licensees. Agreement states regulate waste management services and
disposal sites directly, conducting site inspections and issuing licenses for these sites.  These states
do not have the authority to create new regulations for power plants, which are directly regulated
by NRC.  The state regulations affecting waste management and decontamination firms are  often
identical to those promulgated by NRC, but may be more stringent.
Release Process

       To release scrap metal, NRC faculties must meet the limits in their licenses and comply with
ALARA requirements.  The process involves several steps, previously illustrated in Exhibit 1-2.

       •      Determine whether material is contaminated. Scrap metal originating within
              the radiation control area is assumed to be potentially contaminated, as is
              material  located in any area exposed to radiation through  an accidental
              release.  Otherwise, material is not subject to the formal release guidance,
              although site managers may apply these requirements at  their discretion.

       •      Identify allowable levels of radioactivity.  As discussed previously, residual
              activity levels must generally be .below detection limits (for nuclear power"
              plants) or the license-specific surface requirements based on Regulatory
              Guide  1.86  (for  waste   management  and  decontamination  firms).
              Vohimetricalty-contaminated materials are generally  not releasable.   In
  10  U.S. Nuclear Regulatory  Commission.    IE Circular 81-07, "Control  of Radioactively
Contaminatcrf MaterraT'ltfay 14, 1981. In addition, IE Information Notice 85-92 (1985) describes
in detail the type of detection equipment that should be used to ensure protection from radiation.

  11 Note that requirements for waste management 'and deddhtaminatibn firms' witi" affect releases
from DOE sites as well as NRC licensees, because DOE sites may contract with an NRC-licensed
decontamination firm to process and release their scrap metals.

                                            1-9

-------
                                                             Preliminary Draft: June 12, 1997


              addition, facilities must ensure that materials meet ALARA requirements.
              We  provide  the  Regulatory Guide 1.86 surface activity  requirements in
              Exhibit 1-4.  Like the DOE Order 5400.5 guidelines, these limits are based
              on survey technology and not on dose levels.
                                 /
       •      Survey materials. Facility managers must survey materials to ensure that
              radioactivity levels are below the allowable levels for unconditional clearance.

       •      Make disposition decision.  If activity levels are below the allowable limits,
              the scrap metal is assumed to be non-radioactive and can be released.  If
              radioactivity  is detected, the scrap metal must either  be decontaminated
              before release or subject to disposal or restricted recycling.

       The above discussion applies only to materials that exhibit surface contamination. NRC does
not currently have explicit standards for the release of volumetrically-contaminated materials, and
such materials are rarefy, if  ever, released.
EPA Rulemaking

       EPA is currently in the initial stages of developing its radiation protection standards for the
release of scrap metal from nuclear facilities.  For the purpose of this preliminary analysis, we
assume that the EPA rulemaking will have the following general characteristics.

       •      Focus on Scrap. Metal:  While existing guidance addresses the release of a
              variety of types of materials and real property, EPA's rulemaking will focus
              on scrap metal.

       •      Use  New, Targeted  Risk Model:   While available analyses suggest that
              existing guidance does not lead to unacceptable risks,  current guidance is
              based  primarily  on detection limits.   The  regulatory options EPA is
              considering are likely to be dose-based  and derived  from a  risk model
              developed specifically to support its rulemaking.

       •      Include Volumetric Standards: Once final, EPA's regulations would likely
              provide standards for both surface and volumetric contamination. Current
              guidance focuses  primarily on surface contamination.
                                            1-10

-------
                                                                     Preliminary Draft:  June 11, 1997


                                             Exhibit 1-4

          REGULATORY GUIDE 1.86 SURFACE CONTAMINATION GUIDELINES12
Nuclide*
U-nat, U-235, U-238, and
associated decay products*
Transuranics, Ra-226, Ra-
228, Th-230, Th-228, Pa-231,
Ac-227, 1-125, M29
Th-nat, Th-232, Sr-90, Ra-
223, Ra-224, U-232, 1-126, 1-
131, 1-133
Beta-gamma emitters
(nuclides with decay modes
other than alpha emission or
spontaneous fission) except
Sr-90 and others noted above
Average1*
5,000 dpm a/100 cm2
100 dpm /100 cm2
1000 dpm /100 cm2
5000 dpm /Sy/100cm2
Maximum'*1
15,000 dpm a/100 cm2
300 dpm /100 cm2
3000 dpm /100 cm2
15,000 dpm /Sy/100 cm2
Removable1"
1000 dpm a/100 cm2
20 dpm /100 cm2
200 dpm /1 00 cm2
1000 dpm 0Y/100 cm2
' Where surface contamination by both alpha- and beta-gamma-emitting radionudides exists, the limits established
for alpha- and beta-gamma-emitting radionudides should apply independently.

b As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive  material  as
determined by correcting the counts per minute measured by an appropriate detector for background, efficiency, and
geometric factors associated with the instrumentation,

c Measurements of average contamination should not be averaged over an area of more than 1 m2.  For objects  of
less surface area, the average should be derived for each such object.

d The maximum contamination level applies to an area of not more than 100 cm2.

e The amount of removable material per 100 cm2 of surface area should be determined by wiping that area with dry
filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the
wipe with an appropriate instrument of known efficiency.  When removable contamination on objects of less surface
area is determined, the pertinent levels should be reduced proportionately and the entire surface should be wiped.
  12 Replication of U.S. Atomic Energy Commission, June 1974, page 5.

                                                 1-11

-------
                                                             Preliminary Draft:  June 11, 1997


        •     Exclude NORM: EPA will not address naturally occurring and accelerator
              produced  radioactive  materials  (NARM,  including  NORM)  in  this
              rulemaking due to other ongoing analysis (e.g., by the National Academy of
              Sciences) in this area.

        •     Maintain Current Implementation Responsibilities: We assume that current
              implementation responsibilities will not change.  The rule will apply directly
              to  DOE  facilities,  and  DOE will  determine  related  implementation
              requirements  (e.g., for measurement and documentation).    NRC  will
              determine whether to apply the EPA  requirements or develop its own
              standards, and will establish its own implementation requirement. Note that
              the analysis in this report assumes that the NRC standards will be identical
              to the EPA standards.

     •   Given these assumptions, the EPA standards are likely to affect primarily the numerical
release limits applied to scrap metals, not the other requirements discussed in the previous sections.
For example, requirements related to ALARA and to survey procedures are unlikely to be affected
by  the  rulemaking.  The major effects of the rulemaking will  therefore be:   (1) to increase or
decrease the quantity of scrap metal exhibiting surface contamination that is released (depending
on whether the standards are higher or lower than those resulting from current guidance), and (2)
to simplify (and hence  probably increase)  release of  volumetricalfy-containinated materials  by
establishing specific release standards.
ANALYTIC OPTIONS

       The purpose of the cost-benefit analysis discussed in this document is to compare the effects
of alternate clearance standards to the effects of current requirements. This analysis requires first
estimating likely current and future practices under  existing guidance, then comparing these
"baseline" practices to the practices likely under alternate  standards.

       The analysis considers three options  developed  to  allow comparison of the effects of
significantly different clearance levels. These three options are designed to ensure that the annual
radiation dose to the reasonably maximally exposed (RME) individual would be below 0.1 rarem,
1.0 mrem, or 15.0 mrem, respectively. The risk model used to develop these standards  is described
in Chapter 5.

       In this section, we identify the activity levels that correspond to the dose limit under each
option and compare these  activity levels to current release limits.13  Note that the three options
  13 Due primarily to differences in the type of radiation emitted (along with other factors), the
activity levels associated with a given dose vary depending upon the radionuch'de of interest.  For
example, a surface activity level of 900 dpm/lOOcm2 is equivalent to a dose of 1.0 mrem for Co-60,
while a 1.0 mrem dose is equivalent to an activity level of 9,000 dpm/100 cm2 for Cs-137.

                                         '  1-12

-------
                                                             Preliminary Draft:  June 11, 1997

                                                                                    \.
considered  were  developed for analytic  purposes only  and do  not take  into  consideration
implementation concerns such as the feasibility of detecting radioactivity at corresponding activity
levels. In addition, the release limits are derived from analysis of recycling of carbon steel (which
represents the  majority of the scrap metal potentially affected by the rulemaking); EPA hopes in
subsequent  analyses to assess the risks associated with other metals that may be affected by the
standards.

       Exhibit 1-5 illustrates the RME dose under existing standards (Regulatory Guide 1.86 and
DOE Order 5400.5) for the 44 radionuclides included in the analysis, and compares these doses to
the dose limit  specified under each of the three analytic  options.   As indicated by  the exhibit,
existing standards  lead to annual RME doses below the 15.0 mrem level for all radionuclides, and
below the 1.0 mrem level for 31 of the 44 radionuclides. The current standards lead to higher doses
than the 0.1 mrem level in 26 of the 44 cases.

       The  relationship of the  three analytic  options to current regulatory requirements has
important implications for the effect of EPA's standard on human health risks. Assuming that the
materials released exhibit the maximum activity allowed under each option, a 15.0 mrem standard
would, lead to an increase in risks for all radionuclides  included hi this analysis. In contrast, a 0.1
mrem standard would lead  to a decrease in risks, while the implications of a  1.0 mrem standard
would depend upon the radionuclide in question.14

       Appendix A lists the activity limits associated with each set of release standards.  The first
page of the appendix provides  the surface  release limits under each option.  The second  page
provides the volumetric release limits.
  14 Although Exhibitl-5 indicates thil >sdme tadionuclides do have tower release limits under
current guidelines compared to the 0.1 mrem standard (e.g., Ni-59), none of these radionuclides is
explicitly assessed in this analysis.

                                        ,    1-13

-------
                                   Annual RME Dose (mrem/year)
                                              So
                               —  -    -    '•  •£
p.
o.
s

i
I
5

§
a

1
5
I
o
.....%, i
Btf-152 '••'•-
Co-60
U-series
AgllOm+D
Th-232
Mn-54
U-235+D
U-234
U-Separ.
U-Deplete
U-238+D
Cs-134
Th-Series
Zn-65
Cs-137+D
Sb-125+D
Pb-210-+D
Sr-90+D
Pa-231
Np-237+D
Ra-226+D
Ac-227+D
Ru-106+D
Am-241
Ce-144+D
Th-229+D
1-129
Pu-239
Pu-240
Pu-242
Pu-238
Cm-244
Th-230
Th-228+D
Ra-228+D
C-14
Pu-241+D
Pm-147
Mo-93
Tc-99
Ni-63
Fe-55
Ni-59

i

•


|
•
t
t
1 4
I 4
^
,
,
1 ^
1 A.
^T
,
»
4
4
4
ilk
t\
I \
/ \
*

* / §
^*^_ .0 |
• 1" 1 §•
1 A 	









^
k.
^^
k*















•









^
^^
^
V










> k
• \ A
. \ \
- \ \


-^ u»
o b
i i
• 
-------
                                                             Preliminary Draft: June 11, 1997
ANALYTIC APPROACH
       The analysis  discussed in the subsequent chapters of this report begins by identifying the
scrap metal potentially affected  by the regulations, then determines the change in the amount of
scrap metal likely to be cleared for unconditional use under each of the three analytic options. It
next assesses the changes in risks to human health and the environment attributable to each option,
and also assesses the options' effects on a range of other factors, such as conditions in scrap metal
markets. Exhibit 1-6 provides an overview of the conceptual approach for the .cost-benefit analysis,
each component of which is discussed in detail in the following chapters.

       •      Identify and Characterize Potentially Affected Metals (Chapter 2). The first
              step in the analysis is to identify and characterize  the metals likely to  be
              affected by EPA's clearance standards.  For the preliminary analysis, we
              focus  on scrap  metal potentially  available over the next 55 years  from  11
              major DOE facilities and 123 NRC-licensed commercial nuclear power
              reactors. The analysis addresses only those materials for which disposition
              decisions may change as a result of the rulemaking; e.g., it excludes metals
              from outside radiation control areas, which, provided they have not  been
              accidentally exposed to radiation, can be released regardless of the standards
              established by EPA.

       •      Characterize Baseline  and  Post-Regulatory  Practices  and  Unit  Costs
              (Chapter 3). The second step involves characterizing disposition options and
              related costs under the baseline and analytic options. The initial quantitative
              analysis considers two  disposition  options:  disposal  and unconditional
              clearance (including  any needed decontamination).  For each scenario, we
              developed  estimates of  unit costs  (i.e., costs  per metric ton) based on
              available data sources.   We also include a qualitative  discussion of the
              potential impacts of restricted recycling on our analysis.

       •      Characterize Changes in Quantities Released and Related Costs (Chapter
              4).  The third step in the analysis involves estimating the quantities of scrap
              metal likely to be disposed  or  cleared for unconditional use  under the
              baseline and each analytic option, and the associated costs.  This analysis
              relies primarily on an economic model that assumes that facility managers
              will select the least costly approach  for managing their scrap metal. We
              believe that this model provides high-end estimates of the quantities of scrap
              metal likely to be released, because a number of non-economic factors (such
              as concerns about adverse public reactions) may lead facility managers to
          " -  prefer alternate approaches.    — -.-       --—- -

       •      Assess Effects on Human Health and the Environment, (Chapters 5 and 6).
              Changes in, the quantities and characteristics of 'scrap, metal cleared for"
              unconditional use will in turn affect related risks. Chapter 5 focuses on the
              changes in cancer risks  attributable to each of the options assessed.  The
              effects on other risks are assessed qualitatively in Chapter 6.
                                            1-15

-------
                            Exhibit 1-6
                   CONCEPTUAL APPROACH
                             Identify and
                           characterize metal
                       potentially affected by the
                              standards
    Characterize
baseline practices and
      unit costs
                                                      1
                                       Characterize
                                 post-regulatory practices
                                      and unit costs
                          Estimate change in
                             scrap metal
                           quantity released
                           and related costs
Assess effects
 • on human ,
health arid the
 environment'
                                          Assess market
                                            and other '>
                                             impacts • -

-------
                                                           Preliminary Draft:, June 11, 1997


       •      Assess Market and Other Impacts  (Chapter 6).   Changes in scrap metal
              management practices may also affect scrap metal markets and the demand
             .for virgin materials, as well as the demand for disposal capacity. This initial
              analysis provides a qualitative discussion of these issues.

The final chapter (Chapter 7) summarizes the results of the analysis, comparing estimated scrap
management costs and predicted cancer risks under each of the options considered.

       EPA may refine the data and models used in this preliminary assessment and/or expand the
analysis in the future to  address a number of issues not yet assessed in detail. If EPA develops a
proposal that qualifies as a major rule, the Agency will prepare a full Regulatory Impact Analysis
(RIA) that meets the requirements of Executive Order 12866.15  This RIA would provide a more
comprehensive assessment of the proposed rule and other options. It would also assess other issues
required by government-wide and EPA guidance, such as impacts on small businesses, information
collection'burdens, environmental justice, and burdens on state governments.
  15 Guidance relating to Executive Order 12866 is contained in:  Memorandum to Members of the
Regulatory Working Group, from Salty Katzen, U.S. Office of Management and Budget. "Economic
Analysis of Federal Regulations Under Executive Order No. 12866." January 11,1996.

                                          1-17

-------
Page Intentionally Blank

-------
                                                             Preliminary Draft: June 13, 1997
SCRAP CHARACTERISTICS                                                   CHAPTER 2
INTRODUCTION AND SUMMARY

       The first step in the analysis of EPA's radiation protection standards involves identifying the
materials potentially affected by the rulemaking. These materials include scrap metal generated by
Federal and nonfederal nuclear facilities. Such scrap may be generated by routine operations or
through the demolition of buildings and equipment during decontamination and decommissioning.
The analysis focuses on materials for which disposition decisions may change as a result of the EPA
rulemaking. For example, scrap that is "clean" and can be released for unconditional use regardless
Of the clearance standards  is excluded from the analysis, as is scrap that contains such high levels
of radioactivity that decontamination and release is infeasible.

       This chapter discusses the methodology for collecting information on scrap metal potentially
affected by the rulemaking and provides preliminary data on major scrap sources. It focuses on the
11 key DOE facilities and  123 commercial nuclear power reactors  that are likely to be the major
sources of scrap affected by the rulemaking. It also provides summary information on those physical
and radiological characteristics that  most significantly affect decisions to  release  scrap  for
unconditional  clearance,  such as scrap source, metal type, physical form, and radioactivity levels.
Also included  are estimates of the timing of availability  for potential release.

       The chapter is divided into five major sections. The remainder of this section provides a brief
summary of the available data. We next describe the analytic approach used to collect information
on potentially  affected scrap metals. The third and fourth sections discuss the data compiled for the
major DOE facilities and NRC-licensed power reactors, respectively. In the final section, we discuss
key uncertainties related to these data, as well as possible directions for future research.
Summary of Available Pate- ••   -    •--  -•     *  — •    "i~~——  — •**" --'••.
                     > '  '      '          "     t      '     * i
       The preliminary analysis focuses on  1.6 million  metric tons1 of scrap  metal likely to be
generated by decontamination and decommissioning activities at 11  major DOE facilities and 123
                                            2-1

-------
                                                             Preliminary Draft: June 13, 1997


commercial nuclear power reactors between 1998 and 2053.1 This estimate is based on data from
each of the individual DOE facilities and from a reference boiling water reactor (BWR) and
pressurized water reactor (PWR)  scaled to represent each of the 123 NRC-licensed commercial
nuclear power reactors.

       As indicated in Exhibit 2-1, 59 percent of the scrap assessed is expected to be generated by
the major DOE facilities, with the remaining 41 percent to be generated by the commercial nuclear
power reactors. Three DOE sites (Oak Ridge, Paducah, and Portsmouth) account for 78 percent
of the DOE scrap, while the 83 PWRs account for 53 percent of the NRC scrap.
                                                                 1       i     I
       Most of the scrap (over 80 percent) is carbon steel; other potentially affected metals include
nickel, aluminum, stainless steel, galvanized iron, and copper. Little information is available on the
radiological characteristics of these  metals. For the  preliminary analysis, we  focus on selected
indicator nuclides - including U-238, Cs-137, and Pu-239 for DOE facilities (with activity levels
ranging from 100 dpm/100 cm2 to 100,000,000 dpm/100 cm2); and Co-60, Cs-137, and Ru-106 for
NRC power reactors (with activity levels ranging from 20 dpm/100 cm2 to 130,000,000 dpm/100 cm2).

       The timing of scrap availability differs significantly across the two sectors. About 45 percent
of the DOE metal is either currently in storage or likely to be generated before the year 2000. In
contrast, decontamination and decommissioning activities are just beginning for NRC-licensed power
reactors, and significant quantities of scrap are not likely to be generated until after the year 2020.
The estimated final year of scrap generation is 2038 for the  DOE facilities and  2053 for the
commercial power reactors.
Key Uncertainties and Plans for Future Research

       The purpose of this preliminary analysis is to present the methodology for assessing the cost
and benefits of EPA's rulemaking as well as initial information on the possible effects of alternate
clearance standards. The scrap data presented in this chapter are derived from existing data sources
and subject to considerable uncertainty.  EPA hopes to improve the quality  of these data  in
subsequent phases of the analysis.

       The data we report are likely to understate the total quantities of scrap metal potentially
affected by the rulemaking. DOE facilities (as well as facilities operated by other Federal agencies)
are likely to generate significant quantities of scrap not included in this analysis, and the analysis
does not consider scrap from NRC licensees other than power reactors. In addition, the data on
scrap characteristics are uncertain, particularly the radiological profiles. This uncertainty may lead
us to under-  or  overstate the proportion of the scrap  that could be released under  alternate
clearance standards. The-remainder of this chapter explores the data and related uncertainties in
more detail.
    1 We assume for the purpose of analysis that EPA's rule would become effective in 1998.

                                            2-2

-------
                                    Exhibit 2-1
    PRELIMINARY ESTIMATES OF POTENTIALLY AFFECTED SCRAP METAL QUANTITIES
                                    (1998-2053)
Major DOE Facilities
• Maior NRC Licensees
Femald
4,200 tons

Hanford
92,200 tons
i
Idaho National
Engineering Lab
1£i TAA 4-^wn
Jo. /uu tons

Los Alamos

INallUildl L^dl}
3. 100 tons
-
Nevada Test Site
300 tons
, . i
Oak Midge
252,300 tons

Paducah
279^00.tons

Portsihouth
198,000 tons

RockyfFlats
26,30fttons
'«
Savanna^ River
logoff tDM

Weldon Spring
27,800 tons








































-
^^^^ - ^^^^^^^^
1
DOE Total 1 NRTTntal
^^ U\Jl-i M. Ulul • INIVi.j 1 ULa.1

936,300 tons I 64 1,000 tons



i

^^^^^^^^^m
^^DOEUncIB
^1 ' ' H
^^H NRC Total H-^-1
1 1 577 300 tonsB
^^^^^^^^^^^^^^^^•^H
i




,








                                                                                 Reference





rauiiiucs
299,800 tons

83PWR
rdL-lllUCb
341,200 tons
•^ —

^
^~

BWJK.
8,900 tons

Referencee
P\H7T?
r WK.
4,400 tons

-------
                                                            Preliminary Draft: June 13, 1997
 ANALYTIC APPROACH
       To assess the quantities and characteristics of scrap metal potentially affected by EPA's
 rulemaking, we followed a two  step process.2  First, we conducted a screening analysis to identify
 the major sources of scrap potentially affected by the rulemaking. Second, we collected data on
 individual scrap sources and compiled them into a detailed database.

       The screening analysis  led us  to  focus initially on  11 major DOE sites and  the 123
 commercial power reactors licensed by the NRC. The 11 DOE facilities include the Fernald site, the
 Hanford site, the Idaho  National Engineering Laboratory (INEL), the Los Alamos  National
 Laboratory (LANL), the Oak Ridge facilities (K-25, Y-12, and Oak Ridge National Laboratories -
 ORNL), the Nevada Test site,  the Paducah site, the Portsmouth Gaseous Diffusion Plants, the
 Rocky Flats Environmental Technology site, the Savannah River site, and the Weldon Spring site.3
 The NRC licensees assessed include 123 commercial light water reactors, including 40 BWRs and
 83 PWRs.

       For the DOE sites, detailed data were derived from previous DOE studies, supplemented
 by site visits and interviews with DOE personnel.4  These data include information on scrap metal
 quantities currently in inventory as well as quantities that are  likely to be generated during future
 decontamination  and decommissioning  activities.  The available data sources generally provide
 information on metal type, quantity, and physical form (e.g., "structural steel" or "piping")  for each
 scrap  source, although physical  form data are missing in some cases. Scrap in inventory generally
 is being stored while the facilities consider disposition options; we assume that these quantities will
 remain in storage until EPA's rulemaking becomes effective (i.e., in 1998). For scrap resulting from
 future decontamination and  decommissioning activities, the year available is determined based on
 site-specific planning documents and other sources.
    2 The scrap data reported in this chapter were collected by S. Cohen and Associates, who also
developed and implemented the methodology for estimating radiological characteristics.  See: S.
Cohen and Associates, Technical Support Document:  Evaluation of the Potential for Recycling of
Radioactivelv-Contaminated Scrap Metal. Prepared for the U.S. Environmental Protection Agency,
forthcoming.

    3 Pinellas was originally included in the analysis, but subsequently eliminated because scrap
disposition decisions will be made before EPA's rulemaking becomes effective.

    4 The primary data sources include:  (1) U.S. Department of Energy.  Taking Stock:  A Look
at the Opportnnities and' Challenges from the Cold War Era.  January 1996; (2) O.A. Person, e't al..
G^eous Diffnsiory Facilities Decontamination and Decftmntissidfling Report. U.S. Department of
Energy. December 1995; (3) S. Cohen and Associates.  Scrap Metal Inventories at U.S. Nuclear
Facilities Potentially Suitable for Recycling.  U.S.  Environmental Protection Agency. September
1995; and, (4) U.S.  Department of Energy.  U.S.  Department of Energy Scrap Metal Inventory
Report. March 1995.

                                           2-4

-------
                                                            Preliminary Draft:  June 13, 1997


       For the NRC-licensed commercial reactors, the data are derived from studies of reference
reactors. The Washington Public Power Supply System Nuclear  Project No. 2 in Richland,
Washington is the reference BWR, and Trojan Nuclear Power Plant hi Oregon is the  reference
PWR. These data are extrapolated to the universe of 123 reactors using weighting factors based on
each reactor's size and power rating.5 Larger reactors receive a larger weighting scale and therefore
are estimated to contain more scrap metal than smaller reactors. The data sources for these reactors
provide detailed information on metal type, quantities, and physical form. We assume that scrap
from these facilities will become available  10 years after license expiration; i.e., the reactor will not
be demolished until radioactivity levels are allowed to diminish.

       Available data sources contain little information on the radiological profile of these metals.
For the preliminary analysis, we use a simplified approach to estimate related data. We selected
indicator radionuclides for each site that illustrate the risks posed by the scrap and the feasibility and
cost of decontaminating it (if necessary) to meet alternate clearance standards. For each DOE site,
we  select indicator radionuclides based on the site's operating history and available data. For each
radionuclide, we then estimate ranges of likely activity levels based on knowledge of site operations
and the year when the scrap is likely to become available for disposition.6 For the commercial
power reactors, we focused on radionuclides and corresponding activity levels likely to be present
10 years after license termination. We estimated activity levels based  on process  knowledge of
individual component systems, and scaled them to account for deterioration over the 10-year storage
period.7
   5 The primary reference reactor studies used are:  (1) U.S. Nuclear Regulatory Commission.
Revised Analyses of Decommissioning For the Reference Pressurized Water Reactor Power Station.
1994; and (2) U.S. Nuclear Regulatory Commission. Revised Analyses of Decommissioning For the
Reference Boiling Water Reactor Power Station. 1995.

   6 The primary data sources employed to generate site-specific radiological profiles are:" (1) U.S.
Department of Energy.  U.S. Department of Energy's Weapons Complex Scrap Metal Inventory.
Prepared by John R. Duda, U.S. DOE Morgantown Energy Technology Center.  July 1993; (2) S.
Cohen and Associates.  Scrap Metal Inventories at U.S. Nuclear Facilities Potentially Suitable for
Recycling.   Prepared for U.S.  Environmental  Protection Agency.  September 1995; (3) U.S.
Department of Energy.  Assessment of Risks and Costs Associated with Transportation of DOE
Radioactively  Contaminated  Carbon Steel.  Prepared by S.Y. Chen,  et ah, Argonne National
Laboratory, for  the Office  of  Environmental  Management, November  1995; and (4)  U.S.
Department of Energy.  INEL Metal Recycle Radioactive Scrap Metal Survey Report. Prepared
by D.M. Funk of Lockheed Idaho Technologies Company  for the Idaho  Operations  Office,
September 1994.

   7 The primary data sources employed to generate radiological profiles ^re: '(I) NUREG/CR-0130,
Technology. Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power
Station. Volumes 1 and 2, prepared by R.I. Smith et al., Pacific Northwest Laboratory, for the U.S.
Nuclear Regulatory  Commission;  (2)  NUREG/CR-0672,  Technology. Safety  and Costs of
Decommissioning a Reference Boiling Water Reactor Power Station. Volumes 1 and 2, prepared

                                           2-5

-------
                                                            Preliminary Draft: June 13, -2997


       As noted earlier, the data include only scrap potentially affected by EPA's rulemaking; the
 database is not a complete inventory of all scrap resulting from facility operations or closure.  For
 example, scrap that is free of radioactivity (i.e., from outside the facility's radiation control area) is
 excluded from the analysis. Material that is too radioactive to be cost-effectively decontaminated
 (e.g., materials from inside the reactor) are also excluded. We also exclude scrap in cases where the
 facility has made final disposition decisions prior to the time at which EPA's rule becomes effective.

       The data collection effort  resulted in  creation  of  a  database  that contains detailed
 information on the scrap metal potentially available from each facility. The metals include carbon
 and stainless  steel, along with  nickel,  aluminum, iron  and copper.  The database  also includes
 information on small quantities of graphite,  inconeL, lead, bronze, brass, and other metals.

       For each scrap stream from each facility, the database includes information on the scrap
 source (reactor  or location within the  facility);  scrap characteristics (metal type, physical form,
 weight, year available); and a radiological profile (rough  estimates of radionuclides present  and
 activity levels). In the following sections, we describe these data hi more detail.


 SCRAP METAL FROM MAJOR DOE FACILITIES

       The Federal government operates a number of facilities that use nuclear  materials, including
 U.S. Department of Defense (DOD) operations and various hospitals and laboratories operated by
 other agencies. For example, nuclear-powered naval ships and submarines generate small volumes
 of radioactive scrap metal during normal operations, and  materials within the Aberdeen Proving
 Ground and up to 146 additional DOD sites are known to be contaminated.8

       The most significant Federal source of scrap metal potentially affected by EPA's rulemaking
 is DOE. The majority of DOE scrap is likely to be generated  by  the 11 sites addressed by  this
 analysis; however, many other sites will also be  affected by  the rulemaking.9 The  following
 discussion describes the  characteristics  of the scrap generated by the 11 key sites.
by H.D. Oak, et al., Pacific Northwest Laboratory, for the U.S. Nuclear Regulatory Commission; (3)
"SAFSTOR Decommissioning Plan for the Humboldt Bay Power Plant, Unit 3," Pacific Gas and
Electric Company, July 1994; and (4)  'Trojan Nuclear Plant Decommissioning Plan, PGE-1061,"
Portland General Electric, June 1996.

    8 U.S.  Environmental Protection  Agency.  Radiation  Site Cleanup  Regulations: Technical
Support Document for the Development of Radionuclide Cleanup Levels for Soil.  EPA 402-R-96-
011 A-D (Review Draft), September 1994. Page 1-3 and Appendix A.

    9 The analysis excludes, for example: Pantex, Mound, and six additional national laboratories;
30 sites included in the Formerly Utilized Sites Remedial  Action  Program (FUSRAP); 14 sites
included in the Uranium Mill Tailings Remedial Action Program (UMTRAP); and 33 other sites
identified  under DOE's  Environmental Restoration  Program as  being contaminated  (U.S.
Environmental Protection Agency, September 1994, page 1-3 and Appendix A).

                                           2-6

-------
                                                             Preliminary Draft: June 13, 1997
       Over the next 40 years, decontamination and decommissioning activities at 11 major DOE
sites will generate an estimated 936,300 metric tons of scrap metal potentially affected by EPA's rule
(see Exhibit 2-2). Three  facilities account for 78 percent of this scrap — the gaseous diffusion
reactors  at Oak Ridge (K-25), Paducah, and Portsmouth. The next largest scrap metal sources are
Hanford and INEL, which account for  about 10 percent and four percent, respectively,  of the
available scrap quantities.
Exhibit 2-2
PRELIMINARY ESTIMATES OF SCRAP QUANTITIES
FROM SELECTED DOE FACILITIES
BY FACILITY
Facility
Fernald
Hanford
INEL
LANL
Nevada Test Site
Oak Ridge: ORNL
Y-12
K-25
Paducah
Portsmouth
Rocky Flats
Savannah River
Weldon Spring
Total
Quantity
(metric tons)
4,200
92,200
36,700
3,100
300
1,100
9,100
242,100
279,200
198,000
26,300
16,200
27,800
936,300
Percent of Total
< 1%
10%
4%
< 1%
< 1%
< 1%
1%
26%
30%
21%
3%
2%
3%
100%
Notes: Detail may not add to total due to rounding. Estimates are likely to
understate total scrap available because data are lacking on future scrap
generation for some scrap sources, particularly at Hanford and Savannah River.
       As indicated in Exhibit 2-3, we estimate that 88 percent of the  scrap  metal, by  mass,
generated by these facilities is carbon steel. Nickel, aluminum, and stainless steel are the next most
prevalent metals, together comprising an additional 12 percent of the total. Most of this metal is in
                                            2-7

-------
                                                             Preliminary Draft: June 13, 1997
the form of diffusion cells, which make up 67 percent of all DOE metal. Descriptions of the physical
form categories used hi the analysis are provided in Appendix B. Detailed information on the
physical form of affected scrap from DOE facilities is provided in Appendix C.
Exhibit 2-3
PRELIMINARY ESTIMATES OF SCRAP QUANTITIES
FROM SELECTED DOE FACILITIES
BY METAL TYPE
Metal Type
Carbon Steel
Nickel
Aluminum
Stainless Steel
Copper
Other
Total
Quantity
(metric tons)
824,000
44,900
36,600
24,600
2,300
3,900
936,300
Percent
of Total
88%
5%
4% ,
3%
< 1%
< 1%
100%
Notes: Detail may not add to totals due to rounding. "Other" includes
miscellaneous metals (undefined), graphite, lead, monel, iron, mixed copper,
and brass.
       Much of the DOE scrap metal of interest to this analysis (45 percent) is currently in storage
or will become available before the year 2000. According to available planning documents and other
sources, additional large quantities will be generated in 2008 (primarily from Portsmouth), 2016-2017
(primarily from Paducah and Savannah River) and 2038 (primarily from Hanford). The timing of
scrap generation is illustrated in Exhibit 2-4. The release of scrap metal over time is likely to be
more evenly distributed than suggested in the exhibit; however, we employ this distribution since it
is based on the best available information concerning future D&D activities at DOE facilities.

       Little information is available on the radiological characteristics of DOE scrap metal likely
to be affected by EPA's rule. As mentioned earlier, a relatively simplistic approach is followed for
this preliminary  analysis. For each site, one or two radionuclides were selected as indicators of the
types of radiation present. We also estimate the potential range of'activity levels for each site. For
this  initial analysis, all metal from the site is assumed to match  the  radiological profile, thus
developed  (see Exhibit 2-5). The" unavailability of betteridata on the  radiological characteristics of
this scrap is likely the single most significant source of uncertainty in our estimates of the impacts
of alternate EPA standards.
                                             2-8

-------
         25Q.O
                                             Exhibit 2-4



                           PRELIMINARY ESTIMATES OF SCRAP QUANTITIES

                        FROM SELECTED DOE FACILITIES: BY YEAR AVAILABLE
I §
» I

i H
* a
         200.0
         150.0;
         100.0
          50.0 -•<
              0\
o   cs

8  "§
cs   CN
                           s
                               O
                               o
oo   O
O   •—*
o   o
M   ri
oo
>-<

O
o
f»J

O
                                               o   o   o   o  o


                                                   Year Available
                                                                                            o

-------
                                                               Preliminary Draft: June 13, 1997
Exhibit 2-5
INDICATOR RADIONUCLIDES FOR SELECTED DOE FACILITIES
Site
Fernald
Hanford
INEL:
Test Reactors
Idaho Chemical
Processing Plant
Nevada Test Site
LANL
Oak Ridge:
ORNL
Y-12
K-25
Paducah
Portsmouth
Rocky Flats
Savannah River
Weldon Spring
Radionudide
U-238
U-238
Cs-137

Cs-137
Cs-137
Pu-239
U-238
Cs-137
Cs-137

Cs-137
Cs-137
U-238
U-238
U-238
Pu-239
U-238
U-238
Cs-137
U-238
Assumed Activity Level
(dpm/100 cm2)
Low
1,000
1,000
100

1,000
100
1,000
1,000
100
100

100
1,000
1,000
1,000
1,000
1,000
1,000
1,000
100
1,000
High
10,000,000
1,000,000
100,000

100,000
10,000
1,000,000
100,000
10,000
10,000

100,000
1,000,000
1,000,000
1,000,000
1,000,000
100,000,000
100,000,000
1,000,000
100,000
10,000,000
       As Exhibit 2-5 indicates, the analysis focuses on three indicator nuclides for the selected
DOE facilities:  U-238, Cs-137, and Pu-239. Activity levels vary according to  the radionuclides
present and the operating history of the facilities. Based on our analysis of these factors, the site
with the lowest estimated levels of contamination is Los Alamos, while the site with the highest
                                             2-10

-------
                                                            Preliminary Draft: June 13, 1997


estimated levels is Rocky Flats. In most cases, we assume the contamination is surficial; materials
that are volumetricalty contaminated at the point of generation are usually too highly radioactive
to be cost-effectively decontaminated and released for unconditional use.


SCRAP METAL FROM COMMERCIAL NUCLEAR POWER REACTORS

       In addition to the Federal facilities described above, EPA's radiation protection standards
are likely to affect scrap metal from facilities licensed by NRC. There are currently over 22,000 NRC
licensees,  including hospitals, academic institutions, and industries that make use of radioactive
materials. Most of these facih'ties, however, generate only small  quantities of scrap that could be
affected by EPA's rulemaking. The major sources of scrap are likely to be the nations' 123 operating
or closed commercial nuclear power reactors, all  of which are assumed to terminate operations as
their licenses expire.10

       Based on our preliminary analysis, these reactors are likely to generate approximately 641,000
metric tons of scrap that could be affected  by EPA's rulemaking over the next 60 years. These
estimates are based on data from a reference BWR and PWR, scaled to reflect the quantities likely
to be available from, each of the 123 power reactors.  Although there are fewer BWRs than PWRs
(40 vs. 83), the reference BWR reactor contains roughly twice as much available metal (8,900 metric
tons versus 4,400 metric tons). Therefore, as shown in Exhibit 2-6, only slightly less scrap  is likely
to be generated by decommissioning of BWRs than PWRs.
Exhibit 2-6
PRELIMINARY ESTIMATES OF SCRAP QUANTITIES
FROM COMMERCIAL POWER REACTORS
BY FACILITY TYPE
Facility
Reference BWR
Reference PWR
Total
All BWRs (40 facilities)
All PWRs (83 facilities)
Total
Quantity
(metric tons)
8,900
4,400
13,300
299,800
341,200
641,000
Percent of Total
67%
33%
100%
47%
53%
100%
Note: Detail may not add to total due to rounding.
    10 Some licensees may renew their licenses.  To the extent they do, we are overestimating the
volume of scrap that may be generated by these facih'ties.
                                           2-11

-------
                                                             Preliminary Draft:  June 13, 1997
       We summarize the quantities of potentially available scrap by metal type in Exhibit 2-7. As
illustrated hi this exhibit, carbon steel is the dominant metal, representing roughly 76 percent of the
potentially affected scrap. Stainless steel, galvanized iron, and copper are the next most prevalent
metals, together comprising an additional 24 percent of the total. Much of the metal is large and
small piping, which makes up 44 percent of all NRC scrap metal.  Detailed information on the
physical form of affected scrap from commercial power reactors is provided in Appendix D.11
Exhibit 2-7
PRELIMINARY ESTIMATES OF SCRAP QUANTITIES FROM
COMMERCIAL POWER REACTORS BY METAL TYPE
Metal Type
Carbon Steel
Stainless Steel
Galvanized Iron
Copper
All other
Total
Quantity
(metric tons)
487,100
121,800
18,900
10,000
3,200
641,000
Percent of Total
76%
19%
3%
2%
< 1%
100%
Notes: Detail may not add to total due to rounding. "Other" includes
inconel, lead, bronze, aluminum, brass, nickel, and silver.
       To estimate the availability of scrap metal from commercial power reactors over time, we
assume that these reactors will begin demolition activities 10 years after license termination. We
present the resulting projection of scrap availability in Exhibit 2-8. The majority of the scrap, 84
percent, will become available between the years 2020 and 2040.
                                             •
       As was the case for DOE facilities, limited information is available on the radiological
characteristics of scrap metal from commercial power reactors. We again follow a relatively simple
approach,  identifying one to three  indicator nuclides  as well as activity level ranges for each
component system within each building at the BWR and PWR reference reactors.  We summarize
these profiles hi Exhibit 2-9 (note  that individual systems within each  building have differing
estimates  of  activity levels  within each range).12  Data on other metals is provided separately,.
because the ranges address only carbon and stainless >steel/       - .—H»    .<-,••>•,.
    11 See Appendix B for detailed descriptions of the various physical form categories.

    12 In the BWR reactor building, for example, the HVAC components have a low activity level
of 130 dpm/cm2, while the fuel pool cooling system has a low activity level of 1,300,000 dpm/cm2.
                                            2-12

-------
                                     Exhibit 2-8
                     PRELIMINARY ESTIMATES OF SCRAP QUANTITIES
               FROM COMMERCIAL POWER REACTORS: BY YEAR AVAILABLE
70.0
60.0
50.0
40,0
30.0
20.0
10.0
 0.0
\
•


	 _


-
'



7*000000000



1


1


s





.
>
1
GO O 
-------
                                                              Preliminary Draft: June 13, 1997
Exhibit 2-9
INDICATOR RADIONUCUDES FOR COMMERCIAL POWER REACTORS
Reactor Type
BWR
PWR
Location
Reactor Building
(steel only)
Radwaste Building
(steel only)
Turbine Building
(steel only)
Other Metals
Reactor Building
(steel only)
Auxiliary and Fuel
Storage Building
(steel only)
Other Metals
Radionuclide
Co-60
Cs-137
Ru-106
Co-60
Cs-137
Ru-106
Co-60
Cs-137
Co-60
Cs-137
Co-60
Cs-137
Co-60
Cs-137
Co-60
Assumed Activity Level
(dpm/100 cm2)
Low
130
27
27
130
27
27
130
27
130
27
86
29
86
20
86
High
130,000,000
27,000,000
2,700
130,000,000
27,000,000
2,700
1,300,000
270,000
13,000
2,700
86,000,000
950,000
86,000,000
950,000
8,600
       As illustrated in the exhibit, the activity levels for indicator radionuclides may be somewhat
higher for the BWR than the PWR. For the BWR, the reactor and radwaste buildings contain the
highest levels of contamination. Similarly, the reactor and auxiliary and fuel storage buildings contain
the highest levels of contamination at PWR facilities. The possible activity levels vary considerably
at both types of reactors, ranging from a low of 20 dpm/100 cm2 to 130,000,000 dpm/100 cm2. We
assume the metal  is only surficially contaminated; material that is volumetrically-contaminated  is
usually too radioactive to be cost-effectively decontaminated and released.
                                            2-14

-------
                                                             Preliminary Draft; June 13, 1997

          s
KEY UNCERTAINTIES AND PLANS FOR FUTURE ANALYSIS

       The data collected to date on scrap characteristics is limited in scope and contains some
uncertainties. EPA may assess the effects of these uncertainties and/or collect additional data and
refine the estimates in future stages of the analysis. Below, we discuss the level of uncertainty in
each major data element.

       •      Total quantities are probably understated:  Because the -preliminary analysis
              considers only a  subset  of all  facilities potentially affected by EPA's
              rulemaking, it is likely to understate the quantities of scrap affected. In
              addition, data on future scrap generation are not available for some of the
              major DOE sites. These sources of understatement may be counterbalanced
              or augmented to some (unknown) extent by other sources of uncertainty. For
              example, DOE is currently  considering policies that would lead to re-use
              rather  than  dismantling  of some of its  major facilities, decreasing the
              availability of scrap. Plans for most commercial power reactors are also highly
              uncertain because decontamination and decommissioning activities will not
              be initiated for many years.

       •      Data on metal types appear reasonably accurate: In general, we believe that
              the data  on metal type are reasonably accurate. These data are missing for
              a relatively small percentage of the  database and (except for uncertainty
              regarding the percentage of each metal in certain mixed metal components)
              are reasonably well understood.

       •      More detailed data on physical  form are available  for commercial power
              reactors  than for DOE sites:  The reference reactor studies used for the
  >           analysis of commercial power reactors provide very detailed information on
              individual components (e.g., the number of 10 inch valves), while the DOE
              data are generally more aggregated (Valves and piping")* The DOE data in
              some  cases also lack information on physical  form. Although individual
              reactors may differ in many respects from the reference reactors, the physical
              form data for the NRC sector appear much less uncertain than 'the DOE
              data.

       •      Data on  year available are very uncertain:  Plans for decontamination and
              decommissioning are in  flux for both DOE facilities and  NRGJicensed
              reactors.  Hence, the data on the year in which scrap may become available
              for release  or disposal are very uncertain. While the uncertainty is less for
              scrap  that is  already available (e.g., currently stored at  DOE sites), it is
              unclear when the sites will make disposition, decisions even for this scrap.
•  f'.                         'i,                           ii  ill    " •
                                           2-15

-------
                                                              Preliminary Draft: June 13, 1997


       •      The greatest uncertainty concerns information on radiological characteristics:
              Data on the radiological characteristics of potentially affected scrap (both
              radionuclides present and prevailing activity levels) are highly uncertain and
              based on simplifying assumptions, professional judgement, and limited data.
              This uncertainty is reflected in the small number of nuclides considered and
              the wide ranges of activity levels assessed in the preliminary analysis.

       As discussed in more detail in the following chapters, uncertainty in these factors leads to
uncertainly in our estimates of the costs and ultimate effects of EPA's rule.
                                             2-16

-------
                                                           Preliminary Draft: June 13, 1997
DISPOSAL AND, RECYCLING COSTS                                         CHAPTER 3
INTRODUCTION AND SUMMARY

       The next step in the preliminary analysis of EPA's radiation protection standards involves
determining the costs associated .with unrestricted recycling and other disposition options.1 For the
initial analysis, we focus on two scenarios:

       •      Permanent disposal as low level radioactive waste.

       •      Unconditional clearance, with or without prior decontamination.

We also discuss  the impact of  a  third option, restricted recycling (e.g., the use of scrap to
manufacture nuclear waste containers), on our findings.

       This chapter presents our initial unit cost estimates for each scenario. We first summarize
our findings, then provide an overview of our analytic approach. We next describe the cost estimates
for each disposition scenario. The final section discusses the key uncertainties in these estimates and
plans for future analysis.
Summary of Findings

       The preliminary analysis includes a quantitative assessment of two scrap disposition options,
disposal and unconditional clearance, and a primarily qualitative discussion of the impacts of a third
option, restricted recycling. Because of significant uncertainties in disposal costs, we consider both
high and low cost estimates for the disposal of materials from DOE facilities and  NRC-licensed
nuclear power plants. For unconditional clearance, we include two scenarios:  direct release from
the site (without prior decontamination) and release from a decontamination firm after treatment.
  1 For convenience, we use the term "cost" throughout this report.  However, our estimates reflect
the charges faced by individual generators, and hence in some cases (e.g., when commercial services
are used) represent prices rather than costs.

                                           3-1

-------
                                                             Preliminary Draft: June 13, 1997


        Exhibit 3-1 summarizes the unit cost estimates used in the preliminary analysis. As indicated
 by the exhibit, these costs vary greatly. For disposal, DOE facilities generally face lower costs than
 commercial nuclear power plants;  the reported ranges  for both DOE and commercial reactors
 indicate the differences in costs or  charges across individual disposal sites. For unconditional
 clearance, the range depends largely  on the value of the metal and the costs associated with any
 needed decontamination. For example, carbon steel scrap can be sold for about $110 per metric ton,
 while nickel may be worth as much  as $7,300 per ton. However, these values may be offset by
 decontamination costs, which can vary from less than $1,000 per ton for scrap with relatively simple
 geometry to as much as $60,000 per ton for a steam generator.

SUMMARY OF
Disposal:
DOE facilities
Commercial nuclear power plants
Unconditional Clearance:
Direct release without decontamination
Release after decontamination
Exhibit 3-1
PRELIMINARY COST ESTIMATES
(per metric ton)
$2,090-
$4,650 -
($7,150) -
($6,660) -


$3,140
$9,150
$4,990
$59,540
Note: Negative values indicate cases where scrap value exceeds release-related costs.
Key Uncertainties and Plans for Future Research

       This preliminary analysis presents the methodology for assessing the effects of EPA's
rulemaking as well as the data obtained through our initial research. The cost estimates discussed
in this chapter are somewhat uncertain; EPA hopes to refine the estimates in subsequent phases of
the analysis. In the case of disposal, DOE costs are uncertain because facilities generally have not
selected  a disposal site and because the Department is just beginning to develop a consistent
approach to measuring these costs. For commercial nuclear power plants, disposal costs may change
dramatically depending on decisions to open or close individual low-level waste disposal sites or to
alter disposal  taxes or pricing strategies.

       In  the case  of  unconditional * clearance, tti£'most  significant uncertainty  relates to
decontamination costs for release levels that differ significantly from the existing standards. Almost
all  research and experience to date focuses on  meeting existing standards;  we relied on the
professional judgement of decontamination experts to estimate the costs associated with meeting the
alternate standards considered in this preliminary analysis.
                                            3-2

-------
                                                             Preliminary Draft: June 13, 1997


       Because our cost estimates for each scenario may be under- or overstated, these uncertainties
 may lead us to under- or overstate the proportion of the scrap metal that could be released under
 alternate clearance standards. For example, if our estimates of decontamination costs for the 1.0
 mrem option are overstated, and related costs for the 15.0 mrem option are  understated, our
 estimates of scrap metal quantities released under the 15.0 mrem option would be too high relative
 to the 1.0 mrem option. The remainder of this chapter explores the cost estimates and related issues
 in more  detail.
ANALYTIC APPROACH

       Our approach to assessing likely scrap metal management practices and costs included a
literature review, interviews, and development of estimates by selected experts.  We focused on
developing unit costs (i.e., costs per metric ton) for each disposition option considered:  disposal,
unconditional clearance  without prior decontamination, and unconditional clearance with prior
decontamination. We developed separate burial cost estimates for the DOE and NRC facilities and
different sets of decontamination costs for each of the release standards considered.

       The literature review involved an extensive search to identify relevant published reports and
articles, conference presentations, and analyses prepared for individual faculties or agencies. We
then reviewed these documents for quality and applicability to this analysis.

       Because  disposal  options are  uncertain  and  recycling  practices  are  evolving  as
decontamination and decommissioning activities increase, the literature provides limited insights into
current and potential future costs and practices. Therefore, we supplemented the literature review
with extensive  interviews. We spoke with a large number of personnel from individual DOE
facilities, waste management and decontamination firms, and selected commercial power plants over
a one year period to develop a more detailed understanding of current practices, available cost
estimates, and likely future trends.

       Both the  literature  review and the interviews  yielded  little data  on  the sensitivity of
decontamination costs to changes in the release limits. Available sources focus largely on the costs
of meeting the current  standards, which  differ significantly from some of the release levels
considered in this preliminary analysis. To assess the effects of alternate limits, we  worked with
decontamination experts to  develop cost estimates specifically for the release levels considered in
this report.

       The sources used to develop each cost estimate are noted in the subsequent sections of this
chapter. The final section of the-chapter describes the uncertainties in these estimates and plans for
future analysis.                      ''         "       •    ••- '
                                            3-3

-------
                                                              Preliminary Draft: June 13, 1997
DISPOSAL COSTS
       The costs of low level nuclear waste disposal vary depending on the direct costs of burial,
the need for packaging materials prior to disposal, the incentives for volume reduction, and the
distance the material must be transported. For the preliminary analysis, we consider three scenarios,
as illustrated in Exhibit 3-2:  on-site disposal at DOE facilities; off-site disposal for DOE facilities;
and off-site disposal for commercial power plants.

       Because of the substantial uncertainty  related to disposal options for low level wastes, we
developed both high and low end cost estimates for this preliminary analysis. In all cases, we assume
that scrap metal potentially affected by EPA's rule could be disposed as low level waste. In addition,
we do not consider costs likely to be similar  under all  disposition scenarios (both disposal and
unconditional clearance), such as initial demolition, interim storage, and initial (pre-release) survey
costs. While these costs may vary somewhat depending on whether the scrap is disposed or cleared
for unconditional use, the amount of variance  is difficult to estimate and likely to be small.
DOE Disposal Costs

       DOE  facilities have a number of options for disposing scrap metal. Many have on-site
disposal  facilities, and most can send wastes off-site to other DOE or commercial facilities. The
facilities included in this analysis generally have not  determined where  they would dispose their
scrap (if it were  not recycled)  and related  costs are highly uncertain. Therefore, we follow a
relatively simple approach for the preliminary analysis. First, as a low end option, we assume that
scrap,  in most cases, would  be disposed on-site, at  the average annual disposal cost currently
reported by DOE sites for which cost estimates are available. Second, as a high end option,  we
assume that certain facilities would dispose wastes off-site, at a cost equivalent to average off-site
charges.

       To determine these costs, we follow a four step process.  First,  we consider  the costs
associated  with on-site disposal for the 11 DOE facilities included in  the preliminary analysis.
Second, we consider the off-site disposal options for those facilities likely to  consider such options.
Third, we assess the extent to which related costs  provide incentives for volume reduction prior to
disposal. Finally,  we  determine the  total costs  per metric ton for disposal, including  volume
reduction, handling, transportation and other costs.
                                             3-4

-------
                                        Exhibit 3-2
                      OVERVIEW OF DISPOSAL COST ASSUMPTIONS
                                                    Yes
Generate
 Sera.p.;
                                                                = Preparation + On-Site Disposal
= Preparation + Transport + Off-Site Disposal
                        Costs = Preparation + Transport + Off-Site Disposal

-------
                                                              Preliminary Draft:  June 13, 1997
 On-Site Disposal Options and Costs
        The first step in the analysis is to determine the costs of burying scrap metal on-site. Cost
 factors include facility development costs and operating and maintenance costs. We focus on current
 and planned practices for the time period over which scrap  from each facility is likely to become
. available (as discussed in Chapter 2) — between 1998 and 2018 for most facilities (up to 2038 for
 Hartford). Information from our interviews and review of related documents suggests that all eleven
 facilities may  dispose  at  least some  types of low level waste on-site.2  However,  most facility
 managers are uncertain whether on-site units would be used to dispose scrap metal if recycling were
 not possible. In many cases the facilities are storing rather than burying scrap metal, because storage
 is the least expensive short-term option and/or the facility anticipates  eventually recycling the scrap
 in a manner-consistent with DOE policy goals.3 Also, several facilities now limit the types of wastes
 buried on-site due to capacity constraints, costs, risk-related concerns, or other factors. These issues
 make it difficult to predict whether on-site disposal of scrap  is likely  in the future.

        Characterization of on-site disposal costs is further complicated by current DOE practices.
 In cases where low level wastes are disposed on-site, generators are usually not charged a direct fee,
 nor are there  established, generally accepted cost figures for consideration in  scrap disposition
 decisions.4  Our interviews, however, suggest that generators are concerned about the costs of their
 activities even if no direct fee is charged, although they may also consider other factors such as DOE
 policies regarding waste minimization and recycling. The attention paid to costs is likely to increase
 over time because DOE is  pursuing  policies to ensure that faculties take into account the full
 budgetary impact of their  decisions.

        Exhibit 3-3 draws on a recent  DOE-sponsored  study to characterize on-site disposal costs
 for each of the facilities included in this analysis. As the exhibit shows, the unit costs reported vary
 substantially across the facilities. To some extent this variance reflects differences in how the costs
 were calculated,  rather than differences in "true" costs. The estimates are not directly comparable
 because they reflect different accounting practices as well as differences in disposal conditions. In
 addition, in most cases unit costs are determined by dividing 1995 total costs by 1995 quantities
 disposed, and hence will vary depending on the quantity of material buried each year. For example,
 1994 costs differ from the 1995 estimates by an average of almost 20 percent due to variations in
 total  costs and quantities disposed.
  2 DEc interviews of DOE facility personnel, July through November 1996, and U.S. Department
of Energy, The 1996 Baseline Environmental Management Report (BEMR), June 1996.

  3 Under its Recycle 2000 and waste minimization policies, DOE is encouraging facilities to recycle
metals (either for restricted use or through unconditional clearance) rather than dispose them.

  4 Of the 11 facilities included in the preliminary analysis,  two report direct charges  to on-site
generators for disposal:  Hanford, which charges $42.13 - $95.68 per cubic foot for Category 1 and
2 contact-handled low level waste; and LANL, which charges $5.00 per cubic foot. Hanford charges
arc provided in:  Michael  Gresaifi  (Oak Ridge National Laboratory) and Jayne Tellaricp (U.S.
Department of Energy), "Cost of Low Level Waste Disposal - Baseline;  Back-up Information."
August 30, 1995, page 24. LANL charges are from interviews conducted with site managers.

                                             3-6

-------
                                                            Preliminary Draft: June 13, 2997
Exhibit 3-3
COSTS FOR ON-SITE DISPOSAL
(1995 dollars per cubic foot)
Disposal Site
Fernald Environmental Management Project
Hanford Site
Idaho National Engineering Laboratory
Los Alamos National Laboratory
Nevada Test Site
Oak Ridge Facilities
Paducah Gaseous Diffusion Plant
Portsmouth Gaseous Diffusion Plant
Rocky Flats Environmental Technology Site
Savannah River Site
Weldon Spring Site
Costs
unknown
$40.00
$163.00
$45.00
$25.00
$179.00
unknown
unknown
unknown
$34.00
unknown
Source: Michael Gresalfi (Oak Ridge National Laboratory) and Jayne Tellarico
(U.S. Department of Energy), "Cost of Low Level Waste Disposal - Baseline,"
August 30, 1995.
       In addition to the issues described above, two of the facilities report substantially higher costs
than the others:  INEL and Oak Ridge. On-site disposal of scrap metal may be unlikely at both of
these facilities due to cost considerations and other factors. Information  from the  DOE's 1996
Baseline Environmental Management Report (BEMR) and our interviews suggests that INEL may
begin disposing low level wastes off-site in 1998 due to capacity constraints and other concerns. The
same sources indicate that Oak Ridge only disposes selected wastes on-site (mixed fission products
from ORNL) and is not likely to use on-site faculties for other wastes.
                                            ,. • f-
                                           •m  ,
                                            3-7

-------
                                                              Preliminary Draft:  June 13, 1997
        In light of the factors noted above, this preliminary analysis uses a simple average of on-site
 disposal cost estimates for four facilities — Hanford, LANL, NTS, and SRS ~ as a low end estimate
 of DOE disposal costs. This average, $36.00 per cubic foot, should provide a reasonable lower bound
 on the costs of disposing scrap metal at DOE facilities.
 Off-Site Disposal Options and Costs

        Data from our interviews and other sources suggest that six of the 11 facilities included in
 our analysis may consider disposing wastes off-site due to concerns about costs, disposal capacity,
 risks, or other factors. These six facilities include Fernald, INEL, Oak Ridge (ORNL, K-25, and Y-
 12), Paducah, Portsmouth, and Rocky Flats. Our interviews and literature review suggest that these
 facilities will primarily consider three options for off-site disposal:  Hanford;  the Nevada Test Site;
 or the commercial  Envirocare facility in Utah.5  In most cases, the generating facilities have not
 yet determined which (if any) of these sites they might use for scrap metal, and some may send
 wastes to more than one site.  Exhibit 3-4 lists current off-site charges for these three options.
Exhibit 3-4
FEES CHARGED TO OFF-SITE GENERATORS
(1996-1997)
Disposal Site
Hanford Site
Nevada Test Site
Envirocare
Fees
(per cubic foot)
$42.13 - $95.68
$17.00
$50.00
Source: Michael Gresalfi (Oak Ridge National Laboratory) and Jayne Tellarico (U.S.
Department of Energy), "Cost of Low Level Waste Disposal - Baseline," August
30, 1995.
Note: Envirocare charges are highly uncertain and vary greatly depending on the
generator and waste characteristics.
  5 While other DOE sites  accept wastes from off-site for  disposal,  DOE policy and other
constraints limit the disposal sites likely to be considered by each facility.

                                             3-8

-------
                                                              Preliminary Draft: June 13, 1997


        DOE facilities that accept wastes from off-site are also facing increasing pressure to consider
the full lifecycle costs  of disposal in their pricing.  However, we are uncertain  whether these
pressures will lead to significant changes in the charges imposed by Hanford or NTS.  The data on
costs in Exhibit 3-3  above indicate that Hanford's current charges are higher than the unit costs
($40.00 per cubic foot) faced by the site, while NTS charges are about $8.00 per cubic foot less than
its unit costs.6 DOE is in the process of developing lifecycle cost estimates; the extent to which
these Lifecycle costs will differ from the cost estimates cited above is uncertain.

        Because of the uncertainty regarding which sites are likely to be used for scrap disposal, we
use a simple average of the costs reported in Exhibit 3-4 as the high end disposal cost estimate for
the six facilities likely to consider off-site disposal. Although this average, $45.00 per cubic foot, is
relatively close to the average for  on-site disposal, the total cost of off-site disposal will be higher
due to transportation and other factors. These factors are discussed in more detail  below.
Incentives for Volume Reduction

       Because disposal costs are generally calculated based on volume rather than weight, they
provide incentives for processing scrap prior to disposal. Generators face four related choices:  (1)
to dispose their wastes in bulk form, without any processing for volume reduction; (2) to section the
waste, cutting it into smaller pieces; (3) to compact the waste, using pressure to reduce its volume;
or (4) to melt the waste into ingots. In the disposal cost range considered in this analysis ($36.00 to
$45.00 per cubic foot) it is generally cost-effective to section scrap prior to burial. The costs of
sectioning (averaging $330 per ton) are outweighed by the disposal cost savings attributable to the
lower volume/increased density of the waste (e.g., a density of 45 rather than 20  pounds per cubic
foot for bulk waste).7
  6 The available cost data generally include handling charges that are incurred for wastes received
from on-site (e.g., for placement in containers) that may not accrue when wastes are received from
off-site (e.g., because the generating site has already placed the waste in containers).

  7 Density estimates are based primarily on information from Warren, Stephen et al.  Cost Model
for DOE Radioactivelv Contaminated  Carbon  Steel  Recycling.  U.S. Department of Energy.
December 1995, page 15. The sectioning cost estimate is the mid-point of the range reported by
S. Cohen, and Associates, Analysis of theiPotential Recycling of Department.of Energy Radioactive
Scrap Metal. Prepared for the U.S. Environmental Protection Agency, August 14,1995, page 8-10.
Note that these data are for carbon steel, which represents over 80 percent of the metal addressed
by this analysis. The implications of applying carbon steel estimates to other metals are discussed
at the end of this chapter.

                                             3-9

-------
                                                                    Preliminary Draft:  June 13, 1997
 Total Costs

         The final step in evaluating DOE disposal costs is to calculate total disposal costs for each
 scenario. In  the case of on-site disposal, most related handling is  included in the cost numbers
 displayed earlier. For off-site disposal, preparation costs incurred by the generating facility need to
 .be added to  the disposal charges to determine total costs. In addition to volume reduction, these
 costs include handling (the labor associated with packing the materials into containers), the cost of
 the container itself (e.g., a B-25 box), and transportation to the disposal site. Exhibit 3-5 provides
 estimates of the total unit cost for each option.
                                              Exhibit 3-5

                       ESTIMATED DOE PROCESSING AND DISPOSAL COSTS
                                            (per metric ton)
Option
On-Site Disposal (average =
$36.00 per cubic foot)
Off-Site Disposal (average =
$45.00 per cubic foot)
Sectioning
$330
$330
Handling
N/A1
$110
Container
N/Am
$320
Transport
-0-
$170"
Disposal
$1,760
$2,210
Total
$2,090
$3,140
    Sources: See above for calculation of sectioning and disposal costs.
           Handling, container, and transport costs (per ton-mile) are from: (1) S. Cohen and Associates. Analysis
           of the Potential Recycling of Department of Energy Radioactive Scrap  Metal.  Prepared for the U.S.
           Environmental Protection Agency. August 14, 1995, pages 8-10 and 8-63;  and (2) Stephen Warren et al.
           Cost Mode! for DOE Radioactrvelv Contaminated Carbon Steel Recycling. U.S. Department of Energy.
           December 1995, pages  11 - 12. Container costs assume  use of a B-25 box with a capacity of 92 cubic
           feet, packed at a density of 45 pounds per cubic foot. Transportation costs are  estimated at $0.11 per
           ton-mile, assuming the use of a 45-foot truck.

    Notes:  a.      Handling and container costs are generally included in the calculation of DOE on-site disposal
                   costs.
           b.      Transportation costs are based on an average distance of approximately  1,580 miles calculated
                   by determining the distance from each of the six generating sites to each of the three disposal
                   sites.
         Exhibit 3-6 summarizes the disposal cost numbers used in the preliminary analysis. For the
•low end analysis, we use average annual on-site costs for all facilities except two that are unlikely
 to dispose on-site.  For the high end analysis, we use average off-site numbers for the six sites likely
 to dispose off-site, and the on-site numbers for the remaining facilities.
                                                 3-10

-------
                                                             Preliminary Draft:  June 13, 1997
Exhibit 3-6
COST ASSUMPTIONS FOR PRELIMINARY ANALYSIS
(per metric ton)
Generating Facility
Fernaid Environmental Management Project
Hanford Site
Idaho National Engineering Laboratory
Los Alamos National Laboratory
Nevada Test Site
Oak Ridge Facilities
Paducah Gaseous Diffusion Plant
Portsmouth Gaseous Diffusion Plant
Rocky Flats Environmental Technology Site
Savannah River Site
Weldon Spring Site
Low End Analysis
$2,090
$2,090
same as high end
$2,090
$2,090
same as high end
$2,090
$2,090
$2,090
$2,090
$2,090
High End Analysis
$3,140
same as low end
$3,140
same as low end
same as low end
$3,140
$3,140
$3,140
$3,140
same as low end
same as low end
NRC Disposal Costs
     /
       NRC-licensed power plants currently have three options for the disposal of low level waste,
including scrap metal: the Chem-Nuclear site in Barnwell, South Carolina; the U.S. Ecology site
in Richland, Washington; and the Errvirocare site in Clive, Utah. Most power plants currently send
their scrap to Barnwell for disposal. The Richland site accepts wastes only from states in the Rocky
Mountain and Northwest Compacts.8  Envirocare  focuses on  disposal of high volume,  very low
activity waste (such as soils), but will also accept  scrap metal. The disposal  options for commercial
power  plants can change dramatically over time; for example, Barnwell currently will not accept
radioactive waste  originating  in North Carolina,  and was closed to all generators outside of the
Southeast Compact for six months in 1995.9  Planning is underway for the  siting of additional
disposal facilities, but many observers believe that public opposition will prevent new sites  from
opening in the near future.
 *     e    ,                        .....        - . <•               .     ^   i >        .'Iti
  s Two of the 123 commercial nuclear reactors are located in the eleven states included in these
compacts.

  9 The five nuclear power reactors in North Carolina are currently storing waste on-site while the
state challenges the ban in court.
                                           3-11

-------
                                                             Preliminary Draft: June 13, 1997


        Due to the uncertainty regarding the disposal options likely to be available for commercial
 power plants over the period assessed, we use a high-end and low-end estimate of NRC disposal
 costs in the preliminary analysis. The high end is based on the current price structure for Bamwell,
 while the low end is based on Richland prices.10

        Our analysis of NRC disposal costs involves four steps, similar to the approach for calculating
 DOE costs. First, we estimate high-end costs based on data from Barnwell. Second, we estimate low-
 end costs based on Richland data. Third, we consider incentives for volume reduction. Finally, we
 •determine total disposal costs per metric ton, including the costs associated with volume reduction,
 transportation, and other factors.


 High-End Disposal Costs
                                                          *

       Of the three currently operating sites, Bamwell has the highest rates for scrap metal disposal.
 Therefore, we use Bamwell's current price schedule to predict high-end  disposal costs for  the
 preliminary analysis. Chem-Nuclear changed BarnwelTs price  structure  on November 1, 1996 to a
 weight- and dose-based system.  Under this new system, prices are composed  of three elements:  (1)
 a base  price per pound, which decreases as density increases;  (2) a  dose factor, a per pound
 multiplier based on radioactivity levels; and (3)  a curie surcharge based on total millicuries.11

       Based on information on typical scrap  metal characteristics from the reference reactors
 discussed in Chapter 2, discussions with Chem-Nuclear staff, and Bamwell's price schedule, we
 estimate that scrap metal charges at Barnwell could range from about $6,000 per metric ton to over
 $34,000 per ton, depending on density. As discussed later in this section, this price range provides
 significant incentives for volume reduction prior to burial.


 Low-End Disposal Costs

       To develop a low-end estimate of disposal costs for scrap from commercial power reactors,
we reviewed information on charges imposed by the Richland site. Richland's current price schedule
 includes a number of subcomponents, such as  charges for volume, radioactivity levels, and local
 taxes.12  All of these charges are expressed in dollars per cubic foot Using the information on scrap
  10 We were unable to include Envirocare in these estimates because it determines prices on a case-
by-case basis. Information from our interviews suggests that Envirocare's charges to commercial
faculties may be close to the few-end estimates considered in the preliminary analysis.
                *         -           -           "t
  11 Fitch, Michael. "Disposal Cost Calculation." Chem-Nuclear Systems, Incorporated. September
12,1996-.       •    •           -_    •     •-<..•-'••   tiv.t •     .unni". :   -

  12 U.S. Ecology, Incorporated, Washington  Nuclear Center.  "Schedule A:  Disposal Charge."
August 18,1996.

                                            3-12

-------
                                                            Preliminary Draft:  June 13, 1997


 characteristics discussed in Chapter 2 and Richland's  current price schedule, we estimate  that
 disposal of scrap metal at Richland would cost about $74.00 per cubic foot. Like Barnwell, the cost
 of disposal per metric ton at Richland depends upon the density of the material in question.


 Incentives for Volume Reduction

        As noted above, the cost per ton of disposing scrap  metal at either Barnwell or Richland
 depends upon the material's density, providing an incentive for volume reduction. Rather than being
 disposed in bulk,  scrap from commercial power reactors may be sectioned, compacted, or melted
 prior to disposal.

        The prices currently charged by Barnwell provide significant incentives for volume reduction.
 The costs of melting scrap prior to disposal ($2,800 per metric ton)  are far outweighed by the
 reduction in disposal costs due to the increased density of the metal.13 Using Bamwell's current
 pricing structure, for example, and assuming that melting increases density from 20 pounds per cubic
 foot (for bulk waste) to 450 pounds per cubic foot, melting scrap into ingots prior to disposal would
 reduce disposal costs from about $35,000 per metric ton to $6,200 per ton.14  Based on these
 calculations, every one pound per cubic foot increase in the density of scrap disposed at Barnwell
 would reduce unit disposal costs by approximately $67 per metric ton.

       The Richland prices provide less of an incentive for volume  reduction. The most cost-
 effective approach for scrap sent to Richland is to section it at an average cost of $330 per ton,
 increasing density from 20 pounds per cubic foot to 45 pounds per cubic foot.15


 Total Costs

       'Finally, we calculate total disposal costs for each option. In the high-end scenario, we assume
 that charges related to preparing and melting the scrap are  included hi the melting cost estimate;
 transportation costs are based on the weighted average distance from each of the 123 power reactors
 to one of two regional melt facilities, plus the weighted average distance from each of the  melt
 facilities to Barnwell. For the low-end estimate, we assume the generating facility incurs charges for
  13 We base our estimate of melting costs on the midpoint ($1.27 per pound) of the range of costs
($0.95 to $1.60 per pound) presented in S. Cohen and Associates (August 1995).

  14 The density'estimate forfeited'carbon
-------
                                                                   Preliminary Draft:  June 23, 1997
 sectioning, handling, and  providing the waste container;  transportation costs are  based on an
 average distance of 2,228 miles from each of the reactors to Richland. Exhibit 3-7 summarizes the
 total estimated costs for both the high and low disposal cost scenarios.
Option
High End
Estimate
Low End
Estimate
Processing
$2,800"
(melting)
$330
(sectioning)
Handling
N/A"
$110
Container
N/A"
$320
Transport
$140°
$250"
Disposal
$6,210
$3,640
Total
$9,150
$4,650
                                             Exhibit 3-7

                      ESTIMATED NRC PROCESSING AND DISPOSAL COSTS
                                           (per metric ton)
   Sources: See above for calculation of sectioning, melting, and disposal costs.
           Handling, container, and transport costs (per ton-mile) are from: (1) S. Cohen and Associates. Analysis
           of the Potential Recycling of Department of Energy Radioactive Scrap Metal. Prepared for the U.S.
           Environmental Protection Agency. August 14,1995, pages 8-10 and 8-63; and (2) Stephen Warren et al.
           Cost Model for DOE Radioacuvelv Contaminated Carbon Steel Recycling. U.S. Department of Energy.
           December  1995, pages 11 - 12. Container costs for sectioned  metal assume use of a B-25 box with a
           capacity of 92 cubic feet,  packed at a density of 45 pounds per cubic foot Transportation costs are
           estimated at $0.11 per ton-mile, assuming the use of a 45-foot truck.

   Notes:   a.  Based on the midpoint of ranges presented in S. Cohen and Associates, August 1995.
           b.  Preparation costs incurred by the generator are included in the transportation cost estimate;
           preparation costs incurred by the waste management firm are included in the melt cost estimate.
           c.  Transportation costs for the high end estimate are based  on the weighted average  distance from
           each of the 123 reactors to one of two hypothetical melt facilities, plus the weighted average distance
           from each of the two regional melt facilities to Barnweil.
           d.  Transportation  costs for the  low end estimate are based on  the average distance of 2,228 miles from
           the 123 power reactors to Richland.
For the preliminary analysis, we consider the effects of both the high and low disposal costs above
on decisions to dispose or release scrap metal  from commercial power plants.


UNCONDITIONAL CLEARANCE COSTS

        The costs associated with unconditional clearance of sopp metal from nuclear facilities
include those associated with surveying the scrap to ensure that it meets applicable release levels and
the costs of any needed decontamination. These costs are offset at least in part by the sales value
of the scrap. As illustrated in Exhibit 3-8, the preliminary analysis, considers ^two scenarios; (1) sale
to a scrap dealer directly from the generating site after surveying, for metals that meet the clearance
                                                3-14

-------
                                         Exhibit 3-8
             OVERVIEW OF UNCONDITIONAL CLEARANCE COST ASSUMPTIONS
   Survey
    Scrap
Does Scrap
  Meef
  Release
 Criteria?
                                                         Release from
                                                             Site
                                              Costs = Survey - Scrap Value
                                                         Release aftee
                                                        Decontamination
                                              Costs = Transport + Decontamination +
                                                     Survey - Scrap Value
* Decontamination costs include preparation and secondary waste disposal.

-------
                                                             Preliminary Draft:  June 13,1997


 standards without prior decontamination; and (2) sale to a scrap dealer from a decontamination
 firm, for metals that require  decontamination to meet  the  clearance standards.  Other  -:crap
 management options may also be employed, such as sending the scrap to a  decontamination firm
 for survey and release rather than conducting the survey on-site. The two scenarios analyzed,
 however, capture the major cost factors likely to be associated with unconditional clearance.

        The following discussion describes the data we employ to characterize the costs associated
 with unconditional clearance. As was the case with disposal, we exclude costs that are not likely to
 vary significantly across disposition scenarios, such as those associated with  initial demolition and
 interim storage.


 Scenario 1; Survey and Release from Generating Site

        Under the first unconditional clearance scenario,  a generator surveys scrap and determines
 that it meets the clearance standards,  then sells it to a  scrap dealer. Relevant costs include two
 major components:  the cost of the survey and the sales value of the scrap.


 Survey Costs

       To determine the costs of surveying metals for unconditional clearance from the generating
 site, we considered the types of survey techniques  and equipment most likely to be used at DOE
 facilities and commercial nuclear power plants.16  We assessed the costs of detecting surface
 radioactivity at  the  release limits considered  in  the preliminary analysis for four of the five
 radionuclides included in the database,  based on assumptions regarding the typical physical form of
 the scrap likely to be affected by EPA's rule.17  We also assume that sites will continue to follow
 current measurement practices once EPA's rule is promulgated. Exhibit 3-9 presents the resulting
 estimates.
  16  The research reported in this section was  conducted by Ralph  Kenning of S. Cohen and
Associates, November 1996.  The assumptions used in this analysis are based on current  NRC
guidance, including: (1) Gogolak, C.V., A.M. Huffert, and G.E. Powers.  A Proposed Nonparametric
Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys.
NUREG/CR -1505.  U.S. Nuclear Regulatory Commission.  August 1995; and (2) Huffert, A.M.,
E.W. Abelquist, and W.S. Brown.  Minimum Detectable Concentrations with Typical Radiation
Survey Instruments for Various Contaminants and Field Conditions.  NUREG/CR -1507.  U.S.
Nuclear Regulatory Commission. August 1995.     ,    , ,          *

  17 The radionuclides addressed in the analysis of survey costs include Co-60, Cs-137, and Ru-106
for commercial nuclear power plants, and U-238 and Cs-137 for the eleven DOE facilities (see
Chapter 2 for more information on related assumptions).  Pu-239 was not explicitly considered in
the analysis, but related surveying costs are likely equivalent to those for U-238.  We assume that
materials  exhibit only surface contamination at  the point of  generation; data on volumetric
contamination levels are not available for the small quantity of scrap reported as ingots. The release
levels for each radionuclide under each of the options assessed are provided in Appendix A.

                                            3-16

-------
                                                             Preliminary Draft:  June 13, 1997
Exhibit 3-9
ESTIMATED SURVEY COSTS FOR SURFACE CONTAMINATION
(per metric ton)
Release Limits'
Current standards
0.1 mrem standard
1.0 mrem standard
15.0 mrem standard
Cost
' $180
$450 - $5,100"
$180 - 440"
$150
Source: Ralph Kenning, S. Cohen and Associates, November 1996.
Notes: a. See Appendix A for the release limits (expressed as activity levels) under each option.
b. The low end of the range addresses the release limits for the radionuciides assumed
to be present at DOE facilities; the high end addresses the limits for the radionuciides
assumed present at commercial nuclear power plants.
       As indicated by the exhibit, the initial estimates of survey costs do not vary significantly
under current release standards or the 15.0 mrem standard.18 The variance under the 0.1 mrem
and 1.0 mrem standards is due primarily to the effects of the release levels for Co-60. The high end
costs  ($440 and $5,100) reflect the relatively low release limits (90 dpm/100 cm2 and 900 dpm/100
cm2) assessed for this radionuclide. Because we assume that Co-60 is not present in scrap metal at
the DOE facilities addressed by the preliminary analysis, the low end of the survey cost range is used
for DOE facilities and the high end is used for the commercial power plants.


Scrap Market Value

       The second major component of the costs of releasing scrap directly from the generating site
is  the value of the metal itself; i.e., the price  paid  to the generator by the scrap dealer.19 We
estimate these values based on information provided by experts from the decontamination industry,
as illustrated in Exhibit 3-10.
operating conditions across sites).
  19  We assume that on-site preparation or packaging costs (if any) are negligible in cases where
scrap is released directly from the site, and that transportation costs are borne by the scrap dealer.

                                            3-17

-------
                                                             Preliminary Draft:  June 13, 1997
Exhibit 3-10
ESTIMATED SCRAP VALUES
(per metric ton)
Metal
Carbon Steel
Stainless Steel
Copper
Aluminum
Nickel
Lead
Price Estimate
$110
$660
$2,000
$900
$7300
$500
Source: Dan Burns, James Taylor, and Michael Byrne, November 1996.
Notes: The small quantities of other metals included in the database are valued as
follows: miscellaneous metals and iron are valued at the carbon steel price;
monel, brass, and bronze are valued at the copper price; inconel is valued at
the stainless steel price; graphite is valued at zero. No value is assigned to
silver because, as discussed later, silver is not likely to be released.
As indicated by the exhibit, the majority of scrap in the database (i.e., carbon steel) has a relatively
low market value, while the small quantities of nickel are highly valued.20
Scenario 2; Release after Decontamination

       The second unconditional release scenario we consider is release after decontamination by
a commercial firm. Related costs include transportation to the decontamination firm, preparation,
decontamination,  survey, disposal of secondary waste and reject material, and  scrap  sale. To
determine these costs,  we worked with three decontamination experts.21   The  experts  first
categorized the scrap in the database (described in Chapter 2) according to characteristics affecting
decontamination costs, then determined the decontamination processes likely to be used for  each
category.
  20 The prices' in this table represent the current value for typicah grades1' of unprepared scrap sold
by decontamination firms :and hence are likely' to be lower tnan the prices received for prepared
scrap1'sold by a scrap dealer to a mill.

  21 These experts included Dan Bums of Trinity Environmental Systems, James Taylor of Alaron
Corporation, and Michael Byrne of Manufacturing Sciences Corporation.
                                            3-18

-------
                                                             Preliminary Draft: June 23, 1997


        The experts focused on proven technologies currently used for decontaminating radioactive
 scrap metal, including:

        •      Mechanical methods (e.g. grit  blasting), which physically remove surface
               contamination.

        •      Chemical methods,  which dissolve  the  outer layers of metal  that exhibit
               surface contamination.

        •      Thermal methods, which cause certain radionuclides to partition out of the
               metal into the by-products of the melting process (e.g, gas or  slag). These
               methods may be used for materials that exhibit either surface or volumetric
               contamination.

 The experts  then estimated the  costs associated  with preparing  the scrap  for processing,
 decontaminating it, surveying it to ensure it meets the release standards, disposing secondary waste
 and reject  material, and  selling  the  scrap. We  also  developed an estimate  of  the cost  of
 transportation to a decontamination firm.22
Decontamination Costs

       The resulting decontamination cost estimates for each major scrap category and each option
assessed are provided below in Exhibit 3-11.- As indicated by the exhibit, costs can vary greatly within
each category under each set of release standards.
  22 Transportation costs are from Stephen Warren et  al.   Cost Model for DOE Radioactively
Contaminated Carbon Steel Recycling.  U.S. Department of Energy. December 1995, pages 11 -
12.  These costs are estimated at $55 per metric ton, assuming costs of $0.11 per ton-mile (for use
of a  45-foot  truck)  and  assuming that  regional  processors  will  open hi  areas  on average
approximately 500 miles from each generating site.

                                           3-19

-------
                                                                          Preliminary Draft:  June 13, 1997
Exhibit 3-11
ESTIMATED COSTS FOR DECONTAMINATION AND RELEASE*
(per metric ton)
Scrap Category
Structural Steal:
- Good steel (c.g., heavy gage, flat surfaces)
- Bad steel (e.g., light gage, hard-to-access surfaces)
-Tanks
'rocess Systems:
- Large piping systems
- Small piping systems
.arge-Cgropgnents:
- Steam generators
• Turbine rotors
- Heat exchangers
• Fuel racks
• Diffusion cells and related materials
taeciaj Materials:
• Lead shielding and sheets
• Steel shield casks
• Wide copper wire
• Volumetrically-contaminated copper (e.g., narrow wire, ingots)
• Flat aluminum
• Volumetrically-contaminated aluminum (e.g., compressor blades, ingots)
• Volumetrically-contaminated nickel (e.g., ingots)
crap Metal:
• Mixed scrap piles
• Scrap equipment (e.g., small components such as pumps and motors)
11 categories
Current
standards
51,810
52,490
52360
$3,100
56380
$19,850
51,720
54300
54,480
54370
$4,230
55,840
51,060
N/A"
53,290
N/A"
N/A"
53,000
54,260
51,060-
519,850
Analytic Options
0.1 mrem
standard
54,100
56,090
$5,540
59,990
513,650
$59440
53430
58,930
512,130
59,900
511,250
513,100
54,230
$12450
$10,800
$14,420
$10,650
58,070
$9,770
$3430 -
559440
1.0 mrem
standard
$1,940
52,600
52410
S2,960d
55,270"
522,050
51,830
54,400
$4,940
$3,790"
$4,630
$6,170
$1,120
$70
$3,150"
58430
($3,110)c
52,930"
$2,690"
($3,110) -
$22,050
15.0 mrem
standard
51400
51420
$1,980
S2.120
54,150
$15,440
$1,240
$3,420
$3430
$3,480
$3490
$4,900
5840
($90)c
$2440
$6,770
($6,660)c
$2,230
52^10
($6,660) -
515,440
ource: Dan Bums, James Taylor, and Michael Byrne, November 1996.
less for strong beta/gamma emitters. See Appendix A for information on release levels considered and Chapter 2 for information
on scrap characteristics assessed. Exhibit excludes transportation costs, but includes costs for disposal of secondary waste.
b. Materials in this category would require melting for effective decontamination. Due to the absence of a volumetric clearance
limit, they could not be released under current standards without special approval.
c. Negative values indicate cases where the market value of the scrap metal exceeds the costs of decontamination (for certain
copper and nickel components).
d. Costs are lower relative to current standards due to the establishment of release standards for volumetrically contaminated
scrap. In most cases, the least-cost decontamination method for the scrap metal in this category will include straight metal melt
or some combination of melt and surface  decontamination.

-------
                                                             Preliminary Draft: June 13, 1997


       As indicated  by the  exhibit,  the  costs associated  with unconditional clearance after
decontamination vary significantly,  depending on  metal characteristics and  the release levels
considered. The variance  in the total  costs  reflects variation in the costs of the  different
subcomponents of the process (preparation, decontamination, survey, and secondary waste and reject
disposal). The cost of each of these steps can range from less than $500 per ton to $2,000 or
significantly higher, depending upon the physical form of the scrap and the release standard.

       Not surprisingly, costs decrease as the release limits increase. The costs of releasing materials
under current standards are closest to the costs for the 1.0  mrem option, consistent with  the
information on the release limits provided in Appendix A. The  highest costs are associated with
certain large components (such as steam generators), followed by some specialty metals. The lowest
costs  accrue in cases where the resulting  scrap has high market value (i.e., copper and nickel) or
where the metal  has relatively  simple  geometry  (such as flat structural steel,  simple large
components such as turbine rotors, and copper wire).


Adjustments for High Initial Activity Levels

       The costs reported in the previous section focus on scrap with "typical" initial activity levels;
i.e.,  25,000 dpm/100  cm2 or less for strong alpha  emitters and  250,000 dpm/100 cm2 or less for
strong beta/gamma emitters. Decontamination firms generally  impose a surcharge for scrap with
higher initial activity  levels  due to concerns about worker protection and other factors.  In  this
analysis, we assume that decontamination costs increase by 15 percent, both for alpha, emitters with
starting activity levels greater than 25,000 dpm/100 cm2 and for beta/gamma emitters with starting
activity levels greater than 250,000 dpm/100 cm2.23  For example,  scrap exhibiting strong alpha
activity that would ordinarily cost $1,000 per ton to decontaminate would instead cost $1,150 per ton
to decontaminate if the initial activity level were 50,000 dpm/100 cm2.24
RESTRICTED RECYCLING

       Restricted recycling refers to the fabrication of scrap metal into products for reuse in nuclear
settings, such as shielding blocks and waste containers. Such recycling may be affected by EPA's rule
if the rule makes it more or less cost-effective to release scrap that would  otherwise have  been
  23 Surcharge estimates provided by Dan Burns, James Taylor, and Michael Byrne, January 1997.

  24 Decontamination costs would increase by an additional 10 percent in instances where gamma
doses exceed 100 mR/hour, and by a further 3Q percent where gamma doses exceed 1,000 mR/hour,
however, based on conversion factors providedfcy S. Cohen and Associates, Inc., the surface activity
levels required to produce these doses range from a low of 100,000,000 dpm/cm2 (for Co-60) to a
high of 60,000,000,000,000 dpm/cm2 (for Pu-239).  These activity levels exceed the highest levels
reported for the scrap considered in this analysis.  As a result, the secondary surcharges for the
decontamination of strong gamma emitters do not affect our analysis.

                                           3-21

-------
                                                             Preliminary Draft:  June 13, 1997


 subject to restricted recycling. To date, we have not quantitatively assessed the potential impacts of
 EPA's rule on restricted recycling practices. The following discussion provides some background
 information on this issue. In Chapter 4, we discuss in more detail the effects that consideration of
 restricted recycling could have on our analytic results.

        Relatively little information is available  on restricted recycling of NRC scrap.  While
 anecdotal  evidence suggests that commercial power plants occasionally send materials  to waste
 management  firms for restricted  recycling, more research is needed to determine the typical
 characteristics of affected metals, the Scope of these efforts, and related costs. We are also uncertain
 about the extent to which these efforts may change as more power plants end routine operations and
 enter the decontamination and decommissioning phase.

       More  information is available on restricted recycling programs for materials from DOE
 facilities. DOE recently promulgated a policy of aggressively pursuing opportunities for restricted
 recycling  under  a program referred  to as "Recycle 2000.ll2S  This program  focuses  on the
 manufacture of single use  waste  disposal  containers from contaminated carbon steel.26  DOE
 recommends restricted recycling of carbon steel with activity levels less than 100 times the surface
 contamination limits  in DOE Order 5400.5, excluding any material that can be cost-effectively
 released for unconditional use. Thus, under the current standards, carbon steel would be subject to
 restricted recycling if the activity level criteria were met and such recycling were less expensive than
 free release.

       DOE  has  developed a  cost  model to evaluate the production of B-25 low level waste
 containers  from carbon steel, and is currently considering an extension of the model  to address
 stainless steel and possibly  other metals.27    Under the base case assumptions  used in  this cost
 model, recycling carbon steel to fabricate waste containers would cost about $30 per cubic foot ~
 slightly less than the DOE disposal costs noted earlier. Other studies report a broad range of costs
 for melting scrap and  fabricating containers — e.g., from less than  $1.00 per pound to close to $5.00
  25  Memorandum from Alvin Aim, Assistant Secretary for Environmental Management.  "Policy
on Recycling Radioactivety Contaminated Carbon Steel." U.S. Department of Energy. September
20,1996.

  26 DOE is currently focusing on manufacture of M-100 containers.  See Memorandum from James
M. Owendoff, Deputy Assistant Secretary for Environmental Restoration, and Stephen P. Cowan,
Deputy Assistant Secretary, Office of Waste Management.  "Use of Standardized .Low-Level Waste
Containers."  U.S. •Department of Energy.  April 17,1996.'  "   "     "  '       ,.

  27 Warren*.- Stephen et  al.  Cost Model for  DDE  Radioactively Contaminated Carbon  Steel
Recycling.  U.S. Department of Energy.   December 1995.  This  study  assumes that the DOE
restricted recycling program may involve use of 50,000 tons of carbon steel per year over the next
five years to manufacture waste containers.

                                           3-22

-------
      •  •                                                     Preliminary Draft:  June 13, 1997


 per pound.28 Whether the costs associated with DOE's particular restricted recycling initiative will
 be less than the costs of unconditional clearance is unclear, both because the estimates of restricted
 recycling costs are uncertain and because the costs of unconditional clearance depend heavily upon
 the scrap's characteristics and the release standard applied.

       More research is needed to determine the appropriate restricted recycling cpst estimates for
 the DOE and commercial power plant scrap metal categories used in our analysis. In addition, we
 are uncertain about possible changes in DOE's policy on activity levels, should EPA's  standards
 replace the  existing DOE Order 5400.5  levels.  For  these  reasons,  our preliminary analysis
 incorporates only a qualitative assessment of restricted recycling's implications for the management
 of NRC and DOE scrap. We may conduct additional research on this topic and attempt to include
 a  more detailed assessment of the restricted recycling option in future analyses of EPA's rule.


 KEY UNCERTAINTIES AND PLANS FOR FUTURE ANALYSIS

       The preliminary cost  estimates described above are subject to uncertainties that may lead
 us to over- or understate the quantities of scrap released under alternate clearance standards. Over
 the next year, EPA may research related issues iamore detail. Key observations concerning analytic
 uncertainties are described below.

       •      Disposal costs are uncertain:  While the use of high and low estimates
              attempts to assess the effects of uncertainty in disposal costs, actual costs for
              particular facilities may vary significantly from the range of costs analyzed.
              Policy decisions made by DOE sites, commercial waste  management firms,
              and the states where commercial  sites  are  located or may  be sited are
              difficult to  predict and can  lead to changes in charges, taxes, and waste
              acceptance criteria. Actual costs also can vary substantially  by site and have
              not yet been calculated on a lifecycle basis for DOE facilities. In addition,
              facilities generally have not yet determined which sites they will use for scrap
              disposal. Overstatement of these costs may lead us to overstate the incentives
              for releasing scrap; understatement will have the opposite effect.

       •      Little information is available on decontamination costs for alternate release
              standards:  The costs of decontaminating metals for unconditional clearance
              are uncertain • for those  options that establish  release limits  that differ
              significantly from the current standards. Due to the lack of data on the costs
              associated with meeting these standards, we rely on expert judgement  to
     i~ ,.  ^         I    ' •').—•• '•••«• .•!.,   ; ,,    ,-j; i "J - '   1 -ilPC'1 i^'  IP  -\l<  '1 <•,!  / .   -  v
    »'                      '       '     -'<- j-v  •*-•*,    '     1  °'a'i. 1 J ' '           __
  28 Trinity Environmental Systems. "White Paper Issues Discussion and Recommended Resolution
of Commingling, Production Cost, Mixed Waste,  Throughput Assumptions  and Background for
Recycle 2000 Option 3 (rev. 2)." November 29, 1995.  This paper recommends the use of $1.65 per
pound ($3,640 per metric  ton) as the best estimate of melting, rolling, and fabrication  costs for
DOE's initiative.

                                            3-23

-------
                                                              Preliminary Draft:  June 13, 2997


               estimate these costs and are unable to quantify  related uncertainties.  In
               addition, increases in the availability of scrap from decontamination and
               decommissioning activities or higher release standards may ultimately bad
               to  capacity  increases  and  economies  of  scale  that  could  reduce
               decontamination costs.

        *      Direct release costs may be subject to  less uncertainty:  While the use  of
               simplifying assumptions  leads  to some  uncertainty  in the  survey  cost
               estimates,  these costs are generally better understood than the disposal and
               decontamination costs discussed above. Variation in the physical form of the
               metal and site operating procedures may lead survey costs to vary by about
               a factor of two; the variation in disposal costs reported earlier can be much
               larger.29   Scrap  prices are  also  very  volatile, but generally  exhibit less
               uncertainty than other cost estimates.30  Uncertainty in these estimates may
               lead us to  either under- or overstate quantities released.

        *      More research is needed to assess the  implications of restricted recycling:
               Because restricted recycling in some cases may be a lower cost option than
               disposal or free release, excluding this scenario may lead us to overstate the
               amount of scrap disposed or released.

       To address these concerns, we use both high end and low end disposal costs in our analysis,
and assess the  effects of the other sources of uncertainty qualitatively. The results of this analysis
are reported in the following chapter and will be used  to determine priorities for future research.
  29 The estimate of uncertainty in the survey costs was provided by Ralph Kenning of S. Cohen and
Associates. One example of the variation in disposal costs is provided in Exhibit 3-3, which indicates
that estimates of, on-site disposal costs for DOE vary by a  factor of seven.

  30 FOP example-,- from -1987 to 1995, carton steel scrap prices (heavy melting $1) ranged from S95
to $151 per ton. Other metals showed more volatility; for example, primary  nickel values ranged
from $3,990 to $13,845 per ton.  (See: American Metal Market.  Metal Statistics. 1988, 1994 and
1996 editions.)
                                            3-24

-------
                                                            Preliminary Draft:  June 13, 1997
CHANGES IN COSTS AND QUANTITIES RECYCLED                          CHAPTER 4
INTRODUCTION AND SUMMARY
                                                      *
       This chapter describes the estimated impact of EPA's rulemaking on quantities of scrap
metal that are recycled or  disposed,, and associated costs.   We  describe the analytic method
employed and present preliminary results for the alternate release standards discussed in the
previous chapters, based on the data on scrap characteristics presented in Chapter 2 and the cost
estimates presented in Chapter 3. We also discuss the major uncertainties and limitations of this
preliminary assessment, as well as EPA's potential plans to refine and expand the analysis over the
next several months.
Summary of Findings

       For the preliminary analysis, we consider disposition options for scrap from 11 major DOE
facilities and 123 commercial nuclear power reactors.  We compare the costs of disposal to the costs
of unconditional clearance to determine the least cost disposition option for each scrap source under
each of the four sets of release limits considered. This standard analytic approach assumes that
decisionmakers are rational and will choose the least expensive disposition option for each piece of
scrap metal. We conduct this comparison twice — once with low-end and once with high-end values -
- to  account for uncertainties in the disposal  cost estimates. Because initial activity levels are
uncertain and can have an important impact on scrap management decisions, we also consider the
effects of assuming that all activity levels are at the high end, the low end, and the logarithmic mid-
point of the ranges reported in the scrap metal data (see Chapter 2).1
 1 * We! use a togarithmic^nud-poial rattier thanr an arithmetic mid-point-in our analysis-because-the
distribution of activity levels for each scrap source is likely to be skewed towards the lower end of
the estimated range. •

                                           4-1

-------
                                                             Preliminary Draft: June 13, 1997
       Exhibit 4-1 presents the results of the preliminary analysis assuming mid-point initial activity
levels.2 As indicated by the exhibit, disposition costs are similar under the current standards and
the 1.0 mrem option. Costs increase by 12 to 23 percent under the 0.1 mrem option because of the
need to meet lower standards to release scrap for unconditional use. These lower standards increase
the costs of decontamination and encourage disposal of a larger proportion of the scrap. Under the
15.0 mrem option, the significant increase in applicable release limits relative to the current standard
reduces scrap disposition costs by more than 80 percent.
Exhibit 4-1
PRELIMINARY ESTIMATES OF
TOTAL SCRAP DISPOSITION COSTS BY OPTION
(present value; 1997 dollars)
Option
Current standards
0.1 mrem
1.0 mrem
15.0 mrem
Disposal Costs
Low Disposal Costs
$1.66 billion
$1.86 billion
$1.66 billion
$0.25 billion
High Disposal Costs
$1.98 billion
$2.44 billion
$1.96 billion
$0.33 billion
Notes: Values are total costs of disposal and unconditional clearance for the
scrap available in the DOE/NRC inventory. See Appendix A for information
on the release limits assessed, Chapter 2 for information on scrap
characteristics, and Chapter 3 for information on cost estimates. Results
assume that initial activity levels are at the logarithmic mid-point of the
reported ranges.
Totals in this table may not equal the sum of the agency tables due to
rounding.
 • 2 All costs are-present values (in 1997 dollars) calculated using a real seven percent discount rate,
as recommended by the U.S. Office of Management and Budget (OMB).  This rate represents the
recent marginal pretax rate  of return on an  average private investment.  See:  U.S. Office of
Management and Budget.  Guidelines and Discount Rates for Benefit-Cost Analysis and Federal
Programs (Circular A-94V  October 29, 1992. We may assess the effects of using alternate rates
in future stages of this analysis.

                                            4-2

-------
                                                             Preliminary Draft: June 13, 1997
       The change in costs across options reflects the change in the quantities that can be released
for unconditional use, as illustrated in Exhibit 4-2. Again, the effects of the 1.0 mrem option are
similar to the current standards, while the 0.1 mrem and 15.0 mrem options lead to large changes
in the quantities released.  Under the current standards and the 0.1 mrem and 1.0 mrem options,
most of the scrap released must first be decontaminated. Under the 15.0 mrem option, most can
be released without decontamination, contributing significantly to the cost savings illustrated in the
prior exhibit.
Exhibit 4-2
PRELIMINARY ESTIMATES OF
SCRAP QUANTITY RELEASED BY OPTION
(metric tons)
Option
Current standards
0.1 mrem
1.0 mrem
15.0 mrem
Disposal Costs
JLow Disposal Costs
0.45 million tons
(29 percent)
0.07 million tons
(4 percent)
0.47 million tons
(30 percent)
1.39 million tons
(88 percent)
High Disposal Costs
0.56 million tons
(36 percent)
0.20 million tons
(13 percent)
0.66 million tons
(42 percent)
1.47 million tons
(93 percent)
Notes: Quantities would be released over a 55-year period from 1998 to 2053
(see Chapter 2 for detailed discussion). Percentages are based on the
approximately 1.6 million tons of scrap available in the DOE/NRC inventory.
Results assume that initial activity levels are at the logarithmic mid-point of the
reported ranges.
The exhibits also  demonstrate the effects of changing the estimates of disposal  costs.  Higher
disposal costs make unconditional clearance more attractive (because decontamination and release
becomes less expensive in comparison), increasing the quantities cleared for unconditional use.
However, the increase in disposal costs also causes total costs to increase.

       The costs reported in Exhibit 4-1 are driven primarily by the disposition of scrap from the
11 DOE facilities1 assessed.7 For example;, T)GE costs 'account'fi7r7#.feercferijtitbf-thci total"-'$1.55
biflion df ? 1.98 billion - utfcter the current' standards {high disjWsafMst sbeftaritf^This is In'part
because DOfc accounts'for a larger share  of affected scrap than "the commercial power reactors
(936,300 tons vs. 641,000 tons), and in part because almost hah* of the DOE scrap is expected to be
                                            4-3

-------
                                                             Preliminary Draft:  June 13, 1997


 disposed or released before the year 2000, while most of the scrap from the 123 power reactors
 assessed will not become available until after the year 2020.  As a result, the disposition of DOE
 scrap has a greater impact on the present value of scrap management costs. On an undiscounted
 basis, the costs of managing DOE and commercial nuclear power reactor scrap are similar.  For
 example, the total undiscounted costs under the current standards (high disposal cost scenario) are
 estimated to be $2.7 billion for DOE scrap and $2.8 billion for commercial  power reactor scrap.

       Although DOE accounts for the majority of the scrap in the inventory evaluated, our analysis
 indicates that commercial nuclear power reactors are likely to account for most of the scrap cleared
 for unconditional use.  For example, 84 percent of the  scrap that we estimate would be released
 (0.47 million of the 0.56 million tons) under the current standards  (high disposal cost scenario) is
 from commercial power reactors.  This results in part because the  power reactors face higher
 disposal costs than the DOE facilities, making  decontamination and release relatively more cost-
 effective.

       It is important to note that these preliminary results are based on limited data  on the
 radiological characteristics of the affected materials. The analysis addresses only five radionuclides,
 for which  activity levels  are reported as  ranges that often span several  orders of magnitude.
 Conducting the analysis using the low or high ends of the ranges (rather than the mid-points as in
 the exhibits above)  leads to substantial changes in the estimated effects  of the options.3  For
 example, estimates of total  disposition costs under the current standards range from $0.4 billion
 (using the low end of the reported activity level ranges) to $2.1 billion at the high end.4 Quantities
 cleared for unconditional use range from 1.4 million tons assuming low activity levels to 0.5 million
 tons at high activity  levels.  However, the relative ranking of the options is similar in most cases
 regardless of the activity level assumptions -- costs under the 1.0 mrem option and current standards
 are similar, with a relatively small increase under the 0.1 mrem option and a large decrease under
 the 15 mrem option.
Key Uncertainties and Plans for Future Research
                         «
       This preliminary analysis presents the methodology for assessing the effects of EPA's
rulemaking as well as the results of our initial research.  As discussed above, the estimates of costs
and quantities released are sensitive to uncertainties in the available data. Key among these are the
lack of data on radiological characteristics and other uncertainties related to  scrap characteristics
and disposition options. Uncertainties regarding management costs also limit the precision of our
analysis.
  3 As is discussed later in this chapter, these results primarily reflect the findings for DOE facilities;
the results for the commercial power reactors are relatively similar regardless of whether high or low
estimates of activity levels are used.,  t                               .

  4 This comparison is based on high-end disposal costs.

                                            4-4

-------
                                                             Preliminary Draft:  June 13, 1997


       In general, the preliminary analysis is likely to provide useful information on the relative
ranking of the options but should not be interpreted as providing precise estimates of related costs
or quantities released. We hope to use the results of the preliminary analysis to identify the areas
in need  of  additional research, and to subsequently  refine the analysis to reduce the level  of
uncertainty  in the estimates.
ANALYTIC APPROACH

       To predict changes in costs and in quantities cleared for unconditional use, we developed
an economic model that determines the least-cost approach  for the disposition of scrap under
alternate release standards.  This model predicts the maximum quantity of scrap  that could be
released under each set of standards, assuming that decision-makers always act to minimize their
costs.

       As illustrated in Exhibit 4-3, the major inputs to the model include the scrap characterization
data discussed in Chapter 2 and the cost estimates described in Chapter 3.  Specifically, the model
develops costs for disposal  and unconditional clearance  for each scrap stream  based on the
characteristics of the stream (e.g., metal type, physical form, weight, radiological profile, and source).
The present value of these costs is calculated based on the estimated year the scrap metal quantities
became available. Total disposal costs are based on the weight and the source of the scrap metal
and the per unit burial costs presented in Chapter 3.5  Costs for unconditional clearance are based
on the weight of a particular scrap item, its radiological profile,  its physical form, and the release
standard being analyzed.6
    •s
       The model repeats this process to calculate the costs associated with each disposition option
(disposal or unconditional clearance) for each scrap stream under each set of release standards.  The
model next compares  these costs to determine the least-cost approach under each set of standards,
and summarizes the results.  The process followed by the model is illustrated in Exhibit 4-4.7
    5 For example, the total burial cost for a 0.2 metric ton sheet of carbon steel from Oak Ridge
would equal $628 (0.2 x $3,140), including relevant sectioning and transportation costs.


    6. For example, unconditional clearance costs for a large piping system of carbon steel weighing
one metric ton and re^uirmg,decoptammation-(i.e^.having,Iltvpicari initia^ activity levels) to meet
current release  standards would  equal  $3,045,  including  transportation  costs  ($55)  plus
decontamination and survey costs  ($3,100) minus scrap values ($110).
          I              '              '         ,     • " 5  --     •    I"        1
  7 As noted  previously, this preliminary analysis does not  account for the potential  impact of
restricted recycling activities within the nuclear complex.

                                             4-5

-------
                                                          Exhibit 4-3
                                              COST MODEL COMPONENTS
     Input Tables
Scrap Data Set
• DOE: 11 key facilities
• NRC; 123 power reactors
Cost Data Set
• Disposal: DOE facilities
• Disposal: NRC facilities
» Unconditional Clearance:
  Direct from site
« Unconditional Clearance:
  After decontamination
* Discount rate
Release Limits (by nuclide):
* Current Standards
» Analytic Options
  --0.1 mrem option
  --1.0 mrem option
  — 15.0 mrem option
Disposal and Release Costs
     By Scrap Source
Least-Cost Disposition Option
Calculate, for each scrap source:
• Disposal costs
* Unconditional clearance costs
  — Current standards
  — 0.1 mrem option
  — 1.0 mrem option
  -- 15.0 mrem option
    Calculate, for each scrap
    source, least-cost option
    (dispose or release) under:
     -- Current standards
     -- 0.1 mrem option
     — 1.0 mrem option
     — 15.0 mrem option
 Summary Tables
Calculate total costs,
quantity disposed, and
quantity released under:
 —  Current standards
 —  0.1 mrem option
 —  1.0 mrem option
 —  15.0 mrem option

-------
                                                               Exhibit 4-4
                                                    OVERVIEW OF MODEL LOGIC
   Step 1:
Identify Scrap
Characteristics
                                                 Step 2:
                                                Determine
                                              Disposal Costs
                                  Step 3:
                            Determine whether
                               scrap can be
                              released without
                               \
                              decontamination
   Step4a:
Determine costs
of direct release
   from site
    Step4b:
 Determine costs
 of release after
 decontamination
                                       Step 5:
                                    Identify least-
                                        cost
                                      approach
                                    (disposal or
                                      release)
      Step 6:
 Adjust quantities of
   decontaminated
material to reflect the
    reject rate of
decontamination, and
add reject quantity to
      disposal

-------
                                                              Preliminary Draft:  June 13, 1997
 SCRAP FROM MAJOR DOE FACILITIES

        The preliminary analysis addresses 936,300 metric tons of scrap metal from 11 major DOE
 facilities.  Exhibit 4-5 provides the cost estimates for these facilities under each release standard
 assessed. As indicated by the exhibit, disposition costs are similar under three of the four options
 evaluated. This result reflects the similarity in the disposition of the scrap under these options (see
 Exhibit 4-6). Under the current standards, the 0.1 mrem option, and the 1.0 mrem option, the least-
 cost management alternative for most DOE scrap (about 87 to 100 percent) is disposal.  The
 situation changes under the 15.0 mrem option, where the release levels become high enough that
 unconditional clearance is the least-cost approach for most DOE scrap (98 percent).  The negative
 costs under the 15 mrem option indicate net revenues; i.e., the scrap's sales value is expected to
 exceed the costs associated with surveying and decontamination (if any) prior to  release.  As the
 exhibits show, higher disposal costs under the high end scenario lead to an increase in total costs
 and provide an incentive for releasing larger quantities of scrap.8
Exhibit 4-5
PRELIMINARY ESTIMATES OF
DOE SCRAP DISPOSITION COSTS BY OPTION
(present value; 1997 dollars)
Option
Current standards
0.1 mrem
1.0 mrem
15.0 mrem
Notes: Values are
the scrap available
activity levels are at
Disposal Costs
Low Disposal Costs
$1.34 billion
$1.37 billion
$1.30 billion
($0.03 billion)
High Disposal Costs
$1.55 billion
$1.61 billion
$1.51 billion
($0.02 billion)
total costs of disposal and unconditional clearance for
in the DOE inventory. Results assume that initial
the logarithmic mid-point of the estimated ranges.
  8 This result  occurs because an  increase in disposal costs raises  the  absolute  cost of scrap
disposition for all scrap likely to be disposed under the low disposal cost scenario. For example, if
disposal costs increase from  $100 to $500 per ton, but decontamination and release costs  stay
constant at $700 per ton,' disposal will remain the most cost-effective option, but total disposition
costs will increase by $400 ($500 -100).  If Decontamination and release costs are instead $400 per
ton, release will become cost effective under the nigh cost scenario, but total costs will still increase
by $300 ($400 -100).  This occurs because an increase in disposal costs improves the relative cost-
effectiveness of scrap decontamination and release, but does so by raising the absolute cost of scrap
disposal.
                                             4-8

-------
                                                             Preliminary Draft: June 13, 1997
Exhibit 4-6
PRELIMINARY ESTIMATES OF
DOE SCRAP QUANTITY RELEASED BY OPTION
(metric tons)
Option
Current standards ••
0.1 mrem
1.0 mrem
15.0 mrem
Disposal Costs
Low Disposal Costs
0.05 million tons
(5 percent)
0.0 million tons
(0 percent)
0.08 million tons
(9 percent)
0.92 million tons
(98 percent)
High Disposal Costs
0.09 million tons
(10 percent)
0.0 million tons
(0 percent)
0.11 million tons
(12 percent)
0.92 million tons
(98 percent)
Notes: Quantities would be released over a 40-year period from 1998 to 2038
(see Chapter 2 for a detailed discussion of the timing of releases). Percentages
are based on the approximately 0.94 million tons of scrap available in the DOE
inventory. Results assume that initial activity levels are at the logarithmic mid-
point of the estimated ranges.
       Of the  total amount likely to be cleared for unconditional  use  under both the current
standard and 1.0  mrem option, most  (92 to  95 percent of the total  released) must  first  be
decontaminated.  Under the 15 mrem  option, most (94 percent) can be released without prior
decontamination.   The ability to release the  scrap  as  generated significantly affects the cost
estimates,  because  it means that the  sales value  of  the scrap  is not  offset by  significant
decontamination costs. The only costs associated with release are the costs of surveying the material
to ensure that it meets the release levels.

       As discussed hi Chapter 2, most  of the scrap in the DOE  dataset (over 75 percent) is
generated by three faculties:  the  gaseous diffusion reactors at Oak Ridge (K-25), Paducah, and
Portsmouth.  The extent to which  these three facilities account for the majority of scrap released
varies by option; they represent 37 to 49 percent of the scrap released under the current standard
and 1.0 mrem option, but 79 percent of the scrap released under the 15 mrem standard (no scrap
is released under the 0.1 mrem option). This result largely reflects 1^>e>tioiwhipvOf the estimated
activity levels at each facility to the release iimitsruhcjler each pption; e.g.,' assuming mid-point activity
levels,  scrap can be released without prior decontamination only under the 15 mrem option.
                                            4-9

-------
                                                             Preliminary Draft:  June 13, 1997


       Chapter 2 indicates that most of the scrap in the DOE dataset (88 percent) is carbon steel.
Carbon steel also dominates the metal quantities likely to be cleared for unconditional use under
the alternative release standards assessed.  It represents 69 to 88 percent of the scrap released.
Because the steel is of relatively low value, the revenue from its sale would not entirely offset
decontamination and other processing costs; under the 1.0 mrem standard, for example, we estimate
that the cost associated with the release of carbon steel would be approximately $190 million.  In
contrast, the sale of scrap nickel is expected to generate net revenues of over $25 million under the
1.0 mrem standard, even though nickel represents only six to nine percent of the total metal quantity
released.  This result  demonstrates  that cost impacts are  dependent not only on the available
quantity of scrap, but also on its market value.

       The establishment of a volumetric release standard under the 1.0 mrem option also has an
important impact on  the results.  We estimate that approximately  7,800 tons of volumetrically-
contaminated scrap that would flow to burial under current standards (due to the lack of criteria for
the release of volume tricalry-contaminated material) could be released under the 1.0 mrem standard.
This increase in scrap quantities cleared for unconditional use  also accounts for over 60 percent of
the decline in total disposition costs under the 1.0 mrem option.9

       The results of the analysis reported above hi part  reflect the limited  amount of information
available on the radiological characteristics of DOE scrap. Using the low or high end of the activity
level ranges for each DOE facility  (as reported in Chapter 2)  significantly affects the results.  For
example, estimates of total disposition costs under  the current standards range from $0.19 billion
(using the low end of the reported activity level ranges) to $1.56 billion at the high end.10  Similarly,
the quantity that we estimate would be cleared for unconditional use under the current standards
ranges from 0.08 million tons assuming  high-end  activity to  0.9 million  tons assuming low-end
activity.

       The relative differences between the options change somewhat under the different scenarios.
Under the low activity level scenario, most  DOE  scrap (about 98 percent) can be released for
unconditional use under the current standards, the 1.0 mrem option and the 15.0 mrem option; most
(about 100 percent) goes to disposal only under the 0.1 mrem option. Under the high activity level,
scenario, almost all scrap is disposed  under all but the 15.0 mrem option ~ in which case the
proportion disposed drops slightly, to about 80 percent of the  total.
  9 This comparison is based on high-end disposal costs and assumes* initial activity levels at the
logarithmic mid-point of the estimated ranges.

  10 This comparison is based on high-end disposal costs.

                                           4-10

-------
                                                            Preliminary Draft:  June 13, 1997
SCRAP FROM COMMERCIAL NUCLEAR POWER REACTORS

       For NRC licensees, the initial analysis considers 641,000 metric tons of scrap metal from 123
commercial nuclear power reactors. Exhibit 4-7 provides preliminary scrap disposition cost estimates
for these facilities under each set of release limits. As indicated by the exhibit, disposition costs are
similar under most of the options assessed; only the 0.1 mrem option under the high-end disposal
cost scenario leads to markedly higher costs. These results in part reflect the projected disposition
of the scrap, as illustrated in Exhibit 4-8.  As was the case with  DOE facilities, the increase in
disposal costs under the high disposal cost scenario leads to an increase in total costs and provides
an incentive for releasing larger quantities of scrap.
Exhibit 4-7
PRELIMINARY ESTIMATES OF
COMMERCIAL NUCLEAR POWER Reactor
SCRAP DISPOSITION COSTS BY OPTION
(present value; 1997 dollars)
Option
Current standards
0.1 mrem
1.0 mrem
15.0 mrem
Disposal Costs
Low Disposal Costs
$0.32 billion
$0.48 billion
$0.36 billion
$0.28 billion
Notes: Values are total costs of disposal and unconditional clear
in the commercial nuclear power reactor inventory. Results assu
are at the logarithmic mid-point of the estimated ranges.
High Disposal Costs
$0.43 billion
$0.83 billion
$0.46 billion
x $0 35 billion
ance for the scrap available
me that initial activity levels
       The percentage of scrap that must be decontaminated prior to clearance for unconditional
use varies by option. Under the current standards, 61 to 67 percent of the total released must first
be decontaminated; the remainder can be released as generated.11  For the 0.1 mrem option, the
proportion of scrap needing decontamination is 100 percent; none of the affected scrap from the
  11  The high  and low  disposal cost  scenarios shift quantities from disposal to release after
decontamination; the quantity that can be released without decontamination is constant as long as
the release standards do not change. For example, assume that under release standard "A", totals
of 1,000 and 1,500 tons could be released under the low and  high disposal cost scenarios,
respectively.  The 500 ton difference reflects scrap that would be disposed under one cost scenario
but released after decontamination under the other;  there would be no change in the amount (e.g.,
1,000 tons) that could be released without prior decontamination.
                                           4-11

-------
                                                             Preliminary Draft:  June 13, 1997
commercial power reactors is estimated to meet the release limits as generated.  For the 1.0 mrem
option, 80 to 86 percent of released scrap requires prior decontamination.  About 68 to 72 percent
of released scrap requires prior decontamination at the 15.0 mrem standard.
Exhibit 4-8
PRELIMINARY ESTIMATES OF
COMMERCIAL NUCLEAR POWER Reactor SCRAP
QUANTITY RELEASED BY OPTION
(metric tons)
Option
Current standards
0.1 mrem
1.0 mrem
15.0 mrem
Disposal Costs
Low Disposal Costs
0.40 million tons
(62 percent)
0.07 million tons
(11 percent)
0.39 million tons
(61 percent)
0.47 million tons
(73 percent)
High Disposal Costs
0.47 million tons
(73 percent)
0.20 million tons
(31 percent)
0.54 million tons
(84 percent)
0.54 million tons
(84 percent)
Notes: Quantities would be released over a 55-year period from 1998 to 2053
(see Chapter 2 for a detailed discussion of the timing of releases). Percentages
are based on the approximately 0.64 million tons of scrap available in the
commercial nuclear power reactor inventory. Results assume that initial activity
levels are at the logarithmic mid-point of the estimated ranges.
       As discussed in Chapter 2, about 47 percent of the scrap from commercial power reactors
comes from boiling water reactors (BWRs), while 53 percent comes from pressurized water reactors
(PWRs). The estimates of quantities released reflect a similar split:  scrap from BWRs represents
about 50 to 65 percent of the NRC scrap released under each set of standards.  Most of the scrap
from these reactors (76 percent) is carbon steel. Not surprisingly, carbon steel also represents the
majority of the metal likely to be cleared for unconditional use, accounting for about 73 to 75
percent of the quantity released under each option. The costs to release the steel (i.e., surveying
and decontamination costs) under the 1.0 mrem standard equal approximately $224 million, while
the release ofemore valuable metals (e.g., nickel) is likely to generate small cost savings.
                                           4-12

-------
                                                             Preliminary Draft:  June 13, 1997
                               \

       As was the case with DOE facilities, the analysis of commercial power reactors is based on
limited information on the radiological characteristics of the scrap. As discussed in Chapter 2, the
analysis  of commercial nuclear power reactors relies on estimated activity levels  for  three
radionuclides:  Co-60, Cs-137, and Ru-106.  The activity levels for each of these radionuclides varies
depending on the system from which the scrap is derived, and the range in available estimates of
activity often spans several orders of magnitude.

       To determine the effects of uncertainties in the activity levels, we assessed the effects of
Busing the low or high end of the ranges as  well as the logarithmic mid-point. For the commercial
power reactors, the results are less sensitive to activity level assumptions than was the case for DOE
facilities.12 For example, estimates of total disposition costs under the current standards range from
$0.40 billion (using the low end of the reported activity level ranges) to $0.54  billion at the high
end.13  Quantities cleared for unconditional use range from 0.45 million tons at  high activity levels
to 0.48 million tons at low  activity levels. A similar pattern holds true for the other options.
IMPLICATIONS AND PLANS FOR FUTURE ANALYSIS

       The analysis presented above specifically assesses the effects of uncertainties in disposal costs
and activity levels.  However, several other limitations or sources of uncertainty affect the results.
The most important of these limitations are summarized below.
              The analysis assumes  that generators will select the least-cost disposition
              option, ignoring the effects of factors that discourage release of scrap metal.
              The analysis is likely to overstate the quantities of scrap released because it
              ignores the requirements for ALARA analysis (described in Chapter 1) that
              can lead to lower release limits than the limits contained hi current guidance
              or in EPA's rulemaking.  In addition, site managers are often reluctant to
              release metals due to concerns about public perceptions and other factors.

              The analysis probably understates the total  amount  of scrap potentially
              affected  by EPA's rule. As discussed in Chapter 2, the preliminary analysis
              focuses on major scrap sources; other Federal and nonfederal facilities will
              also be affected by the rulemaking.
  12 This finding reflects in part the greater diversity of activity level assumptions for the commercial
power reactors.  Our analysis applies a single set of activity level ranges to all metal at each DOE
site, while the activity estimates for commercial nuclear power reactors vary by system.
  13 This comparison is based on high-end disposal costs.

                                            4-13

-------
                                                              Preliminary Draft:  June 13, 1997


        •      Much of the data on scrap characteristics is uncertain, particularly the data
               on the year in which the scrap is likely to become available for disposal or
               release.   Radiological  characteristics and physical  form  data are also
               uncertain, particularly for DOE  facilities.  These uncertainties may lead us
               to either under- or overstate the effects of alternative release standards.

        •      The cost estimates are also uncertain, particularly for decontamination and
               disposal. Decontamination costs are likely to change as the industry evolves,
               and disposal options are difficult to predict. The extent to which these costs
               are under- or overstated is uncertain; hence, we are unsure whether they lead
               us to under- or overstate the effects of EPA's rulemaking.

        •      The analysis does not consider restricted recycling.  Restricted recycling may
               provide a lower-cost alternative for  some of the scrap included in this
               analysis.  It would therefore reduce total scrap disposition costs,  as well as
               the quantities disposed or released for unconditional use.14


Over the next several months, we may conduct additional sensitivity analyses to determine the most
important sources of uncertainty. These analyses could include, for example, calculation of break-
even points (e.g., for disposal vs. decontamination costs) and the use of probabilistic models (e.g.,
for the distribution of activity levels) to more fully characterize the implications of key uncertainties.
EPA may then employ the results of these analyses to establish priorities for future  research.
  14 DOE policy expresses a preference for unconditional clearance. However, the extent to which
sites would release materials in cases where restricted recycling results in significantly lower costs
is unclear.  In addition, this policy does not affect NRC licensees.

                                            4-14

-------
                                                            Preliminary Draft: June 13, 1997
CHANGES IN HEALTH EFFECTS
ATTRIBUTABLE TO THE REGULATIONS                                    CHAPTER 5
INTRODUCTION AND SUMMARY

       The introduction of EPA standards for recycling scrap metal from nuclear facilities may
affect the quantity of scrap  entering unrestricted recycling and alter the nature and degree of
exposure to radionuclides present hi the scrap. This chapter describes the methodology employed'
to assess changes in human cancer risks associated with EPA's rulemaking and presents the results
of a preliminary assessment of potential changes in these risks.
Summary of Methodology

       The risk assessment methodology employed hi this chapter is based on a method used in
recent analyses for EPA and the Department of Energy.  An EPA Technical Support Document
(TSD) explains the method in detail.1 The basic approach to evaluating individual risk consists of
assessing the radiation exposures of individuals at various stages of recycling scrap metal, including
transport and processing workers, manufacturing workers, users of products made from recycled
scrap, and workers and members of the general population exposed to byproducts of the steelmaking
process. Based on these exposure assessments, a reasonably maximally exposed (RME) individual
can be identified for each radionuclide that is a potential contaminant of the scrap metal. The
estimated doses and risks to the RME individual (RMEI) are normalized to a concentration of 1
pCi of the radionuclide per gram of scrap; these normalized doses can then be used to develop
release standards that are  protective of the most-exposed individual (e.g., release standards that
reduce doses to less than one millirem per year for the most-exposed individual).


       Using a different set of models, EPA's  analysis also estimates collective impacts on the
general population. This analysis focuses on a set of exposures that may apply to significant segments
of the population (e.g., people exposed to emissions from steel plants that refine scrap).  Using
  1 U.S. EPA Office of Radiation and Indoor Air. Technical. Support Document:  Evaluation of
the Potential for Recycling of Radioactively-Contaminated Scrap Metals. Vols. I and n.  Prepared
by S. Cohen & Associates, Inc. Forthcoming.

                                           5-1

-------
                                                             Preliminary Draft: June 13, 1997


models that incorporate the release standards developed in the individual risk analysis, the collective
impacts analysis estimates the change hi the number of cancer cases that would be expected to
occur. Exhibit 5-1 illustrates the relationship between the individual dose and collective impacts
analyses.
                                                                                          r
       It is important to consider the baseline relative to which incremental risks are being assessed/
As with  other aspects of this preliminary analysis,  we focus on a maximum  permissible-release
baseline  that assumes unrestricted recycling of all scrap that can be cost-effectively recycled while
adhering to existing maximum release levels (e.g.  Regulatory Guide 1.86). The change in risk
therefore depends on the amount of scrap released  and the level of the proposed EPA standards
relative to existing standards.
Summary of Preliminary Findings

       To test the preliminary risk assessment methodology, we have estimated RMEI doses and
collective impacts for a set of five indicator nuclides. As discussed earlier, these five nuclides are
those believed to be most prevalent in scrap from the sources of interest. Specifically, three nuclides
(Cs-137, Pu-239, U-238) occur commonly in scrap from DOE facilities, while Co-60, Cs-137, and Ru-
106 occur commonly in scrap from NRC facilities. As shown in Exhibit 5-2, EPA hopes to revise this
methodology as needed and  complete the analysis with improved information on a fuller set of
nuclides in 1997.

       The key preliminary findings  of the human health risk analysis  include the following:

       •      The reasonably maximally exposed individuals for each nuclide are primarily
              workers involved in the processing of scrap or the handling of steelmaking
              byproducts: scrap cutters or slag yard workers. In only three cases does the
              reasonable maximum exposure scenario involve exposures of individuals other
              than industrial workers: residents near the steel mill drinking contaminated
              ground water or eating food contaminated by airborne effluent emissions.

       •      Based on the indicator nuclides considered, some  of the activity limits
              corresponding to a 1.0 mrem per year target dose would be more protective
              of human health than  current Regulatory Guide 1.86 standards, while others
              would be less protective.
                         , n
                                            5-2

-------
                                           Exhibit 5-1

                         MAJOR COMPONENTS OF HUMAN HEALTH
                                      RISK ASSESSMENT
           INDIVIDUAL RISK ASSESSMENT
                                       COLLECTIVE IMPACTS ASSESSMENT
 Individual Ddse-
 Assessment fgr -
    Relevant ?.
Exposure Scenarios
    Normalized
    RME Dose
 (mrem/y per pCi/g)
. From Each Nuclide/
     in Scrac
Release Standards
(pCi/g) Consistent
 with Maximum
Dose Requirement
 (e.g. 1 mrem/y)
   Estimate of
Collective Impacts,
  Using Release
  Standards from
  Individual Risk
   Assessment
Change in
 Cancer
Incidence
 (cases)

-------
                                   Exhibit 5-2
                 ANALYTIC STEPS AND PLANNED SCHEDULE
Initial Development
'•   of Models
 (December 1996)
Test Method with
     Set of
Indicator Nuclides
 (February 1997)
 Revise Methodology
and Complete Analysis
with full Set of Nuclides
     (Late 1997)

-------
                                                              Preliminary Draft: June 13, 1997
              The analysis of collective impacts is summarized in Exhibit  5-3. Several
              findings are noteworthy. First, the total cancer cases expected over the 1,000-
              year modeling time frame range greatly. For the high disposal cost/mid-point
              activity level scenario, cancer cases associated with the 0.1 mrem standard are
              less  than one; this figure rises to approximately four cases under the 1.0
              mrem standard,  and  to 43 cases under the 15 mrem standard. As we would
              expect, cancer incidence for the low cost disposal option is less (from less
              than 1 to 28) because less scrap enters unrestricted recycling.2
Exhibit 5-3
SUMMARY OF PRELIMINARY ESTIMATES OF CANCER INCIDENCE
(assuming activity levels at the logarithmic mean of the estimated range)
High Disposal Costs
Current
14.4
0.1 mrem
0.1
1.0 mrem
4.4
15 mrem
43.2
Low Disposal Costs
Current
83
0.1 mrem
<0.1
1.0 mrem
1.9
15 mrem
27.5
Note: Cancer incidence equals the total number of cancer cases predicted to occur
over 1,000 years.
              A second key rinding concerns the comparison of cancer incidence under the
              proposed release standards relative to the regulatory baseline. As shown, the
              0.1 and 1.0 mrem per year standards result in fewer cancer cases than in the
              baseline, while the 15 mrem standard is associated with a potential increase
              in total cancer incidence.
Structure of Chapter

       The remainder of this chapter is structured as follows:

       •      Fust, we  review the exposure  scenarios addressed in the assessments of
              individual and collective impacts, and describe the analytic approach used to
              estimate the dose and risk under each scenario.
   2 For this analysis, we assume burial entails zero cancer risk.

                                            5-5

-------
                                                              Preliminary Draft: June 13, 1997
               Next, we present the preliminary risk assessment results for five indicator
               nuclides. These results include the  exposure assessment of  the RME
               individuals, the  associated release standards,  and  the change in cancer
               incidence under the alternate release standards.

               Finally, we describe the key uncertainties in the analysis and propose next
               steps that would address these uncertainties.
ANALYTIC APPROACH

       In  this section, we discuss  the methods used  to estimate individual dose and collective
impacts associated with new release standards for scrap. Note that we discuss the method as it
pertains to the full set of nuclides  governed by the standards. Later, we test this method with a
limited set of data based on five indicator nuclides.3
Individual Dose and Risk

       One important  question in assessing the benefits  and costs of new standards  for scrap
recycling is whether any individuals that come in contact with released scrap or the products of the
recycling process are exposed to unacceptably high radiation  levek. The key steps in estimating
radiation exposures of individuals associated with unrestricted  recycling include:

       •      identifying the full set of potentially exposed individuals (i.e.,  exposure
              scenarios);

       •      identifying the pathways by which each of these  individuals are exposed, and
              characterizing the physical process by which radiation reaches the receptor;
              and

       •      estimating the relative concentrations of  each  radionuclide in the various
              media which are potential sources of radiation  exposure.

       Below, we discuss the analysis associated with each of these components. The results of the
exposure assessment are normalized to a concentration of one picocurie of each radionuclide per
gram of unrestricted-recycled scrap. As described below, the analysis uses the normalized dose to
the RME  individual to determine the concentration of each radionuclide  that corresponds to
alternative clearance standards  (e.g., one  millirem per  year), based on  assessments of various
exposure scenarios.            '• '                      '     	
  3 As noted, a more detailed explanation of the analytic approach employed in the risk analysis can
be found in the Technical Support Document.

                                            5-6

-------
                                                             Preliminary Draft: June 13, 1997
Scrap Recycling and Associated Exposure Scenarios
       The normalized radiological impacts on the RME individual developed for this rulemaking
are based on a systematic analysis of the potential exposures associated with various stages of the
unrestricted scrap recycling process: transporting the scrap away from the nuclear facility; unloading
and processing (e.g., sorting, shredding); transport to a steel mill or other refining facility; loading
and melting of processed scrap; and fashioning molten metal into intermediate products such as
slabs  or coils of sheet metal/ The analysis also considers potential exposures associated with the
management of recycling by-products, such as baghouse dust (from the refining facility) and slag.
Baghouse dust accumulates in the emissions control system and must be periodically collected and
disposed of, usually off-site. Slag is formed in the melting process when impurities,float to the
surface of the molten steel and  are poured off. The slag is stored in a pile and later transported to
a processing facility.
    i                                                    *
       To better define  the scrap recycling process  and identify many  of the relevant exposure
pathways, EPA constructed a reference steel mill employing an electric  arc furnace (EAF). EAF
mills primarily use scrap as input and thus represent a high-end risk scenario. Basic oxygen furnaces
use a smaller proportion  of scrap metal  and thus would be the source of lower potential exposures
than EAFs. The TSD explains the features of this reference facility in detail.

       Exhibit 5-4 illustrates the various exposure scenarios evaluated.  As shown, the individual
exposure assessment considers a comprehensive set of detailed exposure scenarios. These primarily
consist of worker exposures during the transport and processing of scrap, exposures to  a variety of
workers at the steel plant, and exposures to users of the products from the steel mill (steel and slag).
Exposure Pathways

       Summary descriptions of the  exposure scenarios are found in Exhibit 5-5. Most of the
scenarios involve one or more of the following major environmental pathways: external exposure;
inhalation of contaminated dust; inadvertent ingestion of soot or deposited dust; and ingestion of
contaminated food or water. As discussed below, each exposure pathway comprises a set of analytic
assumptions.
  4 Initial decontamination of scrap at the decommissioned facility is not considered because these
operations take place in a radiation controlled area.

                                            5-7

-------
                                    Exhibit 5-4
                      INDIVIDUAL RISK EXPOSURE SCENARIOS
  SCRAP       SCRAP
TRANSPORT   PROCESSING
STEEL MILL/RECYCLING
 NEARBY RESIDENTS AND
USERS OF STEEL AND SLAG

-------
                  Preliminary Draft: June 13, 1997
Exhibit 5-5
EXPOSURE SCENARIOS FOR INDIVIDUAL DOSE ESTIMATION
Location/Stage of
Recycling
Transport
Processing
Steel Mill
Users of Mill
Products and End
Products
Exposure Scenario
Truck driver transporting scrap to the processing
facility (A) .
Worker involved in cutting and otherwise
processing scrap (B)
Operator of the crane loading scrap to the
electric arc furnace (C)
Operator of the electric arc furnace (D)
Operator of the continuous caster (E)
Bag house maintenance worker (F)
Truck, driver transporting baghouse dust (G)
Slag pile worker (H)
Slag leaching to groundwater and affecting offsite
drinking well (I)
Subsistence fanner consuming food contaminated
by airborne effluent emissions (J)
Worker using slag in road construction (K)
Worker assembling automobile engines (L)
Manufacturer of industrial lathe (M)
User of kitchen range (N)
Taxi driver (O)
Lathe operator (P)
User of cast iron pan (Q)
Exposure Pathway
External exposure to scrap
External exposure to scrap
Internal (inhalation and ingestion of dust)
External exposure to scrap
Internal (inhalation and ingestion of dust)
- External exposure to scrap
Internal (inhalation and ingestion of dust)
External exposure to steel
Internal (inhalation and ingestion of dust)
External exposure to baghouse dust and to
scrap (when not working in baghouse)
Internal (inhalation and ingestion of dust)
External exposure to baghouse dust
External exposure to slag
Internal (inhalation and ingestion of slag
dust)
Internal (consumption of contaminated
drinking water)
Internal (ingestion of contaminated food)
External exposure to slag
Internal (inhalation and ingestion of slag
dust)
External exposure to cast iron
External exposure to cast iron
Internal (inhalation and ingestion of iron
dust)
External exposure to steel
- External exposure to steel
- External' exposure to cast iron
External exposure to cast iron
Internal (ingestion of contaminated food)
5-9

-------
                                                              Preliminary Draft: June 13, 1997
       External Exposure
       The external exposure pathway is significant hi all scenarios in which individuals are hi the
proximity of material containing radionuclides that emit penetrating radiation in the form of gamma
or x-rays. As noted in Exhibit 5-5, external exposure is relevant to virtually every exposure scenario.

       EPA estimates external radiation doses in most of the scenarios using a commercially-
available computer model, MicroShield* Version 4.21. If we assume a uniform concentration (e.g.,
1 pCi/g) of a given  radionuclide in a given material with a  given geometry, the dose is simply a
function  of the distance from the source, the intervening shielding (if any) and the length of time
the individual is exposed. Each exposure scenario is modeled individually. For example, the analysis
of risks to an operator of the scrap crane at a steel mill assumes that the worker is located 10 meters
away from a rectangular charging bucket filled with scrap and is exposed for 1,750 hours per year.
The technical support document reports the detailed set of exposure parameters for all the exposure
scenarios.

       In cases where the source medium can be approximated by an infinite  or semi-infinite slab
of infinite thickness, the dose was  calculated using the dose coefficients for exposure  to  soil
contaminated to an infinite depth which are listed in Federal Guidance Report No. 12.
       Internal Exposure

       In many of the operations described in Exhibit 5-5, radioactive material can be dispersed in
the air, creating the potential for internal exposure via the inhalation pathway. For example, this may
occur when scrap is cut during processing or When volatile materials are released from the electric
arc furnace during melting. Certain radionuclides concentrate hi by-products of the steelmaking
process such as slag  and baghouse dust.  These materials  can also be  dispersed in the air and
subsequently inhaled by workers handling these materials.

       The inhalation dose of  radioactivity for any given nuch'de is a function of several factors,
including the breathing rate of the exposed individual and the duration of exposure. In addition, the
dose depends on the assumed  concentration of dust in  the air, the respirable fraction (i.e., the
fraction of the dust that consists of particles small enough to inhale), the concentration of the given
radionuclide in the dust, and the dose conversion factor for inhalation of the given radionuch'de.

       The inhalation of contaminated dust is generally accompanied by the incidental ingestion of
contaminated dust, soot or other loose, finely divided material. The doses received via this pathway
depend on the  hourly ingestion rate of such material, the exposure duration, the radionuclide
concentration in the ingested material, and  the dose conversion factor for ingestion of the given
radionuch'de.
                                            5-10

-------
                                                             Preliminary Draft: June 13, 1997
       Unique Exposures
       The analysis of individual risk considers three other exposure pathways that may result in
risks — groundwater contaminated by leachate from slag piles, consumption of food cooked in cast
iron cookware made from scrap, and ingestion of crops and other food products contaminated by
airborne effluent emissions from the recycling facility.

       In the case of groundwater, radionuclides present in slag may migrate from the slag pile and
contaminate nearby drinking water wells. As explained in detail in the technical support document,
the procedure used to  estimate transport  and subsequent exposure is complex. Essentially, the -
analysis incorporates several factors:

       •     the rate at which radionuclides leach from the slag pile (itself a function of
              the physical properties of the slag and the configuration of the slag pile);

       •     the dilution that occurs as the slag leachate travels through soil and mixes
              with the groundwater; and

       •     the radioactive decay 'that occurs during migration.

The analysis of fate and transport yields an estimate of the concentration of each radionuclide in
the drinking water. This concentration is multiplied by the rate of water ingestion (assumed  to be
two liters per day) and the dose conversion factor to estimate the dose of each radionuclide.

       The second unique exposure scenario is the ingestion  of food prepared  hi contaminated
cookware.  Radioactivity could potentially  leach from, such cookware. EPA estimates the dose
associated with this pathway by first estimating the concentration of iron in food  cooked in a cast
iron pan. We then assume that the  concentration of each radionuclide in the food, relative to the
concentration of iron in the food, is the same as the concentration of that radionuclide hi the iron
pan. This is the same as assuming that the iron that dissolves hi the food brings with it its own share
of radioactivity. The estimated concentration in food is then multiplied by the food ingestion rate
and by the dose conversion factor for ingestion of the given nuclide; ingestion of both beef and
vegetables is considered in the analysis.

       The third unique exposure scenario involves nuclides released to the air from the steel  plant.
These contaminants can be absorbed by food crops or by fodder consumed by livestock. People eat
the food crops and other products such  as milk and meat.

       The doses in this scenario are modeled using EPA's CAP-88 computer  code. The  RME
individual is a hypothetical  subsistence  fanner who gets  a large portion of his food  from locally
grown produce and livestock.
                                            5-11

-------
                                                              Preliminary Draft: June 13, 1997
Estimating Specific Activity in Exposure Media
       As noted, an electric arc furnace steelmaking facility is used as the reference facility to
estimate individual risks from scrap recycling. The steelmaking process yields a number of products
and by-products  that can assume the radiation originally contained in the incoming scrap.  The
exposure pathways reviewed above demonstrate that the relevant exposure media include the scrap
itself, steel made from the scrap, cast iron made from the scrap, as well as slag and baghouse dust.
This section examines how EPA estimates the concentration of radionuclides  in each of these
exposure media. These concentrations are then used in the estimation of dose as reviewed above.

       One key exposure medium is the scrap itself. To estimate the specific activity in the scrap,
the analysis first estimates a mix of scrap containing residual radioactivity from decommissioned
nuclear facilities and scrap from other sources at the generating facility. The analysis  assumes that
scrap from nuclear facilities that contains some level of residual radioactivity is first mixed with other
clean metal from the facility and then sent to a steelmaking facility and mixed with more clean
metal. Based on an analysis of the  fraction of scrap from DOE decommissioning activities  that is
formerly radioactive, EPA estimates that about 13 percent of the scrap sent for recycling contains
residual radioactivity. Second, the analysis assumes that 88 percent of the total scrap volume going
to the furnace is from DOE/NRC facilities. Therefore, 11 percent of the scrap fed to the furnace
contains some residual radioactivity (0.13 x 0.88).5 This dilution factor plays an important role in
estimating doses and associated  risks from unrestricted recycling.

       The radionuclides present in scrap loaded into the furnace are distributed among the various
media produced in steelmaking. Elements vary greatly in terms of how they partition to the different
media. For example, as much as 100 percent of carbon can remain in the steel. In contrast, no
radium remains in steel; instead, 95  percent partitions to slag, while five percent collects  in the
baghouse dust.

       Depending upon the mass of slag and baghouse dust generated per unit of scrap fed to the
furnace, as well as upon the partitioning behavior of the particular nuclide, nuclides may concentrate
or be diluted in the slag and dust. Therefore, each element is assigned a concentration factor in the
analysis.  For example, 0.117 tons of slag are produced for each ton of metal fed to the furnace.
Therefore, when 95  percent of the radium from the charge partitions to the slag, the radium is
concentrated by a factor of 7.8. Exhibit 5-6 summarizes the partition ratios and concentration factors
used in the analysis.
  5 See the Technical Support Document for a more detailed discussion of the development of this
dilution factor.

                                            5-12

-------
                                                             Preliminary Draft: June 13, 1997
Maximum Exposure Scenarios
       As described above, we develop the annual dose to each exposed individual, normalized to
a unit specific activity (i.e., 1 pCi/g) of each radionuclide in scrap. This analysis allows us to identify
the reasonably maximally exposed (RME) individual for each radionuclide. Exhibit 5-7 presents a
summary of the maximum exposure scenarios and the normalized dose and lifetime risk of cancer
to the RME individual from one year of exposure. Several findings are noteworthy. First, except in
three cases, the RME individuals are workers. One exception is Sr-90; the teachability and mobility
of this nuclide suggests that it may leach readily from the slag pile, potentially contaminating well
water drunk by a nearby resident. The other two exceptions are C-14 and 1-129; the RME individual
in both cases is the subsistence fanner residing near the steel plant.

       Four worker scenarios account for the RME individuals for the remaining nuclides. The
scenario associated with the greatest number of nuclides is that involving the slag pile worker. The
25 relevant nuclides are those that readily partition to and concentrate in slag (see above). For five
of these nuclides,  external  exposure is the dominant pathway.6 For the remaining nuclides, dust
inhalation is the primary pathway.

       Other worker scenarios that account for maximum  exposures include the worker cutting
scrap, the furnace  operator, and the lathe operator  using a lathe made from scrap. In the results
section below, we will review how these RMEI doses are used to develop release  standards.
  6,The  external exposure pathway is  significant for the strong  emitters of gamma  radiation,
including Nb-94, Ce-144+D, Eu-152, Ra-226 and Ra-228.

                                           5-13

-------
                   Preliminary Draft: June 13, 1997
Exhibit 5-6
PARTITION RATIOS (PR) AND CONCENTRATION FACTORS (CF)
Element
Ac
Ag
Am
C
Ce
Cm
Co
Cs
Eu
Fc
I
Mn
Mo
Mb
Ni
Np
Pa
Pb
Pm
Po
Pu
Ra
Ru
Sb
Sr
Tc
Th
U'
Zn
Steel
PR
(%)
0
' 99/75

100/27


99


97

24/65
99

99







99
99/80

99'"


20/0
CF
0
1.02
0
1.03
0
0
1.02
0
0
1
0
0.67
1.02
0
1.02
0
0
0
0
0
0
0
1.02
1.02
0
'• 1.02
0
0
0.2
Cast Iron
CF
0
1.01
0
1.01
0
0 '
1.01
0
0
1
0
0.98
1.01
0
1.01
0
0
0
0
0
0
0
1.01
1.01
0
• 1.01
O1
0
0.02
Slag
PR
(%)
95

95

95
95

0/5
95
2

72/32

95

95
95

95

95
95


95
,*. i
95
•95

CF
7.79
0
7.79
0
7.79
7.79
0
0.41
7.79
0.19
0
6.15
0
7.79
0
7.79
7.79
0
7.79
0
7.79
7.79
0
0
7.79
••"o ' "
7.79n
7.'79
0
Baghouse
PR
(%)
5
1/25
5


5
1
100/95
5
1

4/3
1
5
1

5
100
5
100

5
1
1/20
- 5
• ' r
!
5
80/100
CF
2.6
16.5
2.6
0
2.6
2.6
0.67
633
2.6
0.67
0
2.21
0.67
2.6
0.67
2.6
2.6
63.3
2.6
63.3
2.6
2.6
0.67
13.2
2.6
0.67
2.6 ,
2.6
63.3
5-14

-------
                  Preliminary Draft: June 13, 1997
Exhibit 5-7
MAXIMUM EXPOSURE SCENARIOS AND IMPACTS ON THE RME INDIVIDUAL FROM ONE
YEAR OF EXPOSURE, NORMALIZED TO 1 pO/g IN SCRAP
Nuclide
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
Ag-llOm+D
Sb-125+D
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Maximum Scenario
Airborne effluent emissions
Lathe operator
Cutting scrap
Lathe operator
Cutting scrap
Cutting scrap
Cutting scrap
Slag leachate in groundwater
Slag pile worker
Cutting scrap
Cutting scrap
Lathe operator
Lathe operator
Cutting scrap
Airborne effluent emissions
Cutting scrap
Cutting scrap
Slag pile worker
Slag pile worker
Slag pile worker
EAF furnace operator
Slag pile worker
Slag pile worker
Cutting scrap
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile v/orker
Slag pile worker
Slag pile worker
Dose
(mrem EDE)
8.66e-04
. 2.02e-01
6.69e-06
\
8.99e-01
4.396-06
1.076-05
9.61e-02
1.526+00
4.746-01
5.656-05
2.15e-05
5.16e-02
6.296-01
6.37e-02
7.91e-01
2.46e-01
8.916-02
1.77e-02
1.426-04
3.446-01
3.08e+00
6.276-01
' 3.68e-01
8.00e+00
135e+00
437e+00
6.426-01
2.846-1-00
2.51e+00
3.14e-01
3.286-01
2.896-01
1.53e+00
6.82e-01
7.29e-01
Lifetime Risk
of Cancer
4.28e-10
1.54e-07
2.71e-12
6.84e-07
1.556-12
4.41e-12
731e-08
5.51e-07
3.60e-07
1.17e-ll
1.416-11
3.93e-08
4.78e-07
4.85e-08
5.04e-07
1.87e-07
6.77e-08
1.36e-08
8.31e-ll
2.61e-07
4.37e-07
4.36e-07
2.36e-07
1.35e-07
6.17e-06
. 2.32e-07
3.44e-08
3.34e-08
5.20e-08
3.31e-08
5.90e-08
3.55e-08
136e-07
4.78e-08
4.73e^)8
5-15

-------
                                                             Preliminary Draft: June 13, 1997
Exhibit 5-7
MAXIMUM EXPOSURE SCENARIOS AND IMPACTS ON THE RME INDIVIDUAL FROM ONE
YEAR OF EXPOSURE, NORMALIZED TO 1 pCi/g IN SCRAP
Nuclide
Pu-240
Pu-241+D
Pu-242
Am-241
Cm-244
U-Natural
U-Separated
U-Depleted
Th-Senes
Maximnm Scenario
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Dose
(mrem EDE)
7.29e-01
1.17e-02
6.93e-01
1.21e+00
6.75e-01
3.61e+00
6.18e-01
3.22e-01
4.55e+00
Lifetime Risk
of Cancer
4.73e-08
4.01e-10
4.46e-08
1.07e-07
6.69e-08
4.78e-07
7.14e-08
3.95e-08
8.86e-07
Collective Impacts

       The collective risk assessment estimates the change hi cancer incidence attributable to
changes in the release standards for scrap. As we have noted, the analysis is linked to the individual
risk analysis because the specific activity assumed for the scrap is consistent with the standards
developed in the individual risk model. The analyses diverge, however, in that the collective risk
assessment is based on a separate set of exposure scenarios and dose estimation models. Specifically,
we examine exposure scenarios that involve larger segments of the population. These include the
following:

       •      Risks to surrounding populations when scrap, finished steel, and steelmaking
              by-products are transported;

       •      Risks to surrounding populations from the airborne release of radionuclides
              during steelmaking;

       •      Risks from the various uses of slag generated during steelmaking;

       •      Risks from the processing, disposal, and use of baghouse dust; and

       •      Risks from the use of finished products made from scrap.
                          ,. .    •      j   .,,., t u'.,...    ,;            ..  -.       '.
These exposure scenarios  are summarized in Exhibit 5-8. Betow,5 we discuss each of the scenarios
and the models used to estimate .the impacts associated wftb each;
                                           5-16

-------
                    Exhibit 5-8
SCENARIOS FOR ASSESSING COLLECTIVE RADIOLOGICAL
         IMPACTS FROM RECYCLING SCRAP
                           Airborne Effluents
                             Slag

-------
                                                              Preliminary Draft: June 13, 1997


        Before proceeding, it is important to understand the general method by which the collective
 impact assessment is linked to  other aspects of this analysis. The cost analysis reviewed earlier
 estimates the total scrap entering unconditional clearance and specific activity of each nuclide in the
 scrap; this information tells us  the total amount (in mCi) of each nuclide sent into free  release
 across all shipments of scrap in a given year. In turn, the models developed for each of the exposure
 scenarios estimate the collective  dose per mCi of each nuclide sent into unconditional clearance. For
 external exposure,  this dose is expressed in person-rem per year; the total person-rem are then
 multiplied by EPA risk factors  to estimate total cancers. For internal exposure (inhalation and
 ingestion), the total mCi ingested or inhaled are multiplied by EPA cancer slope factors that convert
 total intake to cancer cases. For all forms of exposure, cancer impacts are calculated over 1000
 years; while this tune frame does not account for complete decay of many radionuclides,  it does
 represent the period beyond which EPA believes cancer impacts cannot be credibly predicted. Given
 this relatively long time horizon,  the change in cancer incidence reported here should be interpreted
 as a total cancer "commitment"  for the estimated scrap sent to unconditional  clearance in  a given
 year.-

        Note that  the  analysis does not require knowledge of  the  total number of individuals
 exposed, as is generally the case in a population  risk assessment. Instead, the impacts are assumed
 to be directly proportional to the total inventory of radioactivity to which people are exposed (i.e.,
 the collective dose). The impacts (i.e., cancer cases) are the same whether the activity affects a small
 population or a large population.


 Transport

      ' The radioactive material  - both the scrap and the products and by-products (slag, baghouse
 dust) produced from recycling —  will be transported by truck or rail (shown as trucks in Exhibit 5-8).
 Populations along the transportation route may be exposed externally to radiation from the material.
To assess the collective impacts from transport, EPA modeled this pathway as a function that
 estimates the  collective dose for each radionuclide in person-rem. This  dose is estimated as a
function of several factors, including the population density along the transportation route, the
length of the  shipment route, the speed of the truck or train,  and the distance to  the  nearest
receptor. The  specific elements of the equation are discussed in more detail in the TSD.
Air Emissions

       Another means by which the general population potentially could be exposed to radioactivity
in scrap is the airborne release of radionuclides from the recycling facility (e.g., the steel plant). To
assess the collective impacts from  air releases, EPA performed simple air dispersion modeling to
estimate the concentrations  of radionuclides that surrounding populations could inhale.  EPA
modeled dispersion around the recycling facility  using the CAP88-PC computer program. The
assessment encompasses typical meteorological, land use, and demographic patterns for the U.S. and
incorporates impacts within a 50-mile radius around the facility.

       In addition, EPA modeled the long-term  impacts from deposition, i.e., the transport of

                                            5-18

-------
                                                            Preliminary Draft: June 13, 1997


radioactive particles from air,to land. Deposited radionuclides can  affect humans by entering the
food chain. Specifically, edible plants  can absorb radionuclides and  grazing livestock can ingest
radionuclides. To estimate the radionuclide concentrations in plants (produce, leafy vegetables) and
animal products (milk and meat), we first assume that all radioactivity is deposited to the ground.
We then use the CU-POP model to estimate uptake by plants and animals and the resulting leaf and
tissue concentrations, as well as exposures via ground water.


Slag Management

       Slag from the steel plant is used  in a variety of ways and  therefore can contribute to
collective impacts via several pathways. These uses include:

       •      Road Base: On a national basis, thirty-five percent of slag from electric arc
              furnaces is assumed to be used as underlayment in road construction.7 The
              risk modeling examines two exposure pathways -  external exposure and
              exposure from leached contaminants. For external exposure, EPA estimates
              the collective dose to occupants of vehicles that travel on roads using slag as
              a base. Exposure to radioactivity leached from the  road base  is a second
              potential exposure pathway. Radionuclides could be leached from the road
              base and be washed to a surface or ground water drinking source. Due to the
              fact that the majority of the road base would be covered with pavement, it
              was assumed that this  leaching would be small, and  exposures from this
              pathway  were  not  explicitly   included.  Nonetheless,  for comparison,
              normalized  collective  exposure impacts  were  calculated  for leached
              radionuclides reaching a river water system.

       •      Road Pavement: Another 13 percent of slag is used in asphaltic concrete
              aggregate for road pavement. For the assumed useful road life of 100 years,
              motorists may be exposed to external radiation while driving on the roads.
              Based on discussions with a number  of state DOT officials,  it was assumed
              that road pavement material would be recycled back into roads for the entire
              1000 year evaluation  period. Dose coefficients published in EPA Federal
              Guidance No. 12 were utilized in  calculating the impacts to motorists
              assumed to be exposed  to external radiation while driving on the roads.

       •      Fill: Another 16 percent of slag is used as fill, e.g., underlayment for lawns
              or gardens. Slag used as fill was modeled as a municipal landfill (i.e., material
              with top soil cover). This analysis uses DOE's  Multimedia  Environmental
              Pollutant Assessment System (MEPAS) to model migration of nuch'des from
                     • '    -          •      '                   ~
   7 See the Technical Support Document for a more detailed discussion.

                                          5-19

-------
                                                             Preliminary Draft: June 13, 1997


              the landfill. MEPAS addresses migration to a variety of media, including air,
              surface water, soil, and ground water, allowing  assessment  of external
              exposure, inhalation, and ingestion.

       •     Railroad  Ballast: Three percent  of slag is used to line railroad beds.
              Exposures from the use of slag as railroad ballast were evaluated with the
              same model that was used for slag in road pavement.

       •     Other Uses: The remaining portion of slag (33 percent) goes to other  uses
              such as soil conditioning (addition to acidic soil) and ice control (cover to icy
              roads). It was assumed that the slag from all other uses would be deposited
              on the ground (either immediately or shortly after use), and the CU-POP
              model was used to characterize exposure via these pathways.


Baghouse Dust Management

       About 87 percent of the baghouse dust from the steel plant is shipped to a processing plant
for recovery of zinc, cadmium, and lead in the dust.  This process generates slag;  we analyze
exposures to the slag hi the same manner as slag from the steelmaking process. The remainder of
the baghouse dust is stabilized and sent to a secure landfill;  landfill exposures were modeled using
MEPAS.
Finished Products

       The finished steel made from scrap is sold in the form of numerous different products. This
analysis looks at a limited set of products and other exposure media and assesses the collective dose
received through external exposure. The products analyzed include automobiles, office desks, beds,
and kitchen appliances. These products, along with a non-accessible product category, were used as
surrogates for all finished steel products. Internal exposures due to radionuclides being leached from
frying pans during cooking were also analyzed.
FINDINGS

       The sections below discuss the release standards developed  on the basis of the RME
individual dose assessment and the collective impacts (i.e., change in total cancer cases) associated
with the rule.
Individoal Dose and Associated Release Standards pr~.        '             f'
                       , i    , i' •       i *  i  '.. •,       '
       Having established  the exposure  scenarios that result in the maximum exposure to each
radionuclide, the normalized doses developed per pCi/g of each nuclide in scrap  can be scaled to

                                           5-20

-------
                                                              Preliminary Draft: June 13, 1997


develop release standards for each nuclide. For example, the maximum exposure scenario for Cs-137
(scrap cutter) results in a dose of 0.0891 mrem per year per pCi/g.8 If we wish to identify a release
standard consistent with a dose of 0.1 mrem per year, we can divide the target maximum dose by
the normalized dose to estimate a release concentration of 1.12 pCi/g (i.e., 0.1/0.0891 = 1.12). That
is, to restrict exposures to Cs-137 to below 0.1 mrem per year, the concentration of Cs-137 in scrap
must be below  1.12 pCi/g.

       Exhibit 5-9 presents the estimates of release standards consistent with maximum target doses
of 0.1, 1.0, and 15.0 mrem per year. For each of the five indicator  nuclides, the release standard is
provided on the basis of volume concentration  (pCi/g) as well as  surface concentration (dpm/100
cm2). In addition, the exhibit shows how the alternative standards compare to the existing Regulatory
Guide 1.86 standards.

       The comparison of the alternative standards analyzed and the existing Regulatory Guide 1.86
standards is more  clearly illustrated in Exhibit 5-10. This chart plots the dose associated with the
Regulatory Guide  1.86 standards relative to a target maximum dose of one mrem/y. Positive bars
reflect Regulatory Guide 1.86 standards  that would allow a dose greater than one mrem/y  while
negative bars show instances where the Regulatory Guide 1.86 standards are lower than the one
mrem/y  standard. As shown, a one mrem/y standard is lower than the current standard for two of
the indicator nuclides (Co-60 and U-238+D), and higher than current standards for three  nuclides
(Cs-137+D, Ru-106+D, Pu-239).

       The results presented above suggest that for some nuclides, such as Co-60, a one mrem/y
release standard would decrease risk relative to that currently experienced under the Regulatory
Guide 1.86 standards. These findings, however, must be interpreted carefully. As we have noted,  we
developed  the  maximum permissible  release baseline to reflect  the greatest possible degree of
baseline recycling (i.e., all scrap that meets Regulatory Guide 1.86 limits is released). In actuality,
all scrap that meets current standards is not released. Furthermore, the scrap that is released is
rarely released  at radioactivity levels close to limits prescribed by  Regulatory Guide 1.86; instead,
specific  activity in released  scrap is generally much lower.
    8  See the Technical Support Document for a more detailed discussion of the scaling method.

                                            5-21

-------
Preliminary Draft: June 13,1997
Exhibit 5-9
CURRENT AND ALTERNATE RADIONUCLIDE RELEASE LEVELS
Nuclide
Co-60
Ru-106+D
Cs-137+D
U-238+D
Pu-239
mrem/y
perpCi/g
8.99E-01
5.16E-02
8.91E-02
2.89E-01
7.29E-01
Surface Concentration
(dpm/100 cm*
0.1
mrem/y
8.70E+01
3.03E+03
8.79E+02
2.71E+02
1.07E+02
1.0
mrem/y
8.70E+02
3.03E+04
8.79E+03
2.71E+03
1.07E+03
15.0
mrem/y
1.31E+04
4.55E+05
1.32E+05
4.07E+04
1.61E+04
Volume Concentration
(pCi/g)
0.1
mrem/y
1.11E-01
1.94E+00
1.12E+00
3.47E-01
1.37E-01
1.0
mrem/y
1.11E+00
1.94E+01
1.12E+01
3.47E+00
1.37E+00
15.0
mrem/y
1.67E+01
2.91E+02
1.68E+02
5.20E+01
2.06E+01
Current Standards
(R.G. 1.86)
Dose
mrem/y
5.75E+00
1.65E-01
5.69E-01
1.84E+00
9.32E-02
dpm/
100cm1
5000
5000
5000
5000
100

-------
10 00 mrem
                                                Exhibit 5-10
                              RME DOSE UNDER EXISTING STANDARDS
 I OOmrem
 0 10 mrem
 001 mrem
IDosc mrem/year

-1 mrem
                Co-60
                                  U-2J8+D
                          Gs-137+D
                         Rsdionndides
                                              Ru-106+D
Pu-239

-------
                                                              Preliminary Draft: June 13, 1997
Collective Impacts
       Collective impacts are expressed as the total cancer cases associated with exposure to scrap.9
Our preliminary findings  are  summarized in  Exhibit 5-11. The total number of cancer  cases
estimated by the collective impacts model varies greatly depending upon the release standard under
consideration. For the high disposal cost/mid-point activity level scenario, cancer cases associated
with the 0.1 mrem standard are only 0.1; this figure rises to 43 cases under the 15 mrem standard.
As we would expect, cancer incidence for the low cost disposal option is less (from less than 1 to
28) because less scrap enters unrestricted recycling.10
Exhibit 5-11
PRELIMINARY ESTIMATES OF CANCER INCIDENCP
High Disposal Costs

DOE Facilities
Commercial Power Plants
Total
Current
<0.1
14.4
14.4
0.1 mrem
0
0.1
0.1
1.0 mrem
<0.1
4.4
4.4
, 15 mrem
1.3
42.0
43.2
Low Disposal Costs

DOE Facilities
Commercial Power Plants
Total
Current
<0.1
8.3
8.3
0.1 mrem
0
<0.1
<0.1
1.0 mrem
<0.1
1.9
1.9
15 mrem
1.3
26.2
27.5
Note: Totals may not sum due to rounding. Results assume initial activity levels at the logarithmic
mid-point of the reported ranges.
" Cancer incidence equals the total number of cancer cases (i.e., fatal and non-fatal) predicted to
occur over 1,000 years.
  9 As noted, we report total cancer cases, both fatal and non-fatal  These cases are associated with
a 1,000 year exposure time frame.  In addition, all model estimates reported here are based on the
assumption that initial activity levels are, at the midpoint of the range noted hi the scrap database.
                               *"     *""   "       -      ijlV    "
  10 It is important to bear in mint! that:all the cancer estimates presented here assume that scrap
is released at the relevant standard. In- reality, activity iff scrap -likely will be below the standard.
As a result, our estimates probably overstate  total cancer incidence.  However, the estimates are
likely to indicate more accurately the relative cancer incidence across the four standards analyzed.

                                            5-24

-------
                                                               Preliminary Draft: June 13, 1997


        Relative to the Regulatory Guide 1.86 regulatory baseline, we estimate that cancer incidence
 would be reduced under the 0.1 and 1.0 mrem release standards. For example, in the high disposal
 cost scenario,  the collective impacts model estimates about four cancer cases  for the 1.0  mrem
 standard, compared  to about  14 cases for the baseline standard. As shown, only the 15  mrem
 standard is  associated with a  potential increase  in  total cancer incidence relative to  current
 standards.                                       ^

        Finally, it is useful to consider the relative contributions of DOE facilities and commercial
 power reactors to collective impacts. As shown, scrap from commercial power reactors yields a much
 higher share of the total cancer cases estimated. For example, at the 0.1 and 1.0 mrem standards,
 scrap released from commercial power reactors accounts for virtually all of the cancer cases, while
 at the 15 mrem standard this scrap accounts for about 97 percent of the cases. This finding may be
 at least partially attributable to the fact that the initial activity levels we assume for scrap from DOE
 facilities are often below the clearance thresholds considered, while scrap from commercial  power
 reactors must be decontaminated to achieve release standards. Therefore, our modeling assumes that
 the activity level in most of the scrap  released from commercial power reactors equals the relevant
 release standard. As a result, estimated risks for commercial power reactors are greater. This finding
 is subject to significant uncertainty, however, given that decontamination procedures may achieve
 activity levels  well below the standard.  In addition, the  results are highly dependent 'on our
 assumptions concerning radiological profiles and should be viewed with appropriate caution.

        To further highlight the differences between DOE facilities and commercial power reactors,
 the sections below provide a detailed discussion of the nuclides and exposure scenarios that drive
 collective impacts.
DOE Facilities

        In this preliminary analysis, Uranium-238 is responsible for virtually all (93 to 99 percent)
of the collective impacts  associated with scrap from DOE facilities. Also notable is the fact that
collective impacts are non-existent under the 0.1 mrem standard. This finding is consistent with the
cost analysis,  which indicates that DOE scrap generators will forego recycling under the stringent
release standard and dispose all scrap.

        Exposure to radionuclides via slag is the key pathway in the analysis of the collective impacts
associated with the release of DOE scrap. As a result, cancer incidence associated with exposure to
slag far outweighs other pathways. For example, under the 15 mrem standard, slag exposures account
for about 95 percent of all estimated cancer cases attributable to DOE scrap. This is likely related
,to  the  prevalence of U-238, a nuclide that partitions readily  to slag. The use  of slag in soil
conditioning is responsible fpr- most of the collective impacts estimated.
                                            5-25

-------
                                                             Preliminary Draft: June 13, 1997
Commercial Power Reactors
       We also break down the collective impacts for power reactors by radionuclide. Co-60 is the
nuclide most responsible for the estimated cancer incidence. For example, for the nigh disposal cost
scenario under a 1.0 mrem standard, Co-60 accounts for about 92 percent of estimated cancer cases.
In comparison, the impacts of Cs-137 and Ru-106 are negligible.

       Finally, the predicted incidence of cancer attributable  to the release  of scrap from
commercial nuclear power reactors appears to be driven almost exclusively by the consumer products
exposure route. This is partially attributable to the strong tendency of Co-60 to partition to steel,
which is used to manufacture consumer products. Likewise, the collective impacts are driven by the
population's extended exposure to kitchen appliances, office appliances, and automobiles, the use
of which comprises this exposure scenario.
UNCERTAINTIES AND NEXT STEPS

       The risk analysis developed to date incorporates a variety of assumptions and is subject to
several uncertainties. The findings must be interpreted carefully in light of these considerations.
Below, we review key issues and describe analyses that EPA may pursue to address them.
Uncertainties

       In the evaluation of the RME individual, the most significant uncertainty may be the dilution
assumed in processing scrap during and after recycling. As discussed, the analysis assumes a certain
mix of scrap from radiation control areas with scrap not exposed to radioactive contamination. In
addition, the analysis assumes a certain mix of scrap from decommissioned nuclear facilities with
scrap from  other sources. The assumptions made are conservative, and therefore may overstate the
dose under the various exposure scenarios, thereby affecting the release standards. For example, the
reference steel plant was located near four decommissioned nuclear facilities, increasing the assumed
amount of scrap handled at the  steel plant. In reality, scrap from decommissioned facilities may be
distributed  more evenly to steel plants and other recycling facilities so that dilution is greater and
risks to exposed individuals are  less than  currently estimated.

       The modeling of the exposure pathways used  to estimate individual dose  incorporates
numerous assumptions and uncertainties. The TSD provides a more complete discussion  of these
issues.  Here we briefly note some of the more significant uncertainties:

       •      Total dust inhaled by the scrap  cutter  was-assumed  to- be. equal:to the •<
              threshold  limit value (TLV)  for nuisance dust; 50 percent of the dust was
              assumed to be respirable. Likewise, the concentration of nuclides in the dust
              was assumed to be the same as that in scrap. Because the scrap cutter is the
              RMEI for several nuclides,  these conservative assumptions may  play  an
              important role in the development of the release standards.

                                           5-26

-------
                                                              Preliminary Draft: June 13, 1997


       •      The dust inhaled by the crane operator, furnace operator, and continuous
              caster operator was assumed to have the same concentration as baghouse
              dust, possibly overstating exposure.

       •      One  nuclide (Sr-90) has a  maximum exposure scenario attributable to
              ingestion of ground water contaminated by leaching from  the slag pile.
              Significant uncertainty surrounds the teachability and mobility factors for Sr-
              90 used to calculate risk for this exposure route.

       The analysis of collective impacts also contains a number of key uncertainties. First, these
impacts are a product of our initial assumptions concerning the radiological profiles of affected scrap
metal. As noted  previously, these  profiles are based  on limited  information  and subject to
uncertainty. As a result, the estimates of cancer incidence could change significantly due to only
small changes in the corresponding radiological profiles. For example, if we add Co-60 as a surrogate
radionuclide at certain DOE facilities and make conservative assumptions concerning the quantity
of Co-60 that is present at these facilities, the cumulative population dose from all DOE facilities
for the 1.0 mrem option could  increase from the current level  of 0.01 person-rems to 93 person-
rems.11 EPA may further examine  these uncertainties by conducting additional sensitivity analyses
over the next several months.

       Second, our cost modeling yields a total maximum quantity of scrap that enters unrestricted
recycling under alternate clearance standards. This quantity has a direct impact on the  collective
dose  estimated in  the risk assessment.  As we  discuss elsewhere, the scrap quantities entering
unrestricted recycling will  likely be less than those estimated, leading us to overstate  collective
impacts. Closely related to this uncertainty is the assumption regarding activity in scrap entering free
release. As noted, the cancer estimates presented here assume that residual activity in  any scrap
requiring decontamination prior to release equals the clearance standard. In reality, activity in scrap
likely will be below the standard. As a result, our estimates probably overstate total cancer incidence.

       Third, as with the individual risk analysis,  each of the exposure pathway models contains
detailed assumptions that bear closer examination. For example, the assumption that steel produced
from scrap will enter a variety of end products may be inaccurate. Instead, geographic or economic
factors may result in scrap entering a more homogenous set of products; this could change the
collective  impacts from those estimated here,  although the direction and magnitude  of the bias is
unclear. Likewise, impacts estimated for the ground water pathway are highly uncertain, but likely
to be conservative, since  the  analysis  uses  relatively  conservative  assumptions regarding the
likelihood of contamination reaching ground water and the likelihood that this water is consumed.
  11 Information provided by S. Cohen and Associates, February 1997.

                                            5-27

-------
                                                              Preliminary Draft: June 13, 1997
Next Steps
       EPA may choose  to pursue  additional research in the coming year  to  address  the
uncertainties discussed above. These tasks may include the following:

       •      A more comprehensive analysis of the likely dilution of formerly radioactive
              scrap with other scrap, both at the facilities generating scrap and at recycling
              facilities; and

       •      Refinement of the  assumptions used  in key RME individual exposure
              scenarios.
                                            5-28

-------
                                                            Preliminary Draft:  June 13, 1997
OTHER IMPACTS                                            <               CHAPTER 6
INTRODUCTION AND SUMMARY

       In addition to the direct cost impacts analyzed in Chapter 4 and the implications for human
cancer risks evaluated in  Chapter 5, EPA's rulemaking  may  have  a variety of other impacts,
including secondary effects on scrap markets and the demand for radioactive waste disposal capacity,
ecological impacts or impacts on non-carcinogenic human health risks, and environmental benefits ,
attributable to reduced demand for virgin materials. This chapter assesses these potential impacts.
The discussion, which is qualitative in nature, identifies the potential magnitude of each impact and
is designed to help develop priorities for future analyses. We will incorporate the results of any such
analyses in subsequent assessments of the impact of EPA's proposed scrap metal clearance standard.

       Based  on our  initial review  of these  issues, we have reached  the following tentative
conclusions:             '

       •      Other economic impacts. None  of the analytical options evaluated is likely
              to have a significant effect on  scrap metal markets  or disposal capacity
              because the quantity of potentially available scrap metal is small relative to
              the total quantity of metal recycled and volume of waste disposed. The only
              possible exception is the release of a significant share of nickel, and perhaps
              stainless steel, from the DOE and commercial nuclear power plant inventory.


       •      Effects of public opinion. Our analysis of direct cost impacts and cancer risks
              assumes that scrap metal will be cleared for unconditional use whenever the
              costs associated with meeting the clearance standard are less than the costs
              of disposal. Concerns about human health and environmental effects on the
              part  of affected industries  and  the general  public  over the release  of
              radioactive  material may be significant enough to limit scrap metal releases
              to quantities tesythzw fthis simple eco«dmic'ev^ufitronilsuggests.   '  '
                 '     "' '             -~' "•                    'ji>'         '
                                           6-1

-------
                                                             Preliminary Draft:  June 13, 1997


        •     Noncancer and ecological effects.  The noncancer human health risks or
              ecological risks associated with EPA's rulemaking are likely to be minor, and
              insignificant in comparison to the rule's estimated impacts on human cancer
              risks.

        •     Toxic and conventional releases.  Effects related to toxic or conventional
              pollutants are not likely to be significant unless the EPA standards lead to
              changes in practices in areas where existing regulations are not adequately
              protective.

        •     Transportation risks.  Each of the disposition alternatives considered  in this
              analysis, including  unconditional  clearance and  burial,   generate similar
              transportation risks, indicating that the clearance standards will not result in
              increased accidents and fatalities during shipping of the material.

        •     Demand  for virgin materials.   The clearance  standards  may  generate
              environmental  benefits resulting from lower demand for virgin materials due
              to increased recycling activity. The magnitude of these benefits is uncertain,
              however, and dependent on the extent to which  scrap metal from nuclear
              facilities might  serve as a substitute for virgin material rather than simply as
              a substitute for scrap metal from other sources.

        This  chapter  does not address  several other potentially significant  impacts, including
environmental justice  issues,  effects  on small businesses, costs imposed by unfunded Federal
mandates, and the relationship of EPA's standards  to other governmental programs. Assessments
of these potential impacts, which are generally required for all major Federal rulemakings, may be
included in future analyses.

        This chapter begins with a discussion of the other economic impacts potentially attributable
to the rulemaking. The second  section discusses noncancer human health effects and other potential
environmental impacts.  Potential effects resulting  from  a reduction in the  demand for virgin
materials are presented in the final section.
OTHER ECONOMIC IMPACTS

       In this section, we discuss the potential effects of EPA's standards on scrap metal markets
and on the demand for disposal capacity. We also discuss potential end uses of recycled metal and
consider the extent to which the attitudes of industry or the general public are likely to affect the
quantities recycled.

       These  impacts are directly related to the direct cost savings and risk results discussed in
previous chapters. For example, the acceptance of recycled metal by the general public may have
significant implications that should be taken into account in Chapter 4's assessment of the quantities
of scrap recycled. In  addition,  the end uses of products  are an important component  of the risk
analysis described hi Chapter 5.
                                            6-2

-------
                                                             Preliminary Draft: June 13, 1997


       Below, we first discuss the economic framework for assessing market effects. We then review
 existing data on the potential effects of the clearance standards on the markets for scrap metal and
 disposal capacity. Finally, we discuss issues related to possible end uses of the materials, and industry
 and public attitudes towards recycling.
Economic Framework
                                                                        ?
       Economists generally assess market effects by considering the relationship between supply
and demand, illustrated in its simplest form in Exhibit 6-1. As indicated by the exhibit, demand for
a good or service can be represented by a downward sloping curve indicating that, as the price falls,
more of a good is generally demanded. Suppliers, on the other hand, are generally willing to provide
more of a good when the price rises, leading to an upward sloping supply curve. The intersection
of the supply and demand curves indicates the equilibrium market price.

       The exhibit also  illustrates  the effects of a change in supply due to changes in market
conditions. An increase in supply can be'represented by an outward shift of the supply curve from
S0S0 to SjSj. In this case, the price  drops (from P0 to PJ because more  of the good or service is
available. The drop in price leads to an increase hi quantity demanded (from Q0 to QJ. The change
in price depends on the slope (or elasticity) of the demand curve.1   In  some cases, the quantity
demanded may be relatively insensitive to changes in price.

       The options we analyzed could have both supply and demand effects in the markets for the
affected metals and in the market for disposal capacity. They could increase the supply of scrap
metal, perhaps providing these materials at a lower cost than other sources. They could also affect
the demand'for disposal  capacity. We discuss these effects in more detail below.
  1 Economists measure the welfare effects (or "benefits") of these types of changes by calculating
the change in consumers' and producers' surplus.  This calculation- takes into account people's
willingness to trade-off among different goods.  A more detailed explanation of this' concept !is
available in:  Freeman, A. Myrick. The Measurement  of- Erivironmentat and Resource Values:
Theory and Methods.  Washington, DC: Resources for  the Future, 1993.

                                            6-3

-------
 CD
 O


£
L0
                             Exhibit 6-1



                 SIMPLE SUPPLY AND DEMAND CURVES
                                  Quantity

-------
                                                             Preliminary Draft: June 13, 1997
 Scrap Market Effects
       If EPA's clearance standards significantly affect the supply of scrap metal, they could affect
scrap prices and/or the demand  for scrap or virgin materials from domestic or foreign sources.
Exhibit 6-2 compares the estimated total available volumes of DOE and NRC scrap metal presented
in Chapter 2 (for the years 1998 through 2053) to U.S. production of old scrap and of all types of
scrap in 1993 alone. Note that while the exhibit likely understates the total quantities of DOE and
NRC metals potentially available for recycling (because of data limitations described in Chapter 2),
it may overstate the  actual amount recycled in response  to the EPA.standards. As indicated in
Chapter 4, some of these quantities will not be cleared for unconditional use (e.g., because they are
too costly to decontaminate) and some metals would be recycled even in the absence of the EPA
standards. Also, clearance is likely to occur over several years, with a fraction of the total quantity
entering the market each year.

       Based on the  quantity of DOE and NRC scrap available, the potential for market effects
appears to be limited. The total quantities of aluminum, copper, and carbon steel available in the
DOE/NRC  inventory equal less .than three percent of 1993 old scrap production in the U.S., and
two percent or less of total U.S. scrap production. These data suggest that EPA's rulemaking would
likely have a small impact on scrap market prices for aluminum, copper, and carbon steel, even if
all of the available DOE and NRC scrap quantities were recycled in a single year. In contrast, the
affected DOE/NRC inventory represents 12 percent of the volume of stainless steel recycled in 1993,
and  75 percent of the  volume of nickel recycled.  As a  result, the potential impacts of EPA's
proposed rule on the  scrap markets for stainless steel and nickel are uncertain. If a large percent
of the total  available volumes were released in one year, significant market effects could result.

       We expect that a change of one percent or less in scrap supply will not affect overall market
prices significantly, even if some of the recycled materials are supplied a,t lower than market prices.
We also do not expect  that a one  percent change  in scrap supply  would significantly affect the
demand for  scrap imports or exports.2 We may investigate these potential effects further in future
analyses, particularly for stainless steel and nickel.
  2 International trade may also be affected if the EPA standards are significantly more lenient than
standards applied by U.S. trading partners. Other countries may be unwilling to accept U.S. scrap
that contains higher levels of radioactivity than allowed by their domestic standards.

                                        -    6-5

-------
Exhibit 6-2
COMPARISON OF TOTAL DOE AND NRC AFEECTED
SCRAP METAL QUANTITIES TO ANNUAL U.S. SCRAP RECYCLING
(thousands of metric tons) '
Metal.
Aluminum
Copper
Nickel
Carbon Steel
1
Stainless
" Steel ^
Total Estimated
DOE and NRC
Affected Metal
Quantities
(1998-2053)
(1)
36.9
12.4
44.9
1,311.0
146.4
Sources: Non-ferrous metals:
Plunkert, and Geral
• I" U.S. Bureau of Mir
* scrap resulting fron
resulting from consi
ferrous and nonferr
Steel: Houck, Gera
! Bureau of Mines, 1
outside sources. "Al
scrap.
Annual
U.S. Recycling
from Old Scrap
(1993)
(2)
1,798.0
612.0
N/A
46,147.0
642.0
Annual
U.S. Recycling
from All Scrap
(1993)
(3)
3,244.0
1,417.0
60.0
66,637.0
1,180.0
DOE and NRC
Scrap as a
Percentage of Old
Scrap Recycled
(l)/(2)=(4)
2.1%
2.0%
N/A
2.8%
22.8%
DOE and NRC
Scrap as a
Percentage of All
Scrap Recycled
(1)/(3)=(S)
1,1%
0.9%
74.8%
2.0%
12.4%
Carlin, James P., Daniel Edelstein, Stephen M. Jasinski, John F. Papp, Patricia A.
d Smith. Recycling - Non-Ferrous Metals: 1993 Annual Report. Washington, DC:
ics, 1995, Table 1, page 11. "Old scrap" includes metal recov
i consumer products; "all scrap" includes metal recovered f
nmer products or manufacturing processes. Figures for nicke
DUS sources.
Id W. Recycling Iron and Steel Scrap: 1993 Annual Report. \
995, pages 6 and 7. "Old scrap" includes receipts from broke
! scrap" includes old scrap plus receipts from other own comp
ered from purchased
rom purchased scrap
1 include nickel from
Vashington, DC: U.S.
rs, dealers, and other
>any plants and home

-------
                                                              Preliminary Draft: June 13, 1997
Effects on Disposal
       If the EPA clearance standards lead to a large change hi the amount of radioactive waste
disposal  (and perhaps storage) capacity needed, these standards could have a significant effect on
the costs of this capacity over time. For example, if the standards decrease the demand for existing
disposal capacity, the fixed costs associated with building that capacity will be spread over a smaller
quantity of wastes,  thus increasing the average costs of disposal. If the standards avert the need to
develop  new disposal sites in  the future, the full costs of developing the sites can be averted.
Disposal cost savings are taken into account hi our analysis of cost savings in Chapter 4; it does not
appear that the quantity recycled is significant enough to affect current estimates of these costs. We
intend to refine this conclusion based on future analyses of disposal costs and related capacities.  •

       If recycling averts the need to develop new disposal sites (or allows areas to be used for
purposes other than interim storage), the EPA standards may have a variety of additional benefits.
For example, they could avert decreases in the value of surrounding properties, allow the continued
use of the area for recreational purposes, or preserve values not associated with the direct  use
of the land (such as the value of simply knowing that fewer disposal sites exist). Again, however,
it  does not  appear that these potential  benefits would be significant, given  the magnitude of
estimated changes in quantities recycled under each of the policy alternatives.
Industry, and Public Concerns

       Both industry and public concerns about radioactive materials may limit the acceptance of
materials cleared for unconditional use and hence influence the effects of EPA's standards. The
extent to which these concerns will affect the quantities of materials recycled will depend on whether
the actual or perceived risk levels resulting from the proposed  standards are below the levels of
concern for the affected groups. These groups can be divided into three categories.

       •      Sensitive  industries:     The   computer,   electronics,   photographic,
              instrumentation, and other industries are concerned with the adverse impacts
              of radioactivity  on their  products. These industries may be reluctant  to
              purchase metals from scrap dealers or mills without some guarantee that the
              metals contain negligible levels of radioactivity.

       •      Scrap dealers and mills:  Industry may be reluctant to accept scrap metal
              from nuclear facilities even if the levels of radioactivity are extremely low, as
              long as scrap is available from other sources at a reasonable price. Under the
              existing release criteria, however, a number of scrap dealers already accept
              scrap metal from nuclear facilities.
                            . _V-'  '.  '.-'-_ -I 'V   .•-•;,••     '   L-  •    1
   -  •  •   r  General  public:  ' The public  nas 'expressed concern  about exposure  to
              radionuclides either in  finished  products  or  at  the plants  used  to
            '  decontaminate or 'recycle"the materials.   '>'-
                                             6-7

-------
                                                             Preliminary Draft: June 13, 1997


       Industry concerns could significantly affect the amount of materials recycled if the sensitive
sectors (i.e., industries sensitive to radiation, including risk averse scrap dealers and mills) are a large
share of the total market. The general public could also exert enough political pressure on individual
nuclear facilities or facilities receiving radioactive materials to discourage recycling. Information from
our site interviews suggests that many managers at nuclear facilities are concerned about negative
public perception and liability issues resulting from the release of material from their sites, and may
limit  the  release  of this material regardless  of the direct economic  incentives under alternate
clearance standards. In the current analysis, we assume that site managers are cost minimizers and
will choose the least-cost disposition alternative; we do not consider the potential effects of public
opinion and other non-economic factors. We may investigate this issue in future analyses to
determine whether the results of the economic model need to be scaled in some manner to account
for these concerns.
OTHER IMPACTS ON HUMAN HEALTH AND THE ENVIRONMENT

       EPA's rulemaking may have impacts on human health or the environment beyond the human
cancer effects analyzed in Chapter 5. These  impacts may include: (1) non-cancer human  health
effects from radiation; (2)  human health and ecological effects from releases of conventional or
toxic pollutants; (3) ecological effects from radiation; and (4) human health effects from accidents.
We discuss these issues and their potential impacts below.


Non-Cancer Human Health Effects from Radiation

       Human exposure to ionizing radiation can  result  in a variety of adverse health effects.
Therefore, the potential health risks associated with the proposed clearance standards are an
important factor in the cost-benefit analysis. EPA's Risk Assessment Guidance for  Superfund
Volume  I; Human Health Evaluation Manual (Part A) provides guidance for Superfund sites
contaminated with radioactive  substances. Chapter 10 of this document considers carcinogenic,
mutagenic, and teratogenic effects from ionizing radiation. For assessment  purposes, EPA states
that, "the risk of cancer is limiting and may be used as the sole basis for assessing the radiation-
related human health risks of a site contaminated with radionuclides."3 The basis for this conclusion
is discussed in more detail below.

       Carcinogenic, mutagenic, and teratogenic effects each have an associated risk factor. Risk
factors refer to the number of effects (e.g., the number of additional cancers or birth defects) per
unit of radiation dose. Exhibit  6-3 presents EPA's teratogenic, mutagenic, and cancer risk factor
values for low linear energy transfer (LET) radiation, the type of radiation that would be generated
by the radioactive materials of interest in this  analysis. The values listed under Risk Factor Range
  3 U.S. Environmental Protection Agency.  Risk Assessment Guidance for Superfund Volume I:
Human Health Evaluation Manual (Part AV  December 1989, page 10-32.

                                            6-8

-------
                                                              Preliminary Draft: June 13, 1997
represent a probability  of incidence  per unit  dose,  or "potency"  of the effect. For  example,
cumulative exposure to 1 Gy of radiation is expected  to result in 0.019 to 0.19 additional cancer
cases in the exposed population.4
Exhibit 6-3
SUMMARY OF EPA'S RADIATION RISK FACTORS8
(all factors for low LET radiation)
Risk
Teratogenic:b
Severe mental
retardation
Genetic:
Severe hereditary
defects, all generations
Carcinogenic:
Fatal cancers
All cancers
Significant Exposure Period
Weeks 8 to 15 of gestation
30-year reproductive generation
Lifetime
In utero
Lifetime
Risk Factor Range
(effect per Gy)
0.25-0.55
0.006-0.11
0.012-0.12
0.029-0.10
0.019-0.19
a In addition to the stochastic risks indicated, acute toxicity may occur at a mean lethal dose of
3-5 Sievert (SV) (300,000 - 500,000 mrem) with a threshold in excess of 1 Sv (100,000 mrem).
b The range assumes a linear, non-threshold dose-response. However, it is plausible that a
threshold may exist for this effect.
Source: Modified from Exhibit 10-5 hi U.S. Environmental Protection Agency. Risk Assessment
Guidance for Superfund Volume I: Human Health Evaluation Manual (Part A). December
1989, page 10-31.
                  "." n     - lit IK
  4 Gray (Gy) units represent levels of absorbed dose, or the energy generated by ionizing radiation.

                                             6-9

-------
                                                             Preliminary Draft: June 13, 1997


        Data in Exhibit  6-3 show that  low LET radiation's  mutagenic potency is similar to  its
carcinogenic potency for these effects' significant exposure periods. The risk of birth  defects such
as mental retardation, however,  can be much higher than that of cancer effects per unit dose.
Specifically, the data suggest that a population exposed to 1 Gy of low LET radiation is likely to
show a significantly higher number of teratogenic effects than cancer effects if the dose is received
during weeks eight to 15 of pregnancy.

        Although the risk factor per unit exposure for teratogenic effects is significantly  greater than
the risk factors for cancer and mutagenic effects, the induction of teratogenic effects primarily occurs
from exposure during weeks eight to 15 of gestation. If the  exposure does not occur within the
specified time frame, severe teratogenic effects are unlikely. Due to this severely limited period of
exposure for teratogenic  effects and the potentially longer period of exposure for cancer effects, the
cumulative incidence of cancer will be significantly greater than the incidence of teratogenic effects
for a population exposed to a given  source of radiation over  a long period of time. Similarly, the
significant exposure period for genetic effects is only 30 years, which represents the time period in
which women are capable of reproduction. For any  given  source  of  radiation  to  the  general
population, the cumulative incidence of cancer will be greater than the incidence of genetic effects
due to the susceptibility of both sexes and the potentially longer period of exposure for cancer.

       The fact that the incidence of cancer will be significantly greater than that of teratogenic and
mutagenic effects in a population exposed to a given long-term source of low LET radiation
underlies  EPA's guidance that the risk  of cancer is limiting and can be used to assess radiation-
related human health risks. We  adopt  this convention  for the preliminary cost-benefit analysis;
however,  we may investigate  in  future analyses  whether any of  the recycling practices or end,
products might  inordinately expose pregnant women. If such exposure  appears possible, we will
employ the proper risk considerations to calculate possible teratogenic effects.


Additional Reported Risk Factors

       To ensure that the risk factors documented by EPA in the Risk Assessment Guidance for
Superfund Volume I; Human  Health Evaluation Manual  (Part A) are up to date, we consulted the
Committee on the Biological Effects of Ionizing Radiations (BEER V) report from  1990, Health
Effects of Exposure  to Low Levels of Ionizing Radiation, and the National  Council on Radiation
Protection and  Measurements (NCRP) report from 1993, Limitation  of Exposure to Toni7ing
Radiation. Exhibit 6-4 presents a summary of the risk factors reported by both of these documents.
All values fall within the ranges of factors estimated by EPA.
                                            6-10

-------
                                                              Preliminary Draft: June 13, 1997
Exhibit 6-4
COMPARISON OF RISK FACTORS
(effect per Gy of low LET radiation)
Risk
Teratogenic:
Severe mental retardation
Genetic:
Severe heredity defects, all generations
Somatic:
Fatal cancers
All cancers
Sources:
1. U.S. Environmental Protection Age
Human Health Evaluation Manual
EPA1
0.25 - 0.55 (weeks 8 to 15 of
gestation)
0.006 - 0.11 (30-yr reproductive
generation)
0.012 - 0.12 (Lifetime)
0.029 - 0.10 (In utero)
0.019 - 0.19 (Lifetime) ,
BEIRV2
0.4"
Not
reported
Not
reported
Not
reported
0.04**
NCRP3
0.4'
0.013C
0.05
0.02-
0.03
0.06
ncv. Risk Assessment Guidance for Superfund Volume I:
(Part AV December 1989, page 10-31.
2. Committee on the Biological Effects of Ionizing Radiations, Board on R
Research, Commission on Life Sciences, National Research Council. He
Exposure to Low Levels of Ionizing Radiation CBEIR V). National Acac
Washington, DC, 1990, pages 6,7.
3. National Council on Radiation Prot
lonizingjiadiation. Bethesda. MD.
Notes:
a. Risk factor adjusted for 1 Gy expos
b. Risk reduced by the suggested factc
c. Lifetime value
ection and Measurements. Limitatior
March 1993, pages 3,30,37,38.
ure instead of 0.1 Gy.
ir of 2 to reflect lower dose rate (BE
adiation Effects
ilth Effects of t
emy Press,
i of Exposure to
[R V, page 6).
Other Health Effects from Radiation

        We reviewed the most recent literature available on the human health effects of ionizing
radiation  to  ensure that other potential 'health risks resulting from recycling practices are not
omitted from our analyses. Exhibit 6-5 summarizes two additional health effects from radiation and
jthe corresponding range of threshold doses required to induce such effects. Due  to the variability
of dose rates- and biological characteristics of individuals, fhese threshold, values do not guarantee
a zero probability of incidence, but reflect ajange^of. dose, JeveJs typically required to induce such
effects for the average individual.   ,             (i"      lf      - ,,••  -            '  .   '
                                             6-11

-------
                                                        Preliminary Draft:  June 13, 1997
                                   Exhibit 6-5
          OTHER EFFECTS OF EXPOSURE TO IONIZING RADIATION
             Effect
                              Threshold Reported
   Threshold
    in mrera
Cataracts1
                          0.6-1.5 Gy (single exposure)
                           8 Sv (protracted exposure)
60,000 - ISO^OCKf
    800,000"
Sterility2
        threshold dose for
        temporary sterility in
        adult human testis

        threshold dose for
        permanent sterility in
        adult human testis

        threshold dose for
        permanent sterility in
        adult ovary
                                    0.15 Sv
                            3.5 Sv (single exposure)
                          2.5 - 6.0 Sv (single exposure)
                          6.0 Sv (protracted exposures)
     15,000"
    350,000b
250,000 - 600,000*
     600,000
Sources:
1.
2.

Notes:
a.
b.
Committee on the Biological Effects of Ionizing Radiations, Board of
Radiation Effects Research, Commission on Life Sciences, National Research
Council. Health Effects of Exposure to Low Levels of Ionizing Radiation
(BEIRV). National Academy Press, Washington, B.C., 1990, page 363.
BEIR V, pages 364-366.
Assuming only beta particles and gamma rays
      .  1 Gy = 100 rad
        rem = rad x RBE (Relative Biological Effectiveness)
        EPA assumes RBE = 1 for beta particles and gamma rays
        Therefore, rad = rem, and
        mrem = rad x 1,000
1 Sv = 100 rem
100 rem »  100,000 mrem
                                     6-12

-------
                                                             Preliminary Draft: June 13, 1997


       The reported threshold values have been converted to the mrem unit to facilitate their
 comparison to each other and to the policy alternatives tinder consideration. All threshold values
 in Exhibit 6-5 exceed the highest clearance standard considered in our analysis (15 mrem) by factors
 of one thousand or greater. Therefore, our health risk analysis does not consider the impact of
 EPA's rulemaking 9n the incidence of cataracts and sterility.5     *
 Human Health and Ecological Effects from
 Releases of Conventional and Toxic Pollutants

       Both baseline and  post-regulatory practices  can result in the release of conventional
 pollutants and toxic substances. For example, storage or disposal practices may allow contaminants
 to leach into groundwater, and air and water pollutants will be emitted by mills involved in recycling
 metals. If current practices do not include adequate measures to protect against conventional and
 toxic releases, changes in the flow of scrap metal to particular disposition alternatives could result
 in associated effects on human health. We may further investigate the potential significance of these
 risks in future analyses. It is likely, however, that any changes in risks associated with conventional
 and toxic pollutants will be minimal.
Ecological Effects from Radiation

    ,   Releasing radionuclides into the environment can result in a variety of adverse effects on
ecological receptors. A common approach in ecological risk assessment, however, is to assume that
measures taken to protect human health from exposure to ionizing radiation are also sufficient to
protect the environment. According to the 1977 Recommendations of the International Commission
•on Radiological Protection (ICRP):

       Although the principal objective of radiation  protection is the achievement and
       maintenance of appropriately safe conditions for activities involving human exposure,
       the level of safety  required for the protection of all human individuals is thought to
       be adequate to protect other species, although  not necessarily individual members
       of those species. The Commission therefore believes that if man is adequately
       protected then other living things are also likely to be sufficiently protected.

ICRP modified this statement in 1990 to include:

       The Commission  believes that the standard of environmental control  needed to
       protect man to the degree currently thought desirable will ensure that other species
  5 The fact that the incidence of cancer will be significantly greater than that of teratogenic and
mutagenic effects in a population exposed to a given long-term source of LET radiation underlies
EPA's guidance that the risk of cancer is limiting and can be used to assess radiation-related human
health risks.

                                            6-13

-------
                                                            Preliminary Draft: June 13, 1997


       are not put at risk. Occasionally, individual members of non-human species might be
       'harmed, but not to the extent of endangering whole species or creating imbalance
       between species.6

       Two recent documents on the ecological damage associated with ionizing radiation again
 evaluate the validity of these statements. We reviewed these recent documents to determine whether
 the potential range of clearance standards is likely to protect ecological receptors. Both documents
 report a threshold dose below which ecological damage at the population level is unlikely.

       The National Council on Radiation Protection and Measurements' (NCRP) 1991 report,
 Effects of Ionizing Radiation on Aquatic Organisms, concludes that a chronic dose rate of 0.4 mGy
 per hour and lower to the maximally exposed individual in a population of aquatic organisms is
 unlikely  to induce ecological damage at the population level.7  In addition, NCRP confirms  the
 notion that the  protection of human health against  radiation ensures the protection  of  the
 environment:  "If man is protected by limiting the exposure via the aquatic pathways to 1 mSv/year,
 populations of aquatic organisms in this environment should also be protected from  deleterious
 effects of radiation."8

       The International  Atomic  Energy Agency's (IAEA)  1992 report,  Effects of Ionizing
 Radiation on  Plants and Animals at Levels Implied by Current Radiation Protection Standards,
 concludes that chronic radiation doses below 1 mGy per day are unlikely to harm terrestrial and
 freshwater plant and animal populations.9   IAEA also confirms the NCRP assumption regarding
 human health as the limiting factor  for radiation protection measures:

       It is highly probable that limitation of the exposure of the most exposed humans (the
       critical human group), living  on and receiving full sustenance from the local area, to
       1 mSv/year will lead to dose rates to plants and animals in the same area of less than
       1 mGy/day. Therefore, specific radiation protection standards for non-human biota
       are not heeded.10

       The thresholds reported  by  these two documents are significantly higher than the levels
presented as analytic options in this  analysis. Due to the relatively high exposure levels required to
induce ecological damage, recycling  practices are highly unlikely to harm* species at the population
level. The relationship between the reported threshold levels and the potential range of clearance
levels is discussed in more detail below.
  6 Both ICRP quotes are from International Atomic Energy Agency. Effects of Ionizing Radiation
on Plants and Animals at Levels Implied bv Current Radiation Protection Standards. Vienna, 1992,
page 1.

  7 National Council on Radiation Protection and Measurements. Effects of Ionizing Radiation on
Aquatic Organisms. Bethesda, MD, August 1991, page 62.

  8 NCRP, page 60.

  9 IAEA, page 54.

  10 IAEA, page 54.

                                           6-14

-------
                                                              Preliminary Draft: June 13, 1997


       A dose limit of 1 mGy per clay is the more conservative  threshold of the  two mentioned
above, and thus will be used in the following example. Using this reported threshold, a dose level
of  100 mrem/day  is required to induce  ecological damage at  the  population  level.11  A  15
mrem/year limit for human exposure translates to an average dose of 0.041 mrem/day. These data
suggest that the threshold below which plants and animals are  unlikely to  manifest ecological
damage is greater than the potential range of clearance levels by a factor of 103. In other words,
ecological receptors would require a dose that is  approximately 2,400  times  greater than  the
maximum human daily dose allowed under a 15 mrem/year clearance  standard in order to manifest
ecological damage at the population level. This substantial difference between human and ecological
receptor exposure levels is unlikely.

       The calculations above support the assumption that measures taken to protect human health
from exposure to ionizing radiation are likely to sufficiently protect ecological receptors  as well.
However, in future analyses, we may continue to review whether any recycling practices or end uses
of products might inordinately expose threatened or endangered species, species with low fecundity,
species subject to significant additional natural environmental stressors, species at high trophic levels
(due to the potential for radionuclides to  biomagnify), or  species in close proximity to radiation
sources.
Transportation Risks

       EPA's standards may also affect human health by causing incremental changes in the number
of transportation  accidents  relative  to baseline  conditions. Radioactive scrap  metal is  often
transported to  decontamination or melting facilities for processing before release  or burial.  In
addition, scrap  may be released directly frpm sites in instances when activity levels are below the
clearance standards or when material is packaged  and shipped directly to burial facilities. In each
of these cases, transportation workers and the general public are exposed to risk from traffic or rail
accidents.

       EPA's standards may affect the number of related accidents and fatalities under the following
conditions:  (1) if the standards lead  to changes in the flow of scrap metal to various disposition
alternatives relative to baseline conditions; and (2)  the  level of  transportation  risk varies across
disposition alternatives. For example, an increase in recycling activity attributable to EPA's standards
(as indicated by the results presented in Chapter 4) would  lead to a  corresponding increase in
transportation fatalities only if the transportation risk attributable  to the recycling alternative were
  11 Assuming gamma rays or beta particles and thus SB RBE-of-1:

              1 Gy * 180 rad
              1 mGy = 0.1 rad
              0.1 rad = 0.1 rem
              0.1 rem = 100 mrem

                                            6-15

-------
                                                              Preliminary Draft: June 13, 1997


greater than for burial.   Based on a limited and preliminary review of the available evidence, it
appears unlikely  that  changes  in  recycling activity will  lead  to any  significant  increase in
transportation risks due to the fact that these risks are similar for both recycling and burial options.

        This preliminary conclusion is based largely on an analysis developed by DOE to assess the
transportation risk related to the agency's restricted recycling initiative.12 This analysis suggests that
per unit transportation risks (i.e., accidents or fatalities per mile driven) are  identical for recycling
and burial options. For example,  the estimated risks for shipping contaminated carbon steel in B-25
type boxes to disposal sites and shipping the same material in SeaLand-type containers to melting
facilities both equal S.lxlO"8 traffic fatalities per shipment mile.13   We assume that these types of
shipments are similar to those that would occur under unconditional clearance scenarios.

        As a result, recycling alternatives would  increase risk only if such options involved greater
shipping distances relative to  burial; however, additional DOE analysis suggests that  shipping
distances are likely to be similar under both recycling and burial options. In its recycling cost model,
developed to assess the relative costs of restricted recycling alternatives versus burial, DOE employs
a mileage estimate of 474 miles for scrap metal going directly to a disposal site, while the Agency's
related estimate for scrap metal going to one of two regional melting facilities equals 482 miles.14
Our own preliminary review of the individual  sites' disposal options and locations (discussed in
Chapter 3) suggests that average shipping distances may be shorter for recycling options compared
to burial, meaning that increases  in material released for unconditional use may lead to decreases
in transportation  risk.15  We  may  investigate this  issue further  in  future  analyses, including
expanding our literature search to identify additional per unit risk factors and conducting additional
analyses of relative shipping distances for each  disposition option.
  12 S.Y. Chen, S. Folga, L.A. Nieves, and J. Arnish.  Assessment of Risks and Costs Associated with
Transportation of DOE Radioactive Scrap Metals (Draft). Prepared for the U.S. Department of
Energy, July 1995.

  13 S.Y. Chen et aL, July 1995, page 13.

  14 Stephen Warren, R. Scott Moore, Robert E. Gant, and Kathleen Robertson. Cost Model for
DOE Radioactivelv Contaminated Carbon Steel Recycling. December 1995, page 14.

  15 For example, over 75 percent of available scrap metal within the DOB complex is held at  the
Oak Ridge, Paducah, and Portsmouth gaseous diffusion plants.  Each of these sites will likely have
to dispose of its material off-site (likely at the Nevada Test Site).  In contrast, the primary melting
and decontamination facilities currently operating are located in the same region as these facilities.

                                            6-16

-------
                                                           Preliminary Draft: June 13, 1997


ENVIRONMENTAL IMPACTS OF REDUCING
DEMAND FOR VIRGIN MATERIALS
     I
       Increased  recycling of scrap metals from nuclear facilities could reduce the need for virgin
material production.16 Reduced virgin materials production would avoid the environmental impacts
associated with mining, benefication, and processing of virgin materials - including both direct
emissions and wastes generated by these processes and the environmental impacts of energy use at
each  stage of production. The net environmental impacts of recycling versus virgin materials
production vary with the specific metal recovered.

       Recycling  has the potential to achieve substantial life cycle environmental benefits, with the
major benefits resulting from avoiding the environmental impacts of mining and the energy-related
emissions of primary metal processing. Past studies comparing life cycle impacts of recycling versus
virgin production have been performed for some metals, and partial estimates are available for other
metals. The most extensive studies have been performed for steel and aluminum used in packaging: ,

       •      A  1992 Tellus  study estimated the life cycle energy and environmental
              impacts of different packaging materials, including steel and aluminum cans,
              from recycled and primary production.17

       •      A 1976 study by Calspan estimated the environmental impacts of producing
              steel and  aluminum from various primary  and secondary processes and
              different types  of scrap.18   The Calspan  study covered more  recycling
              processes  and  types of  scrap than the  Tellus  study,  although some
              assumptions about pollution controls may be out of date.

The results of these studies for steel and  aluminum scrap recycling are summarized below.
    16 Increased recycling of scrap from nuclear facilities might replace virgin materials or might
displace recycling of scrap from other sources. This discussion focuses on the environmental effects
of reducing production of a ton of steel or a ton of aluminum, but does not attempt to predict how
much virgin production will actually be  avoided by  increased recycling of scrap  from nuclear
facilities.

    17 Tellus  Institute,  CSG/Tellus Packaging  Study, prepared  for The Council  of State
Governments, the U.S. EPA, and the NJ. DEPE, 1992 (five reports).

    18 Calspan Corp., Environmental Impacts of Virgin and Recycled Steel and Aluminum, prepared
for  U.S. EPA, EPA/530/SW-117c, NTIS PB-253 487, 1976.

                                           6-17

-------
                                                              Preliminary Draft:  June 13, 1997
Impacts of Recycling Steel Scrap
        The Tellus study estimates percentage changes in pollutant loadings attributable to increased
use of scrap (from 28 to 40 percent of total input) in place of pig iron in a steelmaking  facility's
basic oxygen furnace (EOF).19 We present the study's findings in Exhibit 6-6. As these estimates
show, reductions in pollutant releases can be substantial, even under the limited recycling scenarios
considered by Tellus. The only exception is the potential for increased air emissions if detinning or
other decontamination of the scrap is required. Tellus estimates a reduction in the discharge of
pollutants to all environmental media, except for a potential increase in SOx emissions if detinning
is required and an increase in hydrocarbon emissions in the "controlled  emissions/with detinning"
case.
Exhibit 6-6
ENVIRONMENTAL IMPACTS OF RECYCLING STEEL
Category
Criteria Air Pollutants
with detinning of scrap
without detinning
Non-criteria air pollutants (toxics)
with detinning of scrap
without detinning
Conventional water pollutants
Nonconventional (toxic) water
pollutants
Percent Reduction from Use of 12%
Additional Scrap in BOF
(Per Ton of Steel Produced)
Uncontrolled

N/A
15%

N/A
14%
12%
16%
Controlled

slight increase1
14%

12%
13%
14%
15%
1 Detinning results in substantial increase in SOx emissions.
Source: Tellus 1992, Tables S-l and S-2.
  19 These percentages are calculated by summing pounds of specific pollutants reduced per ton of
steel produced, as estimated by Tellus. The Tellus study shows different reductions for different
individual pollutants.
                                            6-18

-------
                                                             Preliminary Draft: June 13, 1997
       Exhibit 6-7 shows the Calspau estimates for percent reductions from recycling, per ton of
steel produced, for two types of steel scrap (auto and steel can) and two steel production methods
(EOF and electric arc furnace).20  These estimates also  show significant environmental benefits
from recycling, especially in the electric arc furnace (EAF), which allows a larger scrap input than
BOF production. Most of the emissions differences estimated by Calspan result from differences in
energy use (in extraction and processing, assuming the combustion of fossil fuels.)  The overall
estimates of reductions in  air emissions reflect reductions hi participates, sulfur oxide emissions
(except for an increase for  recycling of steel can scrap in an EAF), and ammonia, and increases in
CO, hydrocarbons, and organics.
. Exhibit 6-7
ENVIRONMENTAL IMPACTS OF RECYCLING STEEL
Category
Water Discharges
Air Emissions
Water Pollution
Solid Wastes
Energy Consumption
Percent Reduction with Recycling
(Per Ton of Steel Produced)
10% Auto Scrap
in BOF
13%
8%
14%
15%
10%
10% Can Scrap
in BOF
13%
8%
13%
15%
10%
70% Auto
Scrap in EAF
99%
29%
76%
96%
55%
70% Can
Scrap in EAF
98%
19%
67%
92%
41%
Source: Calspan, Table S-3
       Finally, the Institute of Scrap Recycling Industries (ISRI) estimates the following effects of
using scrap iron and steel instead of virgin ore to make new steel:21

       •      74% savings hi energy,

       •      90% savings in virgin materials used;
  20 Calspan calctdated peftentage increases 6r decreases m releases of specific pollutants per metric
ton of carbon steel. We summarize their results by summing the changes across pollutants and
calculating overall percentage changes.
                                                                               \
  21 Institute of Scrap  Recycling Industries (ISRI), Recycling Scrap Iron and Steel and Recycling
Nonferrous Scrap Metals. 1993.
                                            6-19

-------
                                                            Preliminary Draft:  June 13, 1997
              86% reduction in air pollution;

              40% reduction in water use;

              76% reduction in water pollution; and

              97% reduction in mining waste.
Impacts of Recycling Aluminum Scrap

       Exhibit 6-8 shows the Tellus study estimates of the reductions in environmental releases from
recycling of aluminum beverage can scrap. These percentages reflect reductions for all pollutants
except a potential increase in TSP emissions if demagging is required, and increases in wastewater
discharges of chloride, sodium, aluminum and zinc. Substantial reductions in wastewater discharges
were estimated for a long list of other toxic organic and metal pollutants.  In addition, the Tellus
study notes that demagging would result in increases in emissions of noxious halogen and halogen-
compounds and particulates; estimates of emission factors for these compounds were not available,
however.
Exhibit 6-8
ENVIRONMENTAL IMPACTS OF RECYCLING ALUMINUM

Criteria air pollutants
with demagging of scrap1
without demagging
Non-criteria air pollutants
with demagging of scrap
without demagging
Conventional water pollutants
Nonconventional (toxic) water
Percent Reduction with Recycling
(Per Ton of Aluminum Produced)
Uncontrolled

N/A
100%

N/A
> 99%
99%
88%
Controlled

82%
99%

> 99%
> 99%
N/A
N/A
1 Demagging results in substantial increase in TSP emissions.
Source: Tellus 1992, Tables A-l and A-2.
                                           6-20

-------
                                                            Preliminary Draft: June 13, 1997
       Calspan reports overall percentage reductions in pollutant releases for recycling of three
types of aluminum scrap (municipal solid waste, shredder, and baler) in producing two types of ingot
(wrought or cast). These findings are summarized in Exhibit 6-9. Significant reductions in emissions
were estimated for most pollutants.  The exceptions were increases in chloride air emissions and
increases in wastewater discharges of cadmium, lead, manganese and fluoride.
Exhibit 6-9
ENVIRONMENTAL IMPACTS OF RECYCLING ALUMINUM

Water discharges
Air pollutant emissions
Water pollutant loadings
Solid wastes2
Energy consumption
Reduction in Pollutant Release
(Per Ton of Aluminum Produced)
Municipal Solid Waste
(MSW) ito Wrought Ingot
97%
96%
95%
> 100%
97%
MSW to
Cast Ingot,
97%
92%
Increase1
'99%
95%
Shredder
Scrap to Cast
Ingot
97%
95%
78%
> 100%
97%
Baler Scrap
to Cast
Ingot
97%
95%
77%
> 100%
97%
1 Substantial increase in suspended solids from processing MSW.
2 96-99% reduction in overburden and process wastes plus reduction in post-consumer scrap.
Source: Calspan, Table A-3
Other Estimates of Impacts

       In addition to these comprehensive studies, various sources provided partial estimates of the
environmental impacts of resource extraction and mineral processing that would be avoided by
recycling.
Energy Use

       Primary and secondary production of metals differ in the amount of process energy used.
Recycling generally requires less use of energy than primary production. This reduces the demand
for energy raw materials (oil, coal, uranium). As summarized in Exhibit 6-10, a number of studies
                                           6-21

-------
                                                            Preliminary Draft:  June 13, 1997
have estimated the overall percentage reduction in energy requirements from recycling.22, These
reductions in  energy use conserve resources and avoid the generation of pollutants that would
otherwise be released. For example, the combustion of coal to generate electricity produces carbon
dioxide and other greenhouse gases, and sulfur oxides and nitrogen oxides which are harmful to
human health and contribute to acid rain. In addition, coal combustion releases naturally occurring
radioactive materials (uranium and thorium).23
Exhibit 6-10
REDUCTION IN ENERGY USE FROM RECYCLING
Metal
Steel
Aluminum
Copper
Nickel
Lead
ORNL (1972)
(% reduction)
72%
96%
87%
NA
60%
ASM Metals
Handbook
(% reduction)
58%
96%
84%
89%
64%
ISRI
(% reduction)
NA
95%
85%
NA
65%
Denison (reduction --
mill Btus/ton material)
7.8 - 19.3
129-221
NA
NA
NA
Water Pollution

       Stormwater runoff from mining sites and acid mine drainage can pollute streams and rivers
with acids, heavy metals and sediments. Estimates of the quantities of wastewater generated by
mining range widely due to differences in mining operations. Estimates for  copper mining, for
example, range from 0 to 300 liters per ton of ore in open-pit mines and from 8 to 4,000 liters per
ton in underground mines. Based on a yield of 1  ton of metal per 99 tons of ore, these estimates
translate into upper-bound estimates of 29,700 liters of wastewater per ton of metal recovered for
open-pit mines and 396,000 liters per ton for underground mines.24
  22 Oak Ridge National Laboratory, Energy Expenditures Associated With the Production  and
Recycle of Metals. J.C. Bravard, et al., ORNL-NSF-EO-24, November 1972; American Society for
Metals, Metals Handbook  (Desk Edition). 1985, p. 31-5; Richard  A. Denison, "Environmental
Life cycle Comparisons of Recycling, Landfilling and Incineration: A Review of Recent Studies,"
Annual Review of Energy and the Environment. Vol. 21, Annual Reviews, Inc.: Palo Alto CA, 1996;
Institute  of Scrap  Recycling Industries, Inc.  "Recycling Scrap Materials Contributes to a Better
Environment" (undated).

  23 Alex Gabbard, "Coal Combustion: Nuclear Resource or Danger," ORNL Review. 1996.

  24 LA. Nieves, S.Y. Chen, EJ. Kohout, B. Nabelssi, R.W. Tilbrook, and S.E. Wilson. Evaluation
of Radioactive Scrap Metal Recycling. Argonne National Laboratory, Environmental Assessment
                                           6-22

-------
                                                            Preliminary Draft: June 13, 1997
Mining Impacts
       Mining produces large amounts of solid waste, much of which may contain toxic metals.
Overburden is the rock and soil removed to expose an ore body and develop a mine. Tailings are
.produced in the benefication of ores to  improve their grade prior to smelting. The quantity of
tailings produced per ton of metal recovered depends on the grade of ore. Mining wastes are often
left in huge piles, which can continue to be a source of contamination for many years after a mine
is abandoned.
Caveats and Limitations

       Several factors must be considered when assessing the potential impacts of recycling scrap
on virgin materials demand:
                                                                                     /
       •      The net environmental benefits associated with scrap metal recycling depend
              on the recycling process used and on the extent of contamination present.
              For  example,  most  of the estimated  emissions from the recycling  of
              aluminum cans is  associated with removal of the excess magnesium present
              in can  lids.  If  scrap  from  nuclear  facilities  does  not contain  such
              contaminants,  emissions from aluminum recycling may be minimal.

       •      Our  characterization of overall changes in the emission or discharge of
              pollutants per  ton of metal produced combines multiple pollutants without
              considering  their  relative  impacts on human health or the environment.
              Additional analysis would be required to assess the net impacts of changes
              in pollutant  loadings.

       •      Many of the estimates of energy use relied on hi the  studies cited were
              developed in the  1970s. Changes in production to reduce energy use may
              have occurred  since then, potentially reducing the energy savings associated
              with recycling.  Even with changes in energy use, however, the energy savings
              from avoiding  virgin  material production — especially for aluminum — are
              likely to be substantial.

       •      These estimates do not consider the impacts of radiation decontamination
              processes, which may be required to meet the clearance standards and which
              may  have environmental impacts that offset the benefits attributable to
              reduced virgin material production.
                                   '..  i  -.'^  .- U'*iil <-i^,n. '  '' lfl  K?*1 l '»*.'  V ,' a' '.  l

                                   • ••  i   '   \   4:f   .        i-i^
Division, December 1995.

                                           6-23

-------
                                                            Preliminary Draft: June 13, 1997


              These  estimates do not consider the environmental impacts of potential
              differences in transportation distances for recycled versus virgin materials,
              and the associated impacts on fuel use, vehicle emissions, and accidental
              releases.

              Estimates of environmental releases for different processes may or may not
              take the  effects of pollution controls into account.  Estimates of absolute
              reductions in emissions may be overstated by data assuming no controls are
              in place. In addition, estimates  of the relative environmental impacts of
              recycling versus virgin materials production will be biased if emissions are
              typically controlled to a greater or lesser degree in one source than another
              and the available estimates do not  take these differences into account. We
              have not yet gathered  information on the  extent of controls for primary
              versus secondary metals production that would  help to address this issue.
Summary
       Available  estimates  show  that  recycling  can avoid substantial environmental  impacts
associated with resource extraction, metal processing, and associated energy use. Large reductions
in environmental impacts may result in particular from avoided mining impacts and reduced use of
energy in aluminum production. The net environmental impacts from recycling of scrap depend to
a large extent on how much decontamination is required -- for example, detinning of steel scrap or
demagging of aluminum scrap. Net impacts also  depend on the extent to which environmental
emissions or releases are currently controlled in primary and secondary metals production.
                       , iV
                                           6-24

-------
                                                             Preliminary Draft: June 13, 1997
SUMMARY AND CONCLUSIONS                                              CHAPTER 7
INTRODUCTION                                                         :

       This chapter summarizes the results of our preliminary assessment of the costs and benefits
of alternate standards for the release of scrap metal from DOE facilities and commercial nuclear
power reactors. It discusses the predicted change in scrap metal management costs, cancer incidence,
and other impacts under each of the three analytic options considered.

       The chapter contains two .sections. The first examines the potential incremental effects of
the analytic options relative to baseline conditions, discussing results for both DOE facilities and
nuclear power reactors.  The second discusses the potential implications of these results, as well as
the limitations of the analysis and potential topics for future research.
COMPARISON OF COSTS AND BENEFITS

       Exhibit 7-1 summarizes the results of the preliminary analysis, presenting our findings for
both the low and high disposal cost scenarios, assuming midpoint activity levels. The dollar figures
presented have been discounted to 1997 dollars using a real discount rate of seven percent. As
illustrated in this exhibit, the cost impacts of the three analytic options vary considerably, ranging
from an increase in scrap metal management costs of $0.2 to $0.5 billion at the 0.1 mrem standard
to savings of $1.4  to $1.7 billion at the  15.0 mrem standard. The difference in impacts reflects
differences in the disposition of scrap across options. Under current standards, the 0.1 mrem option,
and the 1.0 mrem option, most of the scrap metal must be decontaminated prior to release. Under
the 15.0 mrem option, most scrap metal can be released directly from the facilities of interest with
no decontamination, making unconditional clearance more cost-effective relative  to burial.

       The results of the collective impacts analysis indicate the change in number of cancer cases
predicted to occur over a 1,000 year period. Like the cost impacts, the predicted change in cancer
cases varies-considerably across the'three jl0pti
-------
                                                             Preliminary Draft: June 13, 1997
an estimated increase in cancer incidence of 19 to 29 cases.  Like the 0.1 mrem option, the 1.0 mrem
option is expected to decrease cancer incidence relative to current standards, yielding a decline of
six to 10 cancer cases, depending on the disposal cost scenario considered;  changes in cancer
incidence are smaller for the low end disposal cost scenario because less scrap enters free release.

       EPA's rulemaking may have additional impacts on  a variety of factors, including: (1) scrap
metal market prices and the demand for radioactive waste  disposal capacity; (2) non-carcinogenic
human health risks; (3) ecological impacts;  and (4) demand for virgin materials (e.g., iron ore).
Based on our preliminary review of these  issues, it  is likely that  these impacts are small and
insignificant compared to direct cost effects  and impacts on cancer risks. As a result, we have not
attempted to quantify these impacts or to differentiate their magnitude across the three analytic
options.1 We may revisit these issues in subsequent phases of this analysis.
Exhibit 7-1
POTENTIAL IMPACTS OF EPA CLEARANCE STANDARDS
Impact1
Change in Costs (present value;
billions of 1997 dollars)
Change in Predicted
Cancer Incidence2
Other Impacts3
Analytic Option
0.1 mrem
Low
Disposal
Costs
$0.20
(8.2)
High
Disposal
Costs
$0.47
(14.3)
uncertain
1.0 mrem
Low
Disposal
Costs
$0.00
(6.3)
High
Disposal
Costs
($0.02)
(10.0)
uncertain
15.0 mrem
Low
Disposal
Costs
($1.40)
19.2
High
Disposal
Costs
($1.65)
28.8
uncertain
Notes:
1 The values shown arc based upon initial levels of radioactivity at the logarithmic midpoint of the range
reported for each scrap metal item.
2 Number of total cases (fatal and non-fatal) predicted to occur over 1,000 years.
3 - Includes other economic impacts potentially attributable to the rulemaking such as effects on scrap
metal markets and noncancer human health and environmental effects. These impacts are likely to be
small, and insignificant relative to, impacts on scrap management costs and cancer incidence.
Totals in this table may not equal the sum of the totals for DOE and NRC due to rounding.
  1 This preliminary anatysis does not address several other potentially significant impacts, including
environmental justice  issues,  effects on small businesses, costs  imposed by unfunded  Federal
mandates, and the relationship of EPA's standards to other governmental programs. Assessments
of these potential impacts may be conducted over the next several months.

                                            7-2

-------
                                                             Preliminary Draft: June 13, 1997
DOE Facilities
       Exhibit 7-2  summarizes the potential impacts of EPA's rule for scrap metal from DOE
facilities. As indicated by the exhibit, the 0.1 and 1.0 mrem standards are predicted to have relatively
modest impacts on DOE scrap metal management costs, increasing costs by up to $60 million and
decreasing costs by  $40 million, respectively. Under these analytic options, most DOE scrap metal
(80 to 100 percent)  is predicted to flow to burial because it is the least-cost alternative. In contrast,
unconditional clearance is predicted to become the least costly alternative for 98 percent of DOE
scrap metal under the 15.0 mrem standard. As a result, this option generates estimated cost savings
of approximately $1.4 billion to $1.6 billion.

       With the  majority of DOE scrap metal predicted to flow to burial  under the 0.1 and 1.0
mrem standards,  the corresponding change in predicted cancer incidence is negligible. In contrast,
the change in DOE scrap management practices under the 15.0 mrem standard is predicted to cause
approximately one additional cancer case over the 1,000 year modeling period.
Exhibit 7-2
IMPACTS OF EPA CLEARANCE STANDARDS:
DOE FACILITIES
Impact1
Change in Costs (present value;
billions of 1997 dollars)
Change in Cancer Incidence2
Analytic Option
0.1 mrem
1 Low
Disposal
Costs
$0.03
negligible
High
Disposal
Costs
$0.06
negligible
1.0 mrem
Low
Disposal
Costs
($0.04)
negligible
High
Disposal
Costs
($0.04)
negligible
15.0 mrem
Low
Disposal
Costs
($1.36)
1.2
High
Disposal
Costs
($1.58)
1.2
Notes:
1 The values shown are based upon initial levels of radioactivity at the logarithmic midpoint of the range
reported for each scrap metal item.
2 Number of total cases (fatal and non-fatal) predicted to occur over 1,000 years.
                                            7-3

-------
                                                             Preliminary Draft:  June 13, 1997
Commercial Nuclear Power Reactors
       Exhibit 7-3 summarizes the potential impacts of EPA's rule for scrap metal from commercial
nuclear power reactors. As indicated  by the exhibit, the estimated impacts on scrap metal
management costs range from an increase of $0.17 billion to $0.41 billion under the 0.1 mrem option
to a savings of up to $0.08 billion under the 15.0 mrem standard. The savings attributable to the 15.0
mrem standard are less than those estimated for managing DOE scrap metal because a much higher
percentage of the scrap metal from power reactors (approximately 72 percent vs. six percent) must
be decontaminated prior to clearance.

       The higher activity levels assumed to be associated with power reactor scrap metal also affect
the risk results, with changes in the management of scrap metal  from power reactors accounting for
most of the total estimated change in cancer incidence under each option. Our analysis assumes that
final activity levels in scrap  metal items decontaminated prior to release are at the maximum levels
allowed under each option, while final activity levels in scrap metal released directly from facilities
with no prior decontamination are equal to starting activity levels.  As a result, the analysis treats
final activity levels for all scrap metal released from DOE facilities as lower, on average, than final
activity levels for scrap metal generated  by power reactors.

       The radionuclides and exposure  pathways that drive the cancer risk analysis also differ for
the two source categories. For DOE, U-238 is responsible for virtually all of the collective impacts
predicted; the key exposure pathway is exposure  to workers handling  slag.  In contrast, Co-60
accounts for the majority of predicted cancer cases related to' power reactor scrap, and the dominant
exposure pathway is through consumer products.
Exhibit 7-3
IMPACTS OF EPA CLEARANCE STANDARDS:
COMMERCIAL NUCLEAR POWER REACTORS
Impact1
Change in Costs (present value;
billions of 1997 dollars)
Change in Cancer Incidence2
Analytic Option
0.1 mrem
Low
Disposal
Costs
$0.17
(8.2)
High
Disposal
Costs
$0.41
(14.3)
1.0 mrem
Low
Disposal
Costs
$0.04
(6.3)
High
Disposal
Costs
$0.03
(10.0)
15.0 mrem
Low
Disposal
Costs
($0.04)
17.9
High
Disposal
Costs
($0.08)
27.6
Notes:
1 The values shown are based upon initial levels of radioactivity at the logarithmic midpoint of the range
reported for each scrap metal item.
2 Number of total cases (fatal and non-fatal) predicted to occur over 1,000 years.
                                            7-4

-------
                                                             Preliminary Draft: June 13, 1997
IMPLICATIONS AND NEXT STEPS
       Although these results are preliminary and subject to uncertainty, we can draw the following
initial conclusions from the analysis.

       •      The analysis provides useful information on the relative impacts of the three
              analytic options. The results, however, should not be interpreted as providing
              precise absolute estimates of disposition costs or potential cancer incidence.

       •      Cost impacts vary considerably across the three analytic options. The change
              in scrap metal  management  costs ranges from an increase of up to $0.5
              billion at the 0.1 mreni standard to savings of up to $1.7 billion at the 15.0
              mrem  standard.  These  results, however,  are  highly dependent  on our
              characterization of baseline practices.  If these practices diverge from those
              assumed in our analysis, the impact of the standards will change accordingly.

       •      Impacts on cancer  risks.also vary  considerably across  the three analytic
              options.  The 0.1 mrem standard yields an estimated decrease in cancer
              incidence of up to 14 cases over 1,000 years, while the 15.0 mrem standard
              yields an  estimated increase  of approximately  29  cases.  Again,  these
              conclusions  are  highly  dependent on our characterization of  baseline
              conditions, as well as on numerous other assumptions  imbedded in our risk
              analysis.

       •      The relationship between predicted changes in costs and predicted changes
              in cancer risks varies across the three analytic options. Under the 15.0 mrem
              standard, costs are predicted  to decrease relative to costs under the current
              standard, while cancer risks are predicted to increase. The 0.1 mrem standard
              yields the opposite result, increasing  predicted costs but decreasing the
              predicted number of cancer cases. In contrast, the analysis suggests that scrap
              metal management  costs would remain the same and cancer risks would
              decline  under the 1.0 mrem option.

       •      The  results  of the analysis are highly dependent upon  the  assumed
              radiological profiles of affected scrap metal. For example, the differences in
              predicted cost and cancer impacts between DOE  facilities and commercial
              power reactors are largely attributable to differences in the assumed mix of
              dominant  radionuclides  and related  activity  levels  in  the  scrap  metal
              generated by the respective  complexes. The  results of  the analysis are
              extremety..sensitive to variation in.radiolqgical profiles.    _ ,_,     ...
                                            7-5

-------
                                                              Preliminary Draft: June 13, 1997
Key Uncertainties
       Each component of the analysis, including identifying and characterizing affected volumes
of scrap metal, estimating baseline and post-regulatory scrap metal management practices and costs,
and estimating changes in human cancer risks, has associated limitations. In general, we can only
broadly characterize how these limitations may affect the results of the analysis.

       Much of the data on scrap metal characteristics is uncertain, particularly data on the year
in which scrap metal is likely  to become available for recycling. Radiological characteristics and
physical form data are also uncertain, particularly for DOE facilities.  In  addition, the  analysis
probably understates the total amount of scrap metal potentially affected by EPA's rule, since it
focuses only on scrap metal from major DOE facilities  and commercial nuclear power reactors;
other Federal and nonfederal facilities will also be affected by the rulemaking. These uncertainties,
may lead us to either under- or overstate the effects of alternate release standards.

       The characterization of future scrap metal disposition practices and related costs is also
uncertain.  For example, the analysis  does not consider restricted recycling, which may provide a
lower cost  disposition alternative  for some scrap metal. All  other things equal, the exclusion of this
option leads us to overstate total disposition costs, as well as the quantities disposed or released for
unconditional use. Moreover, decontamination costs are likely to change as the industry evolves, and
disposal options are difficult to predict. The extent to which these costs are under- or overstated is
uncertain;  hence, we are unsure whether they lead us to under- or overstate the effects of EPA's
rulemaking. Finally, the analysis assumes that generators will select the least cost disposition option,
ignoring the effects of non-economic factors that discourage  release of scrap metal. To the extent
that non-economic factors influence decision-making, we likely understate scrap metal management
costs and overstate the quantity of scrap likely to enter unconditional clearance.

       The analysis of cancer risks is also limited by key uncertainties. First, the risk model employs
a number of conservative assumptions that may lead it to overestimate doses under various exposure
scenarios. Second, our cost modeling estimates the maximum quantity of scrap metal that could
enter free release under alternate standards, leading to potential overstatement of collective cancer
impacts.  Closely related to this  uncertainty is the assumption regarding activity in scrap  metal
entering unconditional clearance. As noted, the cancer risk estimates presented here assume that
decontamination efforts reduce activity levels only to the maximum permitted under each release
standard. In reality, residual activity  in decontaminated  scrap metal is likely in most cases to  be
below the  standard. As a result,  our estimates tend to overstate likely cancer incidence. We are
uncertain,  however, how these limitations may affect our assessment of the  incremental effects of
each of the analytic options.
                                             7-6

-------
                                                             Preliminary Draft: June 13, 1997
Next Steps
       Over the next several months, we may conduct additional work to determine the most
important analytic uncertainties and strengthen our preliminary assessment of EPA's rule. Our
efforts may include the following:

       *      Calculating break-even points (e.g., for disposal vs. decontamination costs)
              and using probabilistic models (e.g., for the distribution of activity levels) to
              identify the most significant sources of uncertainty.

       «      Improving information concerning the volumes of scrap  metal potentially
              affected by the rulemaking, and refining the radiological profiles of individual
              scrap metal items. In addition, we may expand the analysis to address sources
              of scrap  metal other than DOE facilities and commercial nuclear power
              reactors.

       •      Expanding our analysis of disposition options to include the impact  of
              restricted recycling alternatives. This effort may also include refining our cost
              estimates, particularly our estimates of burial costs.

       »      Continuing to refine the risk models employed to calculate cancer incidence.
              This effort may include revisiting the characterization of various exposure
              scenarios and reconsidering other assumptions employed in the risk models.  •
                                            7-7

-------
Page Intentionally Blank

-------
                                     REFERENCES


American Metal Market. Metal Statistics.  1988, 1994 and 1996 editions.

American Society for Metals.  Metals Handbook TDesk Edition).  1985.

Calspan Corp. Environmental Impacts of Virgin and Recycled Steel and Aluminum. Prepared for
       U.S. EPA.  EPA/530/SW/117c. NTTS PB-253 487.  1976.

Carlin, James R, Daniel Edelstein, Stephen M. Jasinski, John F. Papp, Patricia A. Plunkert, and
       Gerard Smith.  Recycling-Non-Ferrous Metals: 1993 Annual  Report. Washington, DC: U.S.
       Bureau of Mines. 1995.

Chen, S.Y., S. Folga, LA. Nieves, and J. Arnish. Assessment of Risks and Costs Associated with
       Transportation, of  DOE Radioactive  Scrap  Metals (Draft).    Prepared for the  U.S
       Department of Energy. July 1995.

Cohen, S., and Associates. Technical Support Document: Evaluation of the Potential for Recycling
       of Radioactively Contaminated Scrap  Metal.   Prepared for the U.S.  Environmental
       Protection Agency. February 1997.

Cohen,  S.,  and  Associates.    Scrap Metal Inventories  at  U.S.  Nuclear  Facilities  Potentially
       Suitable for Recycling. Prepared for the U.S. Environmental Protection Agency. September
       1995.

Cohen, S., and Associates. Analysis of the Potential Recycling of Department of Energy Radioactive
       Scrap Metal. Prepared for the U.S. Environmental Protection Agency. August 14, 1995.

Committee on the Biological Effects of Ionizing Radiations, Board  of Radiation Effects Research,
       Commission on Life Sciences, National Research Council.  Health Effects of Exposure to
       Low Levels of Tnnizing Radiation  (BEIR V). National Academy  Press, Washington, DC.
       1990.

Denison,  Richard A. "Environmental Lifecycle  Comparisons of  Recycling,  Landfilling  and
       Incineration: A Review of Recent Studies." Annual Review of Energy and the Environment.
       Vol. 21. Annual Reviews, Inc.: Palo Alto, CA.  1996.

Fitch, Michael. "Disposal Cost Calculation." Chem-Nuclear Systems, Incorporated. September 12,
       1996.

Floyd,  Dennis.   "Economics  of Recycling Radioactive Scrap  Metal."   Manufacturing Sciences
       Corporation. Presented at Beneficial Reuse '95. .July 28,1995. .       ,    .   ,
Freeman^ A. Myrick.- The .Measurement of Environmental; and 'Refteurce Values:' Theory and
       Methods/  Washington-, DC:-Resources for. the Future.  1993,,    ^  ..;    '•

Gabbard, Alex, "Coal Combustion: Nuclear Resource or Danger," ORNL Review. 1996.

-------
                                     REFERENCES
                                       (continued)
Gogolak, C.V., A.M. Huffert, and G.E Powers. A Proposed Nonparametric Statistical Methodology '
       for the Design and Analysis of Final Decommissioning Surveys.  NUREG/CR-1505.  U.S.
       Nuclear Regulatory Commission. August 1995.

Gresalfi, Michael (Oak Ridge National Laboratory)  and Jayne Tellarico  (U.S. Department  of
       Energy), "Cost of Low Level  Waste Disposal-Baseline; Back-up  Information." August 30,
       1995.

Houck, Gerald W. Recycling Iron and Steel Scrap:  1993 Annual Report. Washington, DC:  U.S.
       Bureau of Mines.  1995.

Huffert, A.M., E.W. Abelquist, and W.S. Brown. Minimum Detectable Concentrations with Typical
       Radiation   Survey   Instruments  for  Various  Contaminants  and  Field   Conditions.
       NUREG/CR-1507.  U.S. Nuclear Regulatory Commission. August 1995.

Institute  of Scrap Recycling Industries (ISRI).  Recycling Scrap Iron  and Steel and Recycling
       Nonferrous Scrap Metals. 1993.

Institute  of Scrap Recycling Industries, Inc.  "Recycling Scrap Materials Contributes  to a Better
       Environment." Date  unknown.

International Atomic Energy Agency.  Effects of Ionizing Radiation on Plants and Animals at Levels
       Implied by Current Radiation Protection Standards.  Vienna, Austria. 1992.

Memorandum from Alvin Aim,  Assistant Secretary for Environmental Management.  "Policy  on
       Recycling  Radioactively Contaminated  Carbon  Steel."   U.S.  Department  of  Energy.
       September 20, 1996.

Memorandum to Members of the Regulatory Working Group, from Sally Katzen, U.S. Office of
       Management and Budget. "Economic Analysis of Federal Regulations Under  Executive
       Order No. 12866." January 11,1996.

Memorandum from James M. Owendoff, Deputy Assistant Secretary for Environmental Restoration,
       and Stephen P. Cowan, Deputy Assistant Secretary, Office of Waste Management.  "Use of
       Standardized Low-Level Waste Containers."  U.S Department of Energy.  April 17,1996.

National Council on Radiation Protection and Measurements. Limitation of Exposure to Ionizing
       Radiation. Bethesda, MD. March  1993.

National Council on Radiation Protection  and Measurements.  Effects  of Ionizing Radiation  on
       Aquatic Organisms.  Bethesda, MD. August 1991.

Nieves, L.A., S.Y. Chen, E.J. Kohout, B. Nabelssi, R.W. Tilbrook, and S.E. Wilson.  Evaluation of
       Radioactive  Scrap   Metal  Recycling.   Argonne National  Laboratory, Environmental
       Assessment Division. December 1995.

-------
                                     REFERENCES
                                      (continued)

NUREG/CR-0130, Technology. Safety and Costs of Decommissioning a Reference Boiling Water
       Reactor Power Station. Volumes 1 and 2.  Prepared by R.I. Smith et al., Pacific Northwest
       Laboratory, for the U.S Nuclear Regulatory Commission.

NUREG/CR-0672, Technology. Safety and Costs of Decommissioning a Reference Boiling Water
       Reactor Power Station. Volumes 1 and 2.  Prepared by H.D. Oak et al., Pacific Northwest
       Laboratory for the U.S Nuclear Regulatory Commission.

Oak Ridge National Laboratory. Energy Expenditures Associated With the Production and Recycle
       of Metals. Prepared by J.C. Bravard, et al. ORNL-NSF-EO-24. November 1972.

Pearson, G.A., et al., Gaseous Diffusion Facilities Contamination and Decommissioning Report.
       U.S Department of Energy. December 1995.

"SAFSTOR Decommissioning Plan for the Humboldt Bay Power Plant, Unit 3." Pacific Gas and
       Electric Company.  July 1994.

10 CFR 20, "Standards for Protection Against Radiation."

Tellus Institute. CSG/Tellus Packaging Study. Prepared for the Council of State Governments, the
       U.S. EPA, and the N.J. DEPE.  1992.

Trinity Environmental Systems.  "White Paper Issues Discussion and Recommended Resolution of
       Commingling, Production Cost, Mixed Waste, Throughput-Assumptions and Background for
       Recycle 2000 Option 3 (rev. ?.)." November 29, 1995.

'Trojan Nuclear Plant Decommissioning Plan, PGE-1061." Portland General Electric. June 1996.

U.S Atomic Energy Commission.  Regulatory Guide 1.86: Terminations of Operating Licenses for
       Nuclear Reactors.  Washington, DC. June 1974.

U.S. Department of Energy.  DOE Order  5400.5:  Radiation Protection of the Public and the
       Environment. Washington, DC.  1990; and "Response to Questions and Clarifications of
       Requirements and Processes.  DOE 5400.5, Section U.5 and Chapter tV Implementation
       (Requirements Relating to Residual  Radioactive Material)."  DOE Assistant Secretary for
       Environment, Safety and Health, Office of Environment (EH-41). November 17, 1995.

U.S. Department of Energy. Taking Stock: A Look at the Opportunities and Challenges from the
       Cold War Era. January 1996.
      i
U.S. Department of Energy. Scrap Metal Inventory Report. March 1995.

U.S. Department of Energy.  U.S.  Department of  Energy's Weapons Complex Scrap Metal
       Inventory. Prepared by John R. Dada, U.S. DOE Morgantown Energy Technology Center.
       July 1993.

-------
                                     REFERENCES
                                      (continued)

U.S. Department of Energy. Assessment of Risks and Costs Associated with Transportation of DOE
       Radioactively Contaminated Carbon Steel. Prepared by S.Y. Chen, et al., Argonne National
       Laboratory, for the Office of Environmental Management. November 1995.

U.S. Department of Energy.  INEL Metal Recycle Radioactive Scrap Metal Survey Report.
       Prepared by D.M. Funk, Lockheed Idaho Technologies Company, for the Idaho Operations
       Office. September 1994.

U.S. Department of Energy.  The 1996 Baseline Environmental Management Report (BEMR).
       June 1996.

U.S. Ecology, Incorporated, Washington Nuclear Center.  "Schedule A: Disposal Charge." August
       18, 1996.

U.S. Environmental Protection Agency.  Risk Assessment  Guidance for  Superfund Volume 1:
       Human Health Evaluation Manual (Part AV December 1989.

U.S. Environmental Protection Agency.  Radiation Site Cleanup Regulations: Technical Support
       Document for the Development of Radionuch'de Cleanup Levels for Soil. EPA402-R-96-
       011 A-D. September 1994.

U.S. EPA Office of Radiation  and  Indoor Air. Technical Support Document: Evaluation of
       Radioactivelv-Contaminated Scrap Metals.  Vols. I and II.  Prepared by S. Cohen  and
       Associates, Inc.  February 1996.

U.S. Nuclear Regulatory Commission. IE Circular 81-07, "Control of Radioactively Contaminated
       Material."  May 14,1981.

U.S. Nuclear Regulatory Commission.  Revised Analyses of Decommissioning for the Reference
       Pressurized Water Reactor Power Station. 1994.

U.S. Nuclear Regulatory Commission.  Revised Analyses of Decommissioning for the Reference
       Boiling Water Reactor Power Station.  1995.

U.S. Office of Management and Budget.  Guidelines and Discount Rates for Benefit-Cost Analysis
       and Federal Programs (Circular  A-94V October 29, 1992.

Warren, Stephen et al. Cost Model for DOE Radioactivelv Contaminated Carbon Steel Recycling.
       U.S. Department of Energy.  December 1995.

-------
                                       Preliminary Draft: June 12, 1997
                        Appendix A
         SURFACE AND VOLUMETRIC RELEASE LIMITS
UNDER CURRENT STANDARDS AND THE THREE ANALYTIC OPTIONS

-------
Page Intentionally Blank

-------
Exhibit A-l
SURFICIAL RELEASE LIMITS
(dpm/100 cm2)
Nuclide
C-14 .
Mn-54
Fe-55 -
Co-60
Ni-59
Ni-63
Zn-65
Sr-9CH-D
Nb-94
Mo-93
Tc-99
Ru-106+D
AgllOm+D
Sb-125+D
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152 ,
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-241
Cnv244
U-Senes
U-Separ.
U-Deplete ,
Th-Senes
0.1 Mrem Option
90,000
400
10,000,000
90
20,000,000
7,000,000
800
300
200
1,000,000
4,000,000
3,000
100
1,000
100
300
900
9,000
600,000
200
30
100
400
50
300
90
100
30
30
200
200
300
50
100
100
100
7,000
100
60
100
100
300
300
100
1.0 Mrem Option
900,000
4,000
100,000,000
900
200,000,000
70,000,000
8,000
3,000
2,000
- 10,000,000
40,000,000
30,000
1,000
10,000
1,000
3,000
9,000
90,000
6,000,000
2,000
300
1,000
4,000
500
3,000
900
1,000
300
300
2,000
2,000
3,000
500
1,000
1,000
1,000
70,000
1,000
600
1,000
1,000.
3,000 - -
• ' 3,000
1,000
15.0 Mrem Option
10,000,000
60,000
2,000,000,000
10,000
3,000,000,000
1,000,000,000
100,000
40,000
20,000
200,000,000
500,000,000
500,000
20,000
200,000
10,000
50,000
100,000
1,000,000
80,000,000
30,000
4,000
20,000
60,000
7,000
40,000
10,000
20,000 '
4,000
5,000
40,000
40,000 '
40,000
8,000
20,000
20,000
20,000
1,000,000
20,000
10,000
20,000
" " '"" 'lO.OOO ".-.
- , 40,000 - - -
40,000
20,000
Current Standards
5,000
5,000
5,000
5,000
5,000
5,000
5,000
1,000
5,000
5,000
5,000
5,000
5,000
5,000
100
5,000
5,000
5,000
5,000
5,000
100
100
100
100
100
100
100
1,000
100
5,000
5,000
5,000
100
100
100
100
100
100
100
100
5»(XK)
-~5,(K)0
5,000
1,000
Source: DOE Order 5400.5 and NRC Regulatory Guide 1 .86
^ote: These levels have been developed for analytic purposes only and do not reflect regulatory options currently under
consideration by EPA.
A-l

-------
Exhibit A-2
VOLUMETRIC RELEASE LIMITS
(pCI/g)
Nuclide
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
AgllOnH-D
Sb-125+D
1-129
Cs-!34
Cs-137+D
Ce-144+D
Pm-14?
Eu-152
Pb-21CH-D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D •
Pu-242
Am-241
Qn-244
U-Series
U-Separ.
U-DepIcte
Th-Senes
0.1 Mrem Option
100
0.5
10,000
0.1
20,000
9,000
1
0.2
0.2
2,000
5,000
2
0.2
2
0.1
0.4
1
6
700
0.3
0.03
0.2
0.3
0.01
0.07
0.02
0.2
0.04
0.04
0.3
0.3
0.3
0.07
0.2
0.1
0.1
9
0.1
0.08
02
0.03
0.2
0.3
0.02
1.0 Mrem Option
1,000
5
100,000
1
200,000
90,000
10
2
2
20,000
50,000
20
2
20
1
4
10
60
7,000
3
0.3
2
3
01
0.7
0.2
2
04
0.4
,3
3
3
0.7
1
1
1
90
1
0.8
1
0.3
2
3
0.2
15.0 Mrem Option
20,000
80
2,000,000
20
3,000,000
1,000,000
200
30
30
300,000
700,000
300
20
200
20
60
200
800
100,000
40
5
20
40
2
10
3
20
5
6
50
50
50
10
20
20
20
1,000
20
10
.,20
4
20
40
3
Mote: Current standards do not include release limits expressed as activity levels for
volumetric contamination. These levels have been developed for analytic purposes only
and do not reflect regulatory options currently under consideration by EPA

-------
                                 Preliminary Draft: June 12, 1997
                Appendix B





DEFBMITIONS OF PHYSICAL FORM CATEGORIES

-------
Page Intentionally Blank

-------
, Exhibit B-l
DEFINITIONS OF PHYSICAL FORM CATEGORIES
Category
Structural
Steel
Process
Systems/
Piping
Large
Components
Special
Materials
Scrap Metal
Sub-Category
Good Steel
Bad Steel
Tanks
Large Piping
Small Piping
Steam Generators
Turbine Rotors
Heat Exchangers
Fuel Racks
Diffusion Cells
Lead
Shield Cask
Copper, Wire
Copper, Volume
Aluminum, Surflcial
Aluminum, Volume
Nickel, Volume
Scrap Metal Pile
Scrap Equipment
Description
Heavy gauge metal with flat accessible surfaces. Surface to mass ratio less than 5 sq ft./lb
Light gauge metal or metal that requires sectioning to gain access to inside surfaces.
Surface to mass ratio more than 5 sq.ft /Ib.
Heavy gauge metal that requires sizing and disassembly. Usually constructed from
carbon steel.
System piping and components from supply, circulation, and drains
greater than 2.5 inches in diameter, allowing access for decon and survey
System piping and components from supply, circulation, and drains
less than 2 5 inches in diameter, requiring sectioning and sizing for access
Complete systems, including support pipe, valves, etc Assumes some in situ process to
lower activity.
Complete rotors; assumes some on-site disassembly already performed to insure '
transportability.
Includes shells and tubes, but assumes any asbestos insulation has been removed.
Spent fuel racks used to store fuel, usually constructed of stainless steel or aluminum with
a neutron poison (e.g , boron)
NA
Sheet, blankets or bricks
Carbon steel or stainless steel containers with lead added (poured or sheet) for shielding
Insulated wire 0.25 inches or larger
Insulated and non-insulated wire and windings greater than 0.25 inches.
Aluminum forms with flat accessible surfaces
Aluminum contaminated in depth
Nickel forms with contamination dispersed throughout the matrix or as an ingot.
Miscellaneous mixed metals with no common configuration, surface to mass ratios, or
metal chemistry. Usually rusted or painted and requires significant sizing and preparation.
Components that require extensive disassembly to access inside for decontamination and
survey.
Examples
I-beams, structural members, crane systems,
decking, railroad rail
Stairs/handrails, grates, damaged I-beams,
damaged structural steel, ductwork, small containers
(drums)
Vessels, tanks, large containers, sump or basin liner
Piping systems greater than 2.5 inches in diameter;
valves and components
Piping systems less than 2.5 inches in diameter;
valves and components
NA
NA
NA
NA
Gaseous diffusion plant converters, compressors,
motors, and enclosures
Shield bncks, shield blankets, sheets
NA
Wire, windings, bus-bars
Ingots, activated components
Tanks, plate, structural/frame
Compressor blades, ingots
Ingots, barrier material, porous membrane
Scrap yards (no significant segregation)
Small pumps and motors, cabinets, vehicles, material
handling equipment, control centers, hand tools,
power tools, air handling equipment, small heat
exchangers, radiators and small coolers and heaters
B-l

-------
Page Intentionally Blank

-------
                                     Preliminary Draft: June 12, 1997
                    Appendix C






DETAILED DATA ON DOE SCRAP CHARACTERISTICS

-------
Page Intentionally Blank

-------
Exhibit C-l
DETAILED DATA ON DOE SCRAP CHARACTERISTICS
BY PHYSICAL FORM
Physical Form
Diffusion Cells
Scrap Metal Pile
Tanks
Good Steel
Bad Steel
Large Piping
Scrap Equipment
RX Components
Nickel, Volumetric,
Heat Exchangers
Aluminum, Surficial
Aluminum, Volumetric
Small Piping
Copper Wire
Shield Casks
Copper, Volumetric
Lead
Total
Quantity
(metric tons)
632,701 • .
84,310
67,925
64,641
27,291
15,002
12,224
9,543
7,806
5,008
3,958
2,163
1,962
1,391
368
9
3
936,305
C-l

-------
Exhibit C-2
DETAILED DATA ON DOE SCRAP CHARACTERISTICS
BY SITE AND PHYSICAL FORM
Facility Name
Femald
Hanford
INEL
Physical Form
RX Components
Copper Wire
Tanks
Scrap Equipment
Total
Tanks
Large Piping
RX Components
Scrap Metal Pile
Small Piping
Good Steel
Aluminum, Surficial
Bad Steel
Shield Casks
Total
Scrap Metal Pile
Bad Steel
Small Piping
RX Components
Scrap Equipment
Good Steel
Large Piping
Shield Casks
Nickel, Volumetric
Tanks
Heat Exchangers
Quantity (metric tons)
2,691
1,273
191
63
4,218
66,172
14,467
6,394
2,073
766
742
685
586
291
92,176
34,154
758
524
444
274
198
166
77
47
26
11
C-2

-------
Exhibit C-2 (continued)
Facility Name
INEL (continued)
LANL
NTS
Oak Ridge Reservation
Paducah
Physical Form
Copper, Volumetnc
Copper Wire
Lead
Total
Scrap Metal Pile
Scrap Equipment
Bad Steel
Good Steel
Aluminum, Surficial
Large Piping
Tanks
Small Piping
Total
Scrap Metal Pile
Diffusion Cells
Good Steel
Scrap Equipment
Scrap Metal Pile
Bad Steel
Aluminum, Surficial
Tanks
Large Piping
Copper Wire
Heat Exchangers
Totil
Diffusion Cells
Scrap Metal Pile
Nickel, Volumetric
Bad Steel
Scrap Equipment
Aluminum, Volumetnc
Aluminum, Surficial
Quantity (metric tons
9
4
3
36,695
2,379
437
125
107
22
16
10
3
3,099
264
212,706
21,998
6,600
5,863
3,569
1,019
285
138
59
21
252,258
230,923
27,438
7,759
6,183
3,105
1,892
1,718
C-3

-------
Exhibit C-2 (continued)
Facility Name
Paducah (continued)
Portsmouth
Rocky Flats
SRS
Physical Form
Good Steel
Tanks
Copper Wire
Small Piping
Heat Exchangers
Large Piping
Total
Diffusion Cells
Scrap Metal Pile
Tanks
Bad Steel
Aluminum, Volumetric .
Good Steel
Copper Wire
Aluminum, Surficial
Large Piping
Total
Good Steel
Heat Exchangers
Good Steel
Scrap Metal Pile
Scrap Equipment
Small Piping
Tanks
Large Piping
Bad Steel
RX Components
Total
Quantity (metric tons)
91
41
34
19
18
14
279,235
189,072
6,642
962
941
271
73
19
5
1
197,986
26,303
4,959
4,544
4,036
1,456
650
239
200
135
14
16,233
C-4

-------
Exhibit C-2 (continued)
Facility Name
Weldon Spring
Physical Form
Bad Steel
Good Steel
Scrap Metal Pile
Aluminum, Surficia]
Scrap Equipment
Copper Wire
Total
Quantity (metric tons)
14,993
10,585
1,460
510 '
289
1
27,838
C-5

-------
Page Intentionally Blank

-------
                                   Preliminary Draft: June 12, 1997
                  Appendix D





DETAILED DATA ON NRC SCRAP CHARACTERISTICS

-------
Page Intentionally Blank

-------
Exhibit D-l
DETAILED DATA ON NRC SCRAP CHARACTERISTICS
BY PHYSICAL FORM
Form Description
Large Piping
Small Piping
Tanks
Turbine Rotors
Heat Exchangers
Good Steel
Scrap Equipment
Bad Steel
Fuel Racks
Copper Wire
Scrap Metal Pile
Lead
Aluminum, Surflcia]
Nickel, Volumetric
RX Components
Total
BWR
BWR Reference
Reactor
Quantity
2,228
2,562
317
1,084
1,414
' 159
290
394
276
144
0
10
4
<1
<1
8,881
Total
BWR
Quantity
75,201
86,483
10,706
36,592
47,740
5,369
9,789
13,286
9,330
4,861
0
333
122 '
7
7
299,825
PWR
PWR Reference
Reactor
Quantity
896
623
1,248
403
' 158
373
295
202
49
73
33
5
2
<1
<1
4,359
Total PWR
Quantity
70,169
48,760
97,693
31,547
12,374
29,159
23,089
15,808
3,842
5,675
2,599
360
141
8
8
. 341,233
Total AH
Reactors
145,370
135,243
108,399
68,139
60,114
34,528
32,878
29,094
13,172
~~~ 10,536 ,
2,599
693
263
15
15
641,058
D-l

-------
Page Intentionally Blank

-------
                            Preliminary Draft: June 12, 1997
             Appendix E

REJECT RATES FOR DECONTAMINATED
  SCRAP METAL BY PHYSICAL FORM

-------
Page Intentionally Blank

-------
Exhibit E-l
REJECT RATES FOR DECONTAMINATED SCRAP METAL
BY PHYSICAL FORM

Physical Form
Good Steel
Bad Steel
Tanks
Large Piping.
Small Piping
Steam Generators
Turbine Rotors
Heat Exchangers
Fuel Racks
Diffusion Cells
Lead
Shield Casks
Copper Wire
Copper, Volumetric
Aluminum, Surficial
Nickel, Volumetric
Aluminum, Volumetric
Scrap Metal Pile
Scrap Equipment
Rx Components
Silver
Baseline
5%
16%
5%
10%
40%
0%
5%
10%
30%
10%
1%
30%
20%
2%
5%
100%
100%
30%
20%
100%
100%
Analytic Options
0.1 mrem
4%
12%
4%
3%
14%
0%
0%
3%
2%
30%
0%
2%
16%
0% '
6%
10%
io%-
48%
38%
100%
100%
1.0 mrem
0%
0%
0%
0%
0%
0%
0%
0%
0%
1%
0%
0%
0%
0%
0%
10%
1%
0%
0%
100%
100%
15 mrem
0%
0%
0%
0%
0%
0%
0%
0%
0%
1%
0%
0%
0%
0%
0%
1%
1%
0%
0%
100% •
100%
Note: These values indicate the percentage of all scrap metal decontaminated for unconditional
clearance that is disposed of during the decontamination process.
E-l

-------
Page Intentionally Blank

-------
Page Intentionally Blank

-------
Page Intentionally Blank

-------
          REVIEW DRAFT
  TECHNICAL SUPPORT DOCUMENT

EVALUATION OF THE POTENTIAL FOR
   RECYCLING OF SCRAP METALS
     FROM NUCLEAR FACILITIES

  VOLUME 1 OF 3: CHAPTERS 1-10
             Prepared by:

       S. Cohen & Associates, Inc.
          1355 Beverly Road
        McLean, Virginia 22101
               Under

        Contract No. 68D20155
       Work Assignment No. 5-13
            Prepared for:

  U.S. Environmental Protection Agency
    Office of Radiation and Indoor Air
          401M Street, S.W.
        Washington, D.C. 20460

            Martin Offutt
       Work Assignment Manager

            July 15, 1997

-------
Page Intentionally Blank

-------
                                  CONTENTS
EXECUTIVE SUMMARY	~	ES-1

1.     INTRODUCTION	1-1

      1.1    Purpose	,	1-1

      1.2    Scope of the Analysis	,	1-1

      1.3    Organization of the f SD	1-3

      References	-	1-5

2.     OVERVIEW OF SCRAP METAL OPERATIONS  	2-1

      2.1  .  Characteristics of Scrap Sources	,	2-1

      2.2    Industry Perspectives	2-3

      2.3    Principal Scrap Metal Operations Considered	2-4

      2.4    Current Recycle Practice for Nuclear Facilities	2-6

      References	.-.	 2-8

3.     SCREENING PROCEDURES TO        THE SCOPE
      OF THE ANALYSIS	."*!	3-1

      3.1    Primary Purpose of the TSD	„.	 3-2

      3.2    Primary Screening Criteria	,	3-4

      3.3    Sources of Scrap Metal Considered - Administrative and
            Functional Categories	3-5

            3.3.1   Administrative Authorities		3-5
            3.3.2   Functional Categories 	'	-.	3-10

-------
                              CONTENTS (Continued)

      3.4    Types of Scrap Metal Considered    	'.	3-12

             3.4.1  Screening Based on Economic Value 	3-12
             3.4.2  Screening Based on Public Health Considerations	3-13

      3.5    Radionuclides Selected for Consideration	3-15

      3.6    Scenarios, Pathways, and Biological Endpoints Considered	3-15

      3.7    Summary of the Screening Process and Associated Limitations	3-19

             3.7.1  Sources of Scrap Metal  	3-19
             3.7.2  Types of Scrap Metal from Nuclear Facilities	-	3-20
             3.7.3  Radionuclides	3-21
             3.7.4  Scenarios, Pathways, Modeling Assumptions, and
                   Biological Endpoints	3-22

      References	'.	".	3-23 '

4.    QUANTITIES AND  CHARACTERISTICS OF POTENTIAL SOURCES OF
      SCRAP METAL FROM DOE FACILITIES AND COMMERCIAL NUCLEAR
      POWERPLANTS	4-1

      4.1    Existing and Future Scrap Metal Quantities Available From DOE	4-1

             4.1.1  Background Information	4-1
             4.1.2  Existing Scrap Inventories at DOE 	4-9
             4.1.3  Summary of Existing Scrap Inventories at DOE Sites 	4-14
             4.1.4  Scrap Metal Inventory by Metal Type	4-15
             4.1.5  Future Scrap Metal Quantities at DOE 	4-17
             4.1.6  Summary and Conclusions  	4-21

      4.2    Potential Sources and Characteristics of Scrap Metal from the
             Commercial Nuclear Power Industry	4-23

             4.2.1  Summary Estimates of Contaminated Steel for Reference B WR/
                   PWR and the Commercial Nuclear Industry	4-24

-------
                              CONTENTS (Continued)

             4.2.2  Contaminated Metal Inventories Other Than Steel	4-25
             4.2.3  Time-Table for the Availability of Scrap Metal from the
                   Decommissioning of Nuclear Power Plants	 4-26

             References	4-29

5.    DESCRIPTION OF UNRESTRICTED RECYCLING OPERATIONS	5-1

      5.1    Introduction	5-1

             5.1.1 Recycling Scrap Steel - An Overview	,	5-1
             5.1.2 Reference Facility	5-3
             5.1.3 Exposure Pathways	.	-.	5-4

      5.2    List of Operations and Exposure Scenarios	 5-5

             5.2.1  Dilution Factors			5-6
             5.2.2  Scrap Processing Operations	5-8
             5.2.3  Steel Mill	.-	5-8-
             5.2.4  Use of Steel Mill Products	5-11

      References	5-13

6.    CALCULATION OF RADIOLOGICAL IMPACTS	6-1

      6.1    Radioactive Contaminants	 1, 6-1

      6.2    Specific Activities of Various Materials	 6-5

      6.3    Exposure Pathways 	6-8

             6.3.1 External Exposures to Direct Penetrating Radiation	6-8
             6.3.2 Inhalation of Contaminated Dust	6-12
             6.313 Incidental Ingestion	6-14
             6.3.4 Radioactive Decay		,	6-16

      6.4    Unique Scenarios	6-17

             6.4.1  Ground Water Contaminated by Leachate from Slag Storage Piles  .. 6-18
             6.4.2  Ingestion of Food Prepared in Contaminated Cookware	6-32

-------
                              CONTENTS (Continued)

             6.4.3  Impact of Fugitive Airborne Emissions from the Furnace
                   on Nearby Residents	6-32
             6.4.4  Potential Doses to Individuals Following Disposal
                   of Recycled Metal	6-33

      References	...._	6-34

7.    RESULTS AND DISCUSSION OF RADIOLOGICAL IMPACTS
      ON INDIVIDUALS	 7-1

      7.1    Normalized Doses and Risks to the RMEI 	7-1

      7.2    Maximum Exposure Scenarios  	-	7-2

             7.2.1  Slag Pile Worker			 7-5
             7.2.2  Cutting Scrap   	7-5
             7.2.3  Lathe Operator	7-5
             7.2.4  EAF Furnace Operator	7-6

      7.3    Evaluation of the Results of the Radiological Assessment	7-6

             7.3.1  Dilution of Potentially Contaminated Steel Scrap	7-6
             7.3.2  Exposure Pathways	7-7
             7.3.3  Mass Distribution and Partitioning of Contaminants  	7-10
             7.3.4  Scenario Selection	7-10

      7.4    References	7-11
                                           ~> tgt
8.    DETECTION AND MEASUREMENT OF SCRAP CONTAMINATION	8-1

      8.1    Statement of Purpose	,"	8-1

      8.2    Guidelines and Standards for Free Release of Scrap
             Established by the NRC and DOE	8-1

             8.2.1  NRC: Regulatory Guide 1.86 (1974 and 1982)	8-2
             8.2.2  DOE Order 5400.5	8-4
             8.2.3  Release Criteria for Volumetric Contaminants ....-	 8-4
                                        IV

-------
                               CONTENTS (Continued)
                                                            j
       8.3    Verification of Residual Contamination on Materials Released
             for Unrestricted Use	.-	8-5

             8.3.1   Total Alpha and Beta-Gamma Direct Measurements 	8-6
             8.3.2   Surface Scanning for Total Alpha and Total
                    Beta Contamination	8-7
             8.3.3   Surveys for Total Gamma Contamination	8-7

       8.4    Lower Limit of Detection and Minimum Detectable Concentration	8-8

             8.4.1   MDCs for Surface Scanning for Small Areas of Contamination	8-9
             8.4.2   MDC for Surface Scanning for Large Areas of Contamination  	8-10
             8.4.3   MDC for Direct Measurements	8-11

       8.5    Radionuclide MDCs for Surface Contamination	8-11

       8.6    The Potential Impact of Introducing Radioactively Contaminated
             Metal in the Production of Steel  	8-27

       8.7    Limitations of Standard Survey Measurements When Scrap
             Is Contaminated Volumetrically  	8-28

             8,7.1   Limitations When Bulk Contaminant is a Beta Emitter  	8-28
             8.7.2   Limitations When Bulk Contaminant is an Alpha Emitter	8-30
             8.7.3   Limitations for Gamma Emitting Bulk Contaminants	8-31
                                                                 v
       8.8    Assessing the Radionuclide Concentration in Steel Produced from Scrap . .. 8-31

       8.9    MDCs and Associated Parameters for Laboratory Analysis of
             Radionuclides	 8-32

       8.10   Summary 	8-39

       References	8-39

9.      NORMALIZED COLLECTIVE IMPACTS MODELS	9-1

       9.1    Transportation	.,	9-3

       9.2    Airborne Emissions	9-6

-------
                              CONTENTS (Continued)

      9.3    Slag 	9-9

             9.3.1  Road-Building	9-10
             9.3.2  Fill	9-14
             9.3.3  Railroad Ballast	9-18
             9.3.4  Other Purposes	,	9-18
             9.3.5  Slag Normalized Collective Doses  	9-19

      9.4    Baghouse Dust	9-20

             9.4.1  Zinc Recovery	9-20
             9.4.2  Disposal in Landfill	9-20

      9.5    Finished Steel	9-21

             9.5.1  Automotive 	9-22
             9.5.2  Kitchen Appliances	9-24
             9.5.3  Office Buildings	9-26
             9.5.4  Cookware	9-28
             9.5.5  Finished Steel Normalized Collective Doses	9-30

      9.6    Total Normalized Collective Doses and Risks	9-31

      References	9-35

10.    EVALUATION OF UNCERTAINTIES PERTAINING TO SCRAP METAL
      QUANTITIES, ESTIMATES OF DOSE AND RISK, AND MINIMUM
      DETECTABLE CONCENTRATIONS	10-1

      10.1   Introduction	10-1

      10.2   Uncertainties in Scrap Metal Source Quantities and Levels
             of Contamination	10-5

             10.2.1 Scrap Metal from Nuclear Power Plants	10-5
             10.2.2 Scrap Metal from DOE Facilities 	10-9

      10.3   Uncertainty for Normalized RMEI Doses and Risks	i	10-13
                                        VI

-------
                         CONTENTS (Continued)

10.4  Uncertainty in Normalized Collective Dose Estimates ....		10-17

10.5  Uncertainties Regarding Minimal Detectable Concentrations for
      Radionuclide Contaminants	10-23

References	10-29
                                  vn

-------
APPENDIX A:

APPENDIX B:

APPENDIX C:

APPENDIX D:

APPENDIX E:

APPENDIX F:

APPENDIX G:


APPENDIX H:

APPENDIX I;

APPENDIX J:

APPENDIX K:

APPENDDCL:
               APPENDICES

Characterization of Scrap Metal Inventories at U.S. Nuclear Power Plants

Recycling of Aluminum Scrap

Recycling of Copper Scrap

Selection of Radionuclides for Radiological Impacts Assessment

Distribution of Radionuclides During Melting of Carbon Steel

Distribution of Radionuclides During Melting of Cast Iron

Dilution of Scrap Metal from Nuclear Facilities in Scrap Metal From
Other Sources

Detailed Description of Exposure Scenarios (not yet available)

Leaching of Radionuclides From Slag

Normalized Doses and Risks to Individuals - By Scenario

Individual Doses and Risks

Uncertainties in Evaluations to Date
                                      V1H

-------
                                  LIST OF TABLES


3-1    Inventory of Sites That Are Known to be Contaminated with Radioactivity	3-7

3-2    Estimates of Existing and Projected Potential
       Sources of DOE Scrap Metal Used for Screening Purposes (from MIN 96)	3-8

3-3    Functional Categories for Nuclear Facilities and Sites Containing or
       Contaminated with Radioactive Materials	3-11

3-4    Potential Economic Value of Types of Scrap Metal	3-13

4-1    Groupings in DOE Materials in Inventory			4-10

4-2    Existing Contaminated Scrap Inventories at DOE Sites	4-11

4-3    Summary Estimates of Existing DOE Scrap Metal
       Inventories (Metric Tonnes)	,	4-15

4-4    MIN Scrap Metal Inventory by Metal Type (Metric Tonnes)	4-16

4-5    Estimated Scrap Inventories by Metal Type Currently Stored at
       DOE Facilities (Metric Tonnes)	4-17

4-6    Summary Data for Existing and Future Contaminated Scrap at
       DOE Facilities	4-22

4-7    Summary Data for Contaminated Steel Inventories Potentially Suitable
       for Recycling			4-24

4-8    Summary of Contaminated Metal Quantities Other Than Steel (Metric Tonnes)	4-26

4-9    Time-Table for Available Contaminated Scrap Metals from Decommissioned
       Nuclear Power Plants Quantities (Metric Tonnes) 	4-27

5-1    Operations and Exposure Parameters for Radiological Assessments of Individuals — 5-7

6-1    Implicit Progenies of Nuclides Selected for Analysis	6-3

6-2    Nuclides Included in Various Combinations and Decay Series 	6-5

6-3    Partition Ratios (PR) and Concentration Factors (CF)  	6-7

                                          ix

-------
                             LIST OF TABLES (Continued)

6-4    Lung Clearance Class and FI Values for Use with FOR 11  	6-15

6-5    Potential Contaminants of Ground Water	6-20

6-6    Composition of Slag Used in Leaching Test			6-23

6-7    Leaching Parameters Values	 6-24

6-8    Diffusion Coefficients for EAF Slag Monolithic Samples	 6-25

6-9    Fractions of Various Toxic Elements Leached from Slags
       Using EPA TCLP Protocol  	6-27

7-1    Maximum  Exposure Scenarios and Normalized Impacts on the RMEI
       From One Year of Exposure		... 7-3

8-1    Regulatory Guide 1.86 Acceptable Surface Contamination Levels 	8-3

8-2    Detectability of Radionuclides (Small Area) by Surface Scan Relative
       to RG 1.86 Limits	8-15

8-3    Detectability of Radionuclides (Large Area) by Surface Scan Relative
       to RG 1.86 Limits  	8-17

8-4    Detectability of Radionuclides by Direct Count Relative to RG 1.86 Limits ...	8-19

8-5    Detectability of Radionuclides (Small Area) by Surface Scan
       Relative to DCLs	8-21

8-6    Detectability of Radionuclides (Large Area) by Surface Scan
       Relative to DCLs	,	,		8-23

8-7    Detectability of Radionuclides by Direct Count Relative to DCLs	8-25

8-8    Laboratory MDCs, Associated Parameters, and Costs  	8-35

8-9    Detectability of Radionuclides by Laboratory Analysis Relative to DCLs	8-37

9-1    Major Transportation Pathway Assumptions	9-4

9-2    Unweighted Transportation Doses (person-rem per Ci-transported)	9-5

-------
                             LIST OF TABLES (Continued)


9-3    Unweighted Airborne Emission Doses (person-rem per Ci-released),	 9-7


9-4    Slag Pathway Normalized Collective Doses (person-rem per Ci-in pathway) 	9-12


9-5    Typical Landfill Values Assumed	,	 9-16


9-6    Landfill Normalized Collective Doses (person-rem per Ci-disposed)		 9-16


9-7    Annual Steel Slag Sales (thousand metric tonnes) 	9-19


9-8    Primary Assumptions Used in the Automobile Model	9-22


9-9    Automobile Doses (person-rem per Ci-in car)	-	9-23


9-10   Kitchen Model Major Assumptions	9-24

                       )
9-11   Finished Steel: Kitchen (Seven Appliances) Doses (person-rem per
       Ci-in appliance)	9-25


9-12   Finished Steel: Office Building Doses (person-rem per Ci-in office)	  9-28


9-13   Finished Steel: Frying Pan Doses (person-rem per Ci-in pan)	  9-29


9-14   Distribution of Finished Steel in Commercial Products  	  9-30


9-15   Total Normalized Collective Dose and Risks (per Ci-in scrap)  	  9-33


10-1   Selection of Data Sources for Scrap MetaMJuantities at DOE Facilities	10-10


10-2   Uncertainty/Variability in Normalized Individual Doses	10-15


10-3   Relative Range in MDCs .	;	10-28
                                          XI

-------
                                LIST OF FIGURES


4-1   The U.S. Nuclear Weapons Complex	  4-2

5-1   Operations Analyzed  	  5-2

9-1   Potential Collective Exposure Scenarios	  9-1

9-2   Collective Impact Calculational Approach	  9-2

9-3   Simplified Flow Diagram of the MEPAS Methodology	 .  9-15

9-4   Office Module General Layout and Construction Details	  9-27

10-1  Comparison of a Deterministic Model and a Probabilistic Model
      (from Little 1983)  	  10-3

10-2  Bounding Normalized RMEI Dose Values (mrem/y per pCi/g)	10-16

10-3  Collective Impact Calculational Approach		10-18

10-4  Effects of Ambient Background on MDC Calculation	10-28
                                        Xll

-------
                                EXECUTIVE SUMMARY

The operation, decommissioning, and cleanup of nuclear facilities owned by the Federal
government and licensed by the U.S. Nuclear Regulatory Commission (NRC) will likely
generate large quantities of scrap metal.- Some of this metal may be moderately contaminated
with radioactivity as a result of deposition or neutron activation. Current practice is to dispose of
such material in a licensed, low-level waste disposal facility. The U.S. Environmental Protection
                  *                                                  *
Agency (EPA) is evaluating the potential for recycling scrap metal from nuclear facilities as an
alternative disposition option. The Agency is also assessing the need for regulatory action to
ensure that the recycle of this scrap metal does not endanger public health and safety.

This Technical Support Document (TSD) summarizes the technical information used by EPA in
its evaluation. In a separate document, "Radiation Protection Standards for Scrap Metal:
Preliminary Cost-Benefit Analysis" (EEC 97), EPA describes its analysis of the potential costs
and benefits of recycling scrap metal from nuclear facilities.

The primary objectives of the Agency's technical analysis were to:

       1.     characterize the potential sources of scrap metal, including uncertainties, that may
             be available for recycling;

       2.     estimate, including uncertainties, the potential normalized annual dose and
             normalized lifetime risk to the reasonably maximally exposed individual (RMEI)
             associated with the release of scrap metal from nuclear facilities;

       3.     estimate, including uncertainties, the potential normalized collective dose and
             normalized collective risk to the exposed population associated with the release of
             scrap metal; and

       4.     estimate, including uncertainties, the minimum detectable concentration (MDC)
             of radionuclides contained within or on the surface of scrap metal.
                                          ES-1

-------
ES. 1   CHARACTERIZATION OF THE POTENTIAL SOURCES OF SCRAP METAL
       FROM NUCLEAR FACILITIES

The principal administrative authorities responsible for the management of scrap metal from
nuclear facilities are the U.S. Department of Energy (DOE), NRC, the Department of Defense
(DoD), and State or Superfund authorities. The nuclear facilities managed by the DOE and
commercial nuclear power plants licensed by NRC are the largest potential sources of scrap
metal from nuclear facilities in the United States. The DOE facilities and NRC-licensed
commercial power reactors were the basis for the evaluations reported in the TSD.

ES.1.1 DOE Scrap Metal

Information in several reports (DOE 95; DOE 96; EPA 95) was used to estimate the inventory of
scrap metal currently in storage at DOE facilities. Based on these reports, the current inventory
of scrap metal potentially available for recycle is about 171,000 metric tonnes. However, DOE
96 cautions that its efforts were to

       "not attempt to capture the exact amount of each material hi inventory. Rather,
       [they attempt] to capture the general magnitude of the inventory of each material
       (MEN 96)."

With respect to potential future sources of DOE scrap metal, DOE's Office of Environmental
Restoration Decommissioning Inventory slated 865 structures for future decommissioning (as of
June 1995). Future DOE scrap metal quantities will be closely linked to projected
decommissioning activities at DOE sites that make up the nuclear weapons complex. At some
sites, virtually all structures and their contents willobe dismantled and removed; at other sites,
decommissioning may be limited, and the DOE will continue selected operations considered
crucial to national security or important to national research. To date, final decisions and
commitments for decommissioning all of its facilities have not yet been made.

Based on available data, the quantity of future sources of DOE scrap metal is estimated to be
about 925,000 metric tonnes. Table ES-1 provides summary estimates of the combined quantity
of existing and future scrap metal from DOE facilities. Of the more than one million metric
tonnes of scrap, about 85 percent represents carbon steel with nearly equal quantities of copper,
nickel, aluminum, and stainless steel representing the remainder. These values may substantially
underestimate the total scrap metal quantities because current plans for future decommissioning
of DOE sites have not yet been finalized.
                                         ES-2

-------
            Table ES-1.  Summary Data for Existing and Future Contaminated Scrap at DOE Facilities* (metric tonnes)
Site
Name
Fernald
Han ford
Idaho
LANL
NTS
ORNL
Y-12
K-25
Paducah
Portsmouth
Rocky Flats
SRS
Weldon Sp -
TOTAL
Percent
of Total
Scrap
- Metal -
•: Database
Volume
139,841
92,175
34,213
5,785
264
1,129
9,065
242,063
279,260
197,986
50,846
16,237
27,839
1,096,703
100.00
; • Metal Type
* --in 	 '.y. :: 	 *!.'. 	 : 	 . .< . . 	 . .. . ...
AI ,
—
684
30
40
1!
18
33
7,988
' 21,161
6,130
...
14
510
36,619
3.34
C. Steel
101,753
87,020
19,195
" 5,568
204
992
8,392
232,953
212,917
191,412
33,666
10,403
26,877
931,352
84.92
s,:$t$ei
—
787
14,733
177
15
117
602
753
190
18
2,454
5,809
406
26,061
2.38
,'Copper
38,088
...
44
•••
...
2
38
304
.,4,98
408
14,726
11
46
53,865
4.'91
Nickel'
—
24
44
...
17
...
...
...
44,794
...
...
—
...
44,879
4.09
MOiiel ,
—
...
...
...
...
...
...
65
...
18
...
...
...
83
0.01
Lead
—
291
110
...
...
...
...
...
...
...
...
—
...
401
0.04
Cast fron
—
...
4
...
...
...
...
...
...
...
...
...
...
4
3.6E-6
Black Iron
—
...
7
...
...
...
...
...
...
...
...
—
...
7
6.4E-6
Graphite
...
1,632
...
...
...
...
...
...
...
...
...
—
...
1,632
0.15
Cu/Brass
—
5
8
••<•
2
...
...
...
...
...
...
—
...
15
1.4E-5
Tjn/Fe
...
1
2
...
1
...
...
...
...
...
...
...
...
4
3.6E-6
Other
—
1,711
36
...
...
...
...
...
...
...
...
...
...
1,747
0.16
Misc, :
—
20
...
...
!4
...
...
> •••
...
...
...
...
•..
^ 34
3.1E-5
Includes all metals which may be available for recycle

-------
ES.1.2 Commercial Nuclear Power Plants

The U.S. commercial nuclear power industry includes 123 reactor plants. At present, eight of
these reactors have been shutdown; in the next two to three decades, most of the reactors
currently in operation will have reached then- projected forty-year lifetime, A great deal of
information and data has been compiled by the NRC and the individual utilities pertinent to the
decommissioning of these facilities and the associated quantities and characteristics of the scrap
metal that will be produced.

In the 1976-1980 time frame, two studies were done for the NRC which examined the
technology, safety, and costs of decommissioning large reference nuclear power plants.  These
studies reflected the industrial and regulatory situation of the time. To support the.final
Decommissioning Rule issued by the NRC in 1988, the earlier studies were updated -with two
additional topical reports. These four reports, along with several other NRC reports and selected
decommissioning plans on file with the Commission, were the primary sources of information
used by EPA to characterize Reference Pressurized Water Reactor (PWR) and Boiling Water
Reactor (BWR) facilities and to derive estimates of scrap metal inventories for the industry at
large.

Quantities of both carbon steel and stainless steel will potentially be available for recycling from
decommissioned commercial nuclear power plants.  Estimates for the entire commercial nuclear
industry were derived by taking Reference BWR and Reference PWR values and applying plant-
specific scaling factors for each of the 40 BWRs and 83 PWRs in existence. Approximately
600,000 metric tonnes of steel may become available over time for recycling.  About 80% of this
metal is carbon steel with stainless steel representing most of the balance.

-------
ES.2   ESTIMATES OF THE NORMALIZED ANNUAL DOSE AND NORMALIZED
       LIFETIME RISK

In order to evaluate the potential impacts of recycling scrap metal from nuclear facilities, a
relationship must be established between the possible levels of radioactivity in the scrap and the
potential doses and risks to individuals that may be exposed to the scrap or to the products and
byproducts of the recycling process.  For the purposes of establishing this relationship, EPA
identified over 60 different categories of individuals that have the potential to receive some level
of exposure to residual radioactivity contained in the scrap. Within each of these categories,
there will be a range of exposure levels depending on the actual activities of each individual.  As
part of its analysis, the Agency determined the limiting category for each of the radionuclides
considered.  (Limiting refers to that category which has the highest potential doses associated
with a given radionuclide). The Agency also determined, within each category, the individuals)
who has the potential to receive high end exposures, e.g., 90th percentile. These individuals are
referred to as reasonably maximally exposed individuals (RMEI). It is unlikely that many
individuals within or outside the group could receive exposures significantly greater than those
received by the RMEI; most individuals that may be exposed are likely to receive exposures that
are substantially lower than those received by the RMEI.

Since doses and risks to individuals are directly proportional to the residual radioactivity in the
scrap, the Agency has chosen to express this relationship in terms of a "normalized dose"  for
each radionuclide of concern: The normalized dose is expressed in units of millirem per year
(rnrem/y) effective dose equivalent (EDE) per picoCurie per gram (pCi/g) of specific
radionuclides in released scrap metal.

The normalized dose is a useful metric because, for any tree release criterion established in units
of mrem/y, the normalized dose can be used to derive the average1 radionuclide concentration
level in scrap metal that corresponds to the criterion.  If the average concentration of a given
radionuclide hi scrap metal is known, the annual dose to the RMEI resulting from its release can
       1 When deriving the normalized annual dose to the RMEI, the volume of scrap metal over which the
radionuclide concentrations are averaged differs depending on the exposed individual. For example, when deriving
the normalized annual dose to a mill worker, the volume of scrap metal of concern is the entire scrap metal
throughput at the mill over a year.  However, when deriving the normalized annual dose to a user of a product made
from scrap metal from a nuclear facility, the averaging volume is the volume of scrap metal required to make the
product.
                                           ES-5

-------
be estimated by multiplying the normalized dose by the average concentration of the
radionuclide in the scrap metal.

The normalized risk is similar to the normalized dose except that it is expressed in units of the
lifetime risk of cancer associated with one year's exposure resulting from recycling operations,
per pCi/g of a given radionuclide in scrap metal. It can be used to derive that concentration of a
given radionuclide in scrap metal that corresponds to a given risk.  It can also be used to derive
the potential lifetime risk of cancer for the RMEI from one year's exposure associated with the
release of scrap metal containing a known average concentration of a given radionuclide.

In order to derive the normalized individual doses, visits were made to scrap yards and steel mills
to gather information on the handling and processing of scrap metal for recycling and on the
disposition of the products and byproducts associated with the overall recycling process for
carbon steel. Table ES-2 presents the results of these analyses. Six of the over 60 categories of
individuals evaluated were found to be limiting. A particular category is considered limiting for
a particular radionuclide because of factors concerning how the radionuclide partitions during
melting; whether the radionuclide is diluted; whether the radionuclide has the potential to escape
the baghouse filter; the mobility and bioavailability of the radionuclide in the environment; and
the activities and living habits of the individuals that may come into contact with the scrap, steel,
slag, baghouse dust, or airborne effluents from the mill.
                                          ES-6

-------
Table ES-2. Derived Normalized Doses and Risks to the RMEI
               from One Year of Exposure
Nuclide
C-I4
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
Ag-llOm+D
(Sb-125+D
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Limiting Scenario
• Airborne effluent emissions
Lathe operator
Cutting scrap
Lathe operator
Cutting scrap
Cutting scrap
Cutting scrap
Slag-leachate in groundwater
Slag pile worker
Cutting scrap
Cutting scrap
Lathe operator
Lathe operator
Cutting scrap
Airborne effluent emissions
Cutting scrap
Cutting scrap
Slag pile worker
Slag pile worker
Slag pile worker
EAF furnace operator
Slag pile worker
Slag pile worker
Cutting scrap
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Dose
(mrern per pCi/g)
8.66E-04
2.02E-01
6.69E-06
8.99E-01
4.39E-06
1.07E-05
9.61E-02
1.52E+00
4.74E-01
5.65E-05
2.15E-05
5.16E-02
6.29E-01
6.37E-02
7.91E-01
2.46E-01
8.91E-02
1.77E-02
1.42E-04
3.44E-01
3.08E+00
6.27E-01
3.68E-01
8.00E+00
1.35E+00
4.37E+00
6.42E-01
2.84E+00
2.51E400
3.14E-01
3.28E-01
2.89E-01
1.53E+00
Lifetime Risk
of Cancer
(per pCi/g)
4.28E-10
1.54E-07
2.71E-12
6.84E-07
1.55E-12
4.41E-12
7.31E-08
5.51E-07
3.60E-07
1.17E-11
1.41E-11
3.93E-08
4.78E-07
4.85E-08
5.04E-07
' 1.87E-07
6.77E-08
1.36E-08
8.31E-11
2.61E-07
4.37E-07
4.36E-07
2.36E-07
L35E-07
6.17E-06
2.32E-07
3.44E-08
3.34E-08
5.20E-08
3.31E-08
5.90E-08
3.55E-08
1.36E-07
                        ES-7

-------
               Table ES-2. Derived Normalized Doses and Risks to the RMEI
                          from One Year of Exposure (Continued)
Nuclide
C-14
Pu-238
Pu-239
Pu-240
' Pu-241+D
Pu-242
Am-241
Cm-244
U-Series
U-Separ.
U-Deplete
Th-Series
Limiting Scenario
Airborne effluent emissions
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
EAF furnace operator
Slag pile worker
Slag pile worker
Slag pile worker
Dose
(mrem per pCi/g)
8.66E-04
6.82E-01
7.29E-01
7.29E-01
1.17E-02
6.93E-01
1.21E+00
6.75E-01
3.61E+00
6.18E-01
3.22E-01
4.SSE+OQ
Lifetime Risk
of Cancer
(per pCi/g)
4.28E-10
4.78E-08
4.73E-08
4.73E-08
4.01E-10
4.46E-08
1.07E-07
6.69E-08
4.78E-07
7.14E-08
3.95E-08
8.86E-07
A semi-quantitative uncertainty analysis was performed which evaluated the uncertainty/
variability in the dose evaluation results due to uncertainty/variability in the calculational
parameters and assumptions. In addition, sensitivity analyses were performed to evaluate which
parameters contributed most to the uncertainty. Table ES-3 provides the results of the semi-
quantitative uncertainty/sensitivity analysis.

In Table ES-3, the critical population group is identified for each radionuclide, along with the
dominant exposure pathway(s). An "upper-end multiplier" and "lower-end divisor" is provided.
The upper-end multiplier and lower-end divisor define the potential range (and therefore the
uncertainty) of the normalized RMEI dose. The values for the multipliers and divisors are
largely based on professional judgment and are designed to bracket estimated uncertainties and
variabilities for normalized RMEI doses.  In general, the analyses demonstrate that the
normalized doses for the RMEI could, in theory, be higher by a factor of 5 to 50, or lower by up
to a factor of 100 to 500, depending on the radionuclide.  The uncertainty in the normalized risks
are similar.
                                          ES-8

-------
Table ES-3. Uncertainty/Variability in Normalized Individual Dose
?' 'ItedfemStJ'ffes'T-'s
?,•, '* ., -?.'
Zn-65*
Sb-125 -
Cs-134*
Cs-137*
Ni-59/63
Mo-93
Tc-99
Ac-227+D
Fe-55
Mn-54
Co-60
Ru-106
Ag-llOm+D
Nb-94
Ce-144+D
Eu-152
Ra-226+D/228+D
T!i-228+D
Pm-147
Th-229/230/232
Pa-231
U-234/235/23S
Np-237
Pu-all
Am-241
Cm-244
Pb-210
,l^ffi^g^iiaiibft .•
>','• :0w' 'i •
Scrap yard workers
Users of metal products
Slag pile
Slag pile workers
Mill workers
^Maij?Sf®iw ,
&»;• ' ''*,*.
Upper end due to eliminating dilution factor.
Lower end due to additional dilution (30 fold), reduced occupancy and
increased distance (3).
Upper end due to eliminating dilution factor.
Lower end due to additional dilution (30 fold), reduced occupancy (2), and
reduced dust loading (10)
Upper end due to eliminating dilution factor.
Lower end due to additional dilution (30), reduced 'occupancy (2), and reduced
soot ingestion (10)
Upper end due to increase in size of component and occupancy time (5).
Lower end due to application of a dilution factor (30) and lower occupancy time
and smaller size component (3).
Upper end due to elimination of dilution factor (9) and increased occupancy
time and slag partition (4).
Lower end due to additional dilution (30) and smaller contaminated area and
occupancy time (3).
Upper end due to elimination of dilution factor (9) and increased occupancy
time and slag partition (2).
Lower end due to additional dilution (30), lower dust loading (10), and lower
occupancy time (2).
Upper end due to elimination of dilution factor (8) and increased occupancy
time and slag partition (2).
Lower end due 19 additional dilution (30), lower soot ingestion (10), and lower
pccupancv time (2).

-------
                                       Table ES-3. Uncertainty/Variability in Normalized Individual Dose (Continued)
• .RadsOTMides t f

C-14
1-129

Sr-90
r Cfiikaf fiflpularfon
: - Gioup ' •-

Offsite residents • Offsite
emissions
Offsite residents -
Ground water
contaminated by slag
leachate
PrinSary Exposure.
=Jathvvay



Ground water
ingestion
pgperEnd ,
MUttipliei-

SO


50
Lower Bfrf Divisor

NA'
100

NA'
' 't -V • ,,5", gaje^C&fttfolliflg ?itafee!er£ ^ "•

Upper end due to elimination of dilution factor (8), closer location (3),
increased intake of crops (2).
Lower end due to additional dilution (30), further distance (2), less intake (2).
Upper end due to less dilution in ground water.
Lower end due to elimination of ground water due to increased transit time, anc
soot ingestion becomes the limiting pathway.
      * These radionuclides partition to baghouse dust.  If it is plausible for individuals to be exposed to reconcentrated stages of the metal recovery process for prolonged periods of time, the upper end
      multiplier for these radionuclides could be as high as a factor of 100.
      t A lower limit for these pathways in not applicable, since the lowest limiting dose will be due to a different pathway (see text).
W
C/3

-------
ES.3   ESTIMATES OF THE NORMALIZED, TIME-INTEGRATED COLLECTIVE DOSE
       AND NORMALIZED, TIME-INTEGRATED POTENTIAL COLLECTIVE PUBLIC
       HEALTH IMPACTS

The normalized, time-integrated collective dose is expressed in units of the collective dose (i.e.,
person rem) to which a population is committed per unit activity contained in free released scrap
metal.  The units can be simply expressed in terms of person rem per Curie, for example. The
normalized collective dose represents the sum of all individual exposures for the entire exposed
population for as long as the radionuclide can reasonably be assumed to result in human
exposures.  The normalized collective dose is a convenient metric because, once a determination
is made of the total radionuclide inventory that may be present in scrap metal, the collective dose
can be determined by simply multiplying the curie inventory associated with the scrap metal by
the normalized collective dose for each radionuclide.

The time-integrated normalized collective risk is similar to the time-integrated normalized
collective dose except that it is expressed in units of the potential numbers of adverse health
effects per Curie of each radionuclide that may be contained in scrap metal. It can be used to
derive the potential number of cancers that may be attributable to the release of a given quantity
of scrap metal containing a known inventory of radionuclides.

Table ES-4 presents the estimated normalized time-integrated collective doses and risks for each
of the 40 radionuclides considered in these evaluations. The values were derived based on
models representing the fate of each of these radionuclides. The models take into consideration
the wide variety of products and byproducts that could be produced from metal recycling and the
populations that may be exposed.                ^

The column in Table ES-4 labeled "Trans" refers to the collective dose associated with the
transportation of scrap metal, metal products, and byproducts comprised of scrap metal from
nuclear facilities. "Air" refers to the collective off site population doses that may occur due to
airborne emissions associated with steel mills that may recycle scrap metal from nuclear
facilities. "Slag" refers to the population impacts associated with radionuclides in slag produced
at mills that recycle scrap from nuclear facilities and used in a wide variety of applications, such
as road building. "Dust" refers to the population doses associated with the radionuclides that
may accumulate in baghouse dust during mill operations. "Steel" refers to the population doses
                                        ES-11

-------
associated with the radionuclides that may partition to steel which is then used in a variety of
products.

The models also take into consideration the time period over which populations may be exposed.
Any radionuclides hi steel products made from scrap from nuclear facilities have the potential to
cause exposures for as long as the products remain in the accessible environment.  The values in
Table ES-4 are therefore referred to as "time-integrated" collective impacts. As is the case for
the normalized doses for the RMEI, the time-integrated normalized collective doses are derived
to ensure that the potential collective health impacts are not underestimated; i.e., each value
represents an upper end estimate given the range of uncertainties.

A semi-quantitative uncertainty analysis for collective dose was performed which evaluated the
uncertainty in the results due to uncertainty in the calculational parameters and assumptions. In
addition, sensitivity analyses were performed to evaluate which parameters contributed most to
the uncertainty.  In general, the results of the uncertainty analyses reveal that the collective doses
could be higher or lower by less than a factor of two to three. The collective risks could also be
marginally higher, but the possibility exists that the risks could be zero for extremely low doses
and dose rates.
                                          ES-12

-------
Table ES-4.  Normalized Time-Integrated Collective Dose and Risks (per Ci—in scrap)
KudWr:
'* , •! *
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
Ag-llOm
Sb-125
1-129
Cs-134
Cs-137+D
Ce-144+D .^
Pm-147
Eu-152
Pb-210+D
. ;<. -~;> .,
;;jT&<$;"-""
O.OOE+OO
1.42E-02
O.OOE+OO
6.35E-02
O.OOE+OO
O.OOE+OO
5.31E-03
O.OOE+OO
5.77E-03
5.60E-07
1.45E-09
4.63E-03
6.58E-02
8.78E-03
8.19E-07
1.64E-02
5.82E-03
1.29E-04
3.27E-09
4.02E-03
9.19E-07
",- *'• '•"' '
' T;$r*"Y;/
4.74E+01
2.73E-04
1.61E-06
1.22E-03
8.80E-06
4.86E-05
2.44E-02
6.40E-02
8.02E-02
2.71E-04
U8E-03
1.47E-04
3.53E-04
3.41E-05
1.43E+03 -
6.00E-02
1.79E-01
5.07E-04
4.41E-05
4.94E-03
4.08E+00
,- :^m-^
/'"'^p'g;';."
O.OOE+OO
1.67E-01
3.98E-03
O.OOE+OO
O.OOE+OO
O.OOE+OO
O.OOE+OO
5.87E+01
5.35E+02 .
O.OOE+OO
O.OOE+OO
,< O.OOE+OO
O.OOE+OO
O.OOE+OO
O.OOE+OO
1.33E-01
7.42E-01
5.04E-02
6.49E-01
1.01E+01
O.OOE+OO
mwro) ' .
•."".H$br> -
O.OOE+OO
O.OOE+OO
1.64E-17
O.OOE+OO
O.OOE+OO
O.OOE+OO
1.65E+02
3.09E+00
2.81E+01
O.OOE+OO '
O.OOE+OO
O.OOE+OO
O.OOE+OO
O.OOE+OO
O.OOE+OO
O.OOE+OO
O.OOE+OO
2.65E-03
3.42E-02
5.34E-01
O.OOE+OO
\ ^ ^ .,..,.*
I .xSfcfcj'":;
2.12E-02
3.80E+02
2.65E-04
-1.01E+04
1.33E-03
6.84E-03
6.13E+01
O.OOE+OO
O.OOE+OO
2.53E+01
1.91E-01
1.76E+02
1.49E+03
4.02E+01
O.OOE+OO
O.OOE+OO
O.OOE+OO
. O.OOE+OO
O.OOE+OO
O.OOE+OO
O.OOE+OO
( ,jj- •-• . , -,,-~
I c- ir^t"". "
4.75E+01
3.80E+02
4.24E-03
1.01E+04
1.34E-03
6.89E-03
2.26E+02
6.18E+01
5.63E+02
2.53E+01
1.92E-01
1.76E+02
1.49E+03
4.02E+01
1.43E+03
2.09E-01
9.27E-01
5.36E-02
6.83E-01
, 1.07E+01
4.08E+00
. .^,^.,pm
."r/'Tm"*' :
2.34E-02
2.88E-01
1.57E-07
7.66E+00
1.18E-06
6.60E-06
1.72E-01
2.25E-02
3.81E-01
1.92E-02
1.58E-04
1.34E-01
1.13E+00
3.05E-02
9.39E-01
1.29E-04
6.39E-04
2.73E-05
2.05E-06
7.77E-03
3.67E-04
sere -v'r^''.
••"-'• y^rajr/-
1.62-02
1.93E-01
1.56E-07
5.13E+00
1.18E-06
6.58E-06
1.15E-01
1.78E-02
2.54E-01
1.29E-02
1.24E-04
9.01E-02
7.61E-01
2.05E-02
9.64E-02
8.68E-05
4.27E-04
1.74E-05
1.09E-06
5.17E-03
2.90E-04

-------
                 Table ES-4.  Normalized Time-Integrated Collective Dose and Risks (per Ci—in scrap) (Continued)
4:Nl«sHcte
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-241
Cm-244
Doses (persothrem)
Twn$ :
6.44E-03
3.31E-03
9.45E-04
5.60E-03
7.07E-04
2.57E-07
8.08E-08
7.04E-05
6.35E-08
2.48E-04
6.19E-05
5.09E-04
1.18E-08
4.78E-08
1.16E-08
3.53E-09
1.02E-08
4.66E-06
1.02E-08
Air
2.60E+00
1.49E-02
6.86E-01
2.48E-01
8.30E-01
9.42E-01
9.83E-01
1.52E+00
2.98E-01
2.83E-01
2.72E-01
5.01E+00 4
3.25E-01
3.65E-01
3.64E-01
5.43E-01
3.47E-01
5.70E-01
2.85E-01
Slag
2.81E+03
1.12E+01
2.00E+01
2.18E+00
1.99E+02
7.49E+02
1.34E+03
9.68E+02
1.55E+02
1.82E+02
1.50E+02
4.94E+03
2.55E+00
3.34E+01
1.55E+01
7.08E-02
1.55E+01
3.13E+01
3.02E-01
DUBt
1.48E+02
5.88E-01
1.05E+00
1.15E-01
1.05E+01
3.94E+01
7.03E+01
5.10E+01
8.18E+00
9.60E+00
7.92E+00
2.60E+02
1.34E-01
1.76E+00
8.15E-01
3.73E-03
8.13E-01
1.65E+00
1.59E-02
Steel
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
Total
2.96E+03
1.18E+01
2.17E+01
2.55E+00
2.11E+02
7.89E+02
1.41E+03
1.02E+03
1.64E+02
1.92E+02
1.59E+02
5.20E+03
3.01E+00
3.56E+01
1.67E+01
6.17E-01
1.66E+01
3.35E+01
6.03E-01
Cancers , .,
Total •
7.82E-01
8.11E-03
3.67E-03
1.73E-03
8.09E-02
2.11E-01
9.18E-01
2.07E-02
2.59E-02
5.35E-02
4.02E-02
3.41E-01
2.23 E-04
1.29E-0'3
1.24E-03
6.79E-05
1.23E-03
3.57E-03
3.37E-05
PafaE !"
5.49E-01
5.41E-03
2.50E-03
1.16E-03
5.47E-02
1.53E-01
6.15E-01
1.46E-02
1.60E-02
3.46E-02
2.50E-02
2.78E-01
1.97E-04
1.12E-03
1.08E-03
5.88E-05
1.07E-03
2.86E-03
2.92E-05
I
K*
£>.

-------
ES.4   ESTIMATES OF THE MINIMUM DETECTABLE CONCENTRATION (MDC) OF
       RADIONUCLIDES CONTAINED WITHIN OR ON THE SURFACE OF SCRAP
       METAL

The minimum detectable concentration (MDC) for each radionuclide considered in the analysis
must be known in order to be able to assess the feasibility and implementation costs of
potential free release criteria. MDCs, which were derived for a variety of instrument types
and survey techniques, were compared to the radionuclide contamination levels that correspond
to potential release criteria. Tables ES-5 through ES-8 present the MDCs for alternative types
of field survey instruments (alpha, beta, and gamma detectors), alternative survey techniques
(using scans versus stationary counts), and using laboratory analysis of samples instead of field
surveys.

Small areas of contamination can be detected for all but one radionuclide relative to a 15
mrem/y derived concentration limit (DCL)2.  At a DCL corresponding to 1 mrem/y,
detectability drops to 75%; only 25% of the radionuclides are detectable at a DCL
corresponding to 0.1 mrem/y.  Significant improvement is noted when surveying for
distributed (large area) sources of contamination. In this case, one hundred percent of the
radionuclides are detectable at DCLs corresponding to both 15 mrem/y and 1 mrem/y. At a
DCL corresponding to 0.1 mrem/y, almost 70%  of the radionuclides are detectable while
scanning for large areas, while almost 90% could be detected using direct measurements.

The reported values represent ideal conditions, which are not always encountered in the field
(for example,  low background radiation levels and .smooth surfaces  are assumed). In addition,
                                              -*--i
the results are presented for both large and small areas of surface contamination.  An
evaluation of how the MDCs may change under alternative conditions was performed and is
presented in Table ES-9.  Less than ideal conditions could lessen sensitivity for beta and alpha
counting by several fold.  The extent of the loss of detectability is less for scanning for large
areas of contamination than for small, as well as for instances hi which direct measurements
are made. Surrogate methods may be useful in situations where there are multiple
radionuclides present.
       2 A DCL is the radionuclide concentration or surface contamination level in or on scrap metal that
corresponds to a given annual dose.
                                        ES-15

-------
Any assessment of volumetrically contaminated metal by standard field survey techniques is
severely restricted by the limited range of alpha and beta radiation.  Only those radionuclides
with DCLs greater than a few hundred pCi/g can be detected reliably.  However, laboratory
analysis of samples of scrap steel or steel derived from the recycling of scrap metal provides
significantly improved results. State-of-the-art laboratory methods are quite effective at
detecting low levels of volumetric contamination, even down to a few tenths to even
hundredths of a pCi/g.  At levels corresponding to DCLs of 15 mrem/yr and 1 mrem/yr, 100%
of fee radionuclides evaluated can be detected. Even at a DCL corresponding to 0.1 mrem/yr,
85% of the radionuclides are detectable.
                                        ES-16

-------
                       Table ES-5.  Detectability of Radionuclides (Small Area) by Surface Scan* Relative to DCLs
>'R*to&ij&""'
•' •• '*, .' ?i:? :**|
'^:f^''"','mm'4";y:l''
.- ;'Q3?-;'' ' -i" CM'"-.
440 2,500
3,300 23,000
4,300 78,000
3,900 31,000
650 2,800
24,000 610,000
2,300 13,000
1,800 9,600
1,300 6,600
2,900 15,000
31,000 ND
3,600 30,000
3.9E+06 1.5E+07
46,000 ND
1,600 8,000
700,000 ND
8,800 ND
1,100 7,000
4,600 220,000
960 5,600
3,300 22,000
]$lHfcnnmejtt*y"<"'; ';:
,jH';V^$^,.Y'?
:£.. •Q|....:?l:...:.::^./
85 320
ND ND
420 1,600
ND ND
ND ND
420 1,600
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND - ND
440 1,600
430 1,600
420 1,600
ND ND
' ':^;"::
-'•$mfa";-
-?<&& '•
120,000
120,000
340,000
ND
600,000
8.1E+06
200,000
160,000
420,000
140,000
ND
320,000
360,000
1.2E+06
180,000
ND
ND
160,000
860,000
3.9E+06
3.5E+06
' ' ' ' •' A ' t '''''.'''
' '''- r "-fdC'...t:f '."•' ,; ""', r,. '', *:, ' ';;-
« • i5 mr&ls}?' ^ '>,¥)f&e(i&Mt
7,300 Yes
19,000 Yes
9,700 Yes
1.4E+07 Yes
1.3E+06 Yes
17,000 Yes
13,000 Yes
48,000 Yes
130,000 Yes
34,000 Yes
1.8E+09 Yes
15,000 Yes
58,000 No
2.1E+08 Yes
25,000 Yes
2.7E+09 Yes
1.1E+09 Yes
7,700 Yes
4,700 Yes
3,800 Yes
8.3E+07 Yes
; ^^yfaaftWMy .
'f/l' '/'%&$ '::'j.r:^ -&h :'^;rx
'' i'&fKftfy '-'.. l^ect4>Je- '"'' ;,
490 Yes
1,200 No
640 Yes
900,000 Yes
89,000 Yes
1,200 Yes
870 No
3,200 Yes
8,800 Yes
2,300 No
1.2B+08 Yes
990 No
3,900 No
1.4E+07 Yes
1, 700 Yes
1.8E+08 Yes
7.3E+07 Yes
510 Yes
310 No
250 No
5.5E+06 Yes
'- '" •/'"&$'". 'j--f?, Y;"^'i^"-
'•- 0-J 'tetem?^ '/|lteBte-bi^;'|
49 No
120 No
64 No
90,000 Yes
8,900 Yes
120 No
87 No
320 No
880 No
, 230 No
1.2E+07 Yes
99 No
390 No
1.4E+06 Yes
170 No
1.8E+07 Yes
7.3E+06 Yes
51 No
31 No
25 No
550,000 Yes
I
t—>

-J

-------
                  Table ES-5. Detectability of Radionuclides (Small Area) by Surface Scan* Relative to DCLs (Continued)
Radlpauclife
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
Mj
B«a
J <3P ' GW
19,000 460,000
37,000 750,000
19,000 370,000
18,000 ND
23,000 470,000
77,000 390,000
1,100 5,600
840 3,500
2,100 13,000
550 2,300
2,200 12,000
580 2,700
400 2,200
15,000 120,000
17,000 160,000
16,000 180,000
1,200 10,000
680 3,000
120,000 600,000
i'C (dpin/Jlftcm' j
Alpha
GP-'f < &!$ '
420 " 1,600
420 1,600
420 1,600
1.7E+07 6.5E+07
420 1,600
420 1,600
ND ND
ND ND
ND ND
ND ND
ND ND
85 320
85 320
420 ( 1,600
420 %J'* 1,600
420 1,600
430 1,600
420 1,600
ND ND
JJamms
W
8.5E+06
2.2E+06
8.7E+06
6.9E+06
1.1E+07
5.0E+06
370,000
l.OE+06
280,000
8.7E+09
ND
140,000
110,000
1.2E+07
1.5E+07
9.9E+06
160,000
1.2E+06
740,000
for- ; :
"45miemfr VtbxtaW '-'
-17,000 Yes
16,000 Yes
16,000 Yes
l.OE+06 Yes
17,000 Yes
19,000 Yes
64,000 Yes
460,000 Yes
180,000 Yes
40,000 Yes
5.5E+08 Yes
44,000 Yes
13,000 Yes
18,000 Yes
4,100 Yes
37,000 Yes
36,000 Yes
41,000 Yes
120,000 Yes
t>CL SpWtflfcni*)
M :
••' 1 mrerjj/y . :tJe
-------
                      Table ES-6. Detectability of Radionuclides (Large Area) by Surface Scan* Relative to DCLs
•H . • ^ / ,x% ' ,;•
,'„ ^^fl^P*, ,,5
•''* ''•', ''." ' -
Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D
Cm-244
Co-60-
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D
Pa-23f
Pb-210+D
Pm-147
;;:^:< - ;?%W
,^'ai; „,"%$' '?>,.•%,( ,
•"":><$"Jl,{ •'.''<$&??*"
160 790
1,200 7,300
1,500 25,000
1,400 9,800
230 890
8,400 193,000
830 4,100
640 3,000
480 2,100
1,000 4,800
11,000 ND
1,300 9,500
1.4E+06 4 8E+06
16,000 ND
570 2,600
250,000 ND
3,100 ND
370 2,200
1,600 70,000
340 1,800
1,200 6,900
j<^d£*ji^i?);;, -f
'^''l. $&/,„%,
- ^tM... 1\LLM[
18 €8
ND ND
90 340
ND ND
ND ND
89 330
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
92 '350
91 340
89 330
ND ND
..l:M:.Il!
270
270
770
ND
1,400
19,000
470
370
960
320
ND
730
820
2,800
410
ND
ND
360
2,000
9,000
7.9B+06
: "r''.::. ' '-r : H" :
1 > ' tftjtl 1 ! A , ft •ft/' ''t 'ft ff
' 15 mffiip/f'' ... . fietee&bte ' • •
7,300 Yes
19,000 Yes
9,700 ' Yes
1.4E+07 Yes
1.3E+06 Yes
17,000 Yes
13,000 Yes
48,000 Yes
130,000 Yes
34,000 Yes
1.8E+09 Yes
15,000 Yes
58,000 Yes
2.1E+08 Yes
25,000 Yes
2.7E+09 Yes
1.1E+09 Yes
7,700 Yes
4,700 Yes
3,800 Yes
8.3E+07 Yes
>,.: ', '^^
-------
                Table ES-6.  Detectability of Radionuelides (Large Area) by Surface Scan* Relative to DCLs (Continued)
RMloiHWlMs
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Eu-106+D
Sb-125+D
Sr-90+D
Te-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-6S
Pi
8M» -
.... Q? ...*.* 'M
6,800 150,000
13,000 240,000
6,800 120,000
6,400 ND
8,300 150,000
27,000 120,000
380 1,800
300 1,100
730 4,000
190 730
800 3,900
210 860
140 710
5,200 37,000
6,000 50,000
5,600 57,000
410 3,200
240 970
42,457 190,000
iCtivmfW&at'J
Aljte
0? 2&
89 340
89 340
89 340
3.7B4-06 1.4E+07
89 • 340
89 330
ND ND
ND ND
ND ND
ND ND
ND ND
18 67
18 67
89 340
89 340
89 340
91 340
89 340
ND ND
Gamma.
Jlaf
20,000
51,000
20,000
1.6E+07
24,000
12,000
860
2,400
650
2.QE+07
ND
310
250
28,000
34,000
23,000
360
2,800
1,700
for "''
• JSratsoi/y- DstestaMs ;
17,000 Yes
16,000 Yes
16,000 Yes
l.OE+06 Yes
17,000 Yes
19,000 Yes
64,000 Yes
460,000 Yes
180,000 Yes
40,000 Yes
5.5E+08 Yes
44,000 Yes
13,000 Yes
18,000 Yes
4,100 Yes
37,000 Yes
36,000 Yes
41,000 Yes
120,000 Yes
JXJL('
fOf
i carere/y Usteclatfe
1,200 Yes
1,100 Yes
1,100 Yes
67,000 Yes
1,100 Yes
1,300 Yes
4,300 Yes
30,000 Yes
12,000 Yes
2,600 Yes
3.6E+07 Yes
2,900 Yes
900 Yes
1,200 Yes
280 Yes
2,500 Yes
2,400 Yes
2,700 Yes
8,100 Yes
4 "" , ' " _ ~
O.i mr«Bi& ' BetekaWe '
120 Yes
110 Yes
110 Yes
6,700 Yes
110 Yes
130 Yes
430 Yes
3,000 Yes
1,200 Yes
260 Yes
3.6E+06 Yes
290 Yes
90 Yes
120 Yes
28 No
250 Yes
240 Yes
270 Yes
810 No
s
       * Meter time constant = 10 s

-------
                            Table ES-7. Detectability of Radionuelides by Direct Count* Relative to DCLs
" "'ISllsiik'ite' '-1
. •.••:.•« 'A -v-l
..•..•..j?..v.?....\4.:: 	 ;.,.,
Ac-227+D
Ag-llOm-fD
Ara-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D
Pa-231
Pb-210+D ,
Rm-147
......... ..^
••"^•'^:^^
' , •• : .•"••,„'* " -S
,,:QP.v..., U . df... •;
70 260
560 2,600
690 8,000
620 3,200
100 290
3,800 63,000
370 1,300
290 990
210 680
460 1,600
4,900 ND
580 3,100
620,000 1.6E+06
7,400 ND
260 830
' 110,000 ND
1,400 ND
170 720
730 23,000
150 580
520 2,300
;tt$ftlttt»cf*) >" ' , •
.,,. ,', , Apf'j ' c
^±:M.?..L'^.!.™]^.r*".
6 18
ND ND
32 91
ND ND
ND ND
32 90
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
33 93
33 92
32 90
ND ND
'•,"'•, •" '"sf
; &&$i&v|
:-_.M£
91
93
270
ND
480
6,400
160
130
330
110
ND
250
280
970
140
ND
ND
120
680
3,100
2.7E+06
• , X * | f • £ ' ,* j"" &A£V&<
!;^l 1 ',, ; - 'M
|S msei^f . ] ^.Oa^8!alfe ;'«'. .
7,300 " Yes
19,000 Yes
9,700 Yes
1.4E+07 Yes
1.3E+06 Yes
17,000 Yes
13,000 Yes
48,000 Yes
130,000 ' Yes
34,000 Yes
1.8E+09 Yes
15,000 Yes
58,000 Yes
2.1E+08 Yes
25,000 Yes
2 7E+09 Yes
1.1E+09 Yes
7,700 Yes
4,700 Yes
3,800 Yes
8.3E+07 Yes
/>•", -m$M»WK''T
^•:i&"«* ™-tr;';-rv ,
..:.,;.. £.BfftB{B/^ t ••' U^ClaJ)^.. . .,,
490 Yes
1,200 Yes
640 Yes
900,000 Yes
89,000 Yes
1,200 Yes
870 Yes
3,200 Yes
8,800 Yes
2,300 Yes
1.2E+08 Yes
990 Yes
3,900 Yes
1.4E+07 Yes
1,700 Yes
L8E+08 Yes
7.3E+07 Yes
510 Yes
310 Yes
250 Yes
5.5E+06 Yes
"'' ' i'.X/t'V % •" ' ' '
',&* /i >*> *; J.v' x?
'.• 8,1 m&w$ ,- :JEJBi$cfstilB'..A"
49 Yes
120 Yes
64 Yes
90,000 Yes
8,900 Yes
120 Yes
87 No
320 Yes
880 Yes
230 Yes
1.2B+07 Yes
99 No
390 Yes
1.4B+06 Yes
170 ' Yes
1 8E+07 Yes
7.3E+06 Yes
51 Yes
31 No
25 No
• 550,000 Yes
w
OT3

tb

-------
                      Table ES-7. Detectability of Radionuclides by Direct Count* Relative to DCLs (Continued)
i
RadiQiarcllfc.
j
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228-fD
Ru-106+D
Sb-125-fD
Sr-90-fD
Tc-99
Th-228+D
Th-229-fD
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
Mi
6«a
t & GM ,
3,100 47,000
5,900 78,000
3,100 39,000
2,900 ND
3,700 49,000
12,000 41,000
170 580
130 360
330 1,300
87 240
360 1,300
93 280
63 230
2,300 12,000
2,700 16,000
2,500 18,000
190 1,000
110 320
19,000 63,000
iCCdpiJ/JIKfem^
Alpha
<3t» 2fa$
32 90
32 90
32 90
1.3E+06 3.7E+06
32 90
32 90
ND ND
ND ND
ND ND
ND ND
ND ND
6 18
6 18
32 90
32 90
32 90
33 92
32 90
ND ND
;: '
for -, '%"";
s; ' y.}*.,':,. 7-' ,
0,J mteonfy, v .Deteftafala
120 Yes
110 Yes
110 Yes
6,700 Yes
110 Yes
130 Yes
430 Yes
3,000 Yes
1,200 Yes
260 Yes
3.6E+06 Yes
290 Yes
90 Yes
120 Yes
28 No
250 Yes
240 Yes
270 Yes
810 Yes
10
      * Count time = 60 s

-------
                    Table ES-8. MDC of Radionuclides by Laboratory Analysis Compared to DCLs
10
RMomjeliao
Ag-llOm
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Ni-63
Pb-210+D
Pm-147
Pu-238
Pu-239
Pu-241
Ra-226+D
Ra-228+D
Ru-106+D
Sb-125
Sr-90+D
Tc-99
Th-228+D -
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
MDC (pCi/g)
0.04
0.05-0.5
0.2 - 37
0.23
0.05-0.1
0.01-0.3
0.02-0.4
0.007-0.3
0.02-0.9
1-30~
0.4-2
0.2-0.3
1-100
0.1-5
0.5-5
0.02 - 0.4 %
0.02 - 0.4
0.02 - 20
0.02 - 0.7
0.1-2
0.2-1
0.11
0.03 - 5
0.3 - 15
0.05-0.4
0.05 - 0.5
0.05 - 2
0.05 - 0.2
0.02-0.3
0.02-0.1
0.09 - 0.6
for
15 'mr&Aly Detectable
24 Yes
5.5 Yes
l.OE+06 Yes
310 ' Yes
- 9.9 Yes
17 Yes
180 Yes
490 Yes
13 Yes
2 3E+6 Yes
8,400 Yes
22 Yes
1.4E+06 Yes
41 Yes
5.5E+05 Yes
9.0 Yes
8.4 Yes
530 Yes
7.7 Yes
14 Yes
290 Yes
670 Yes
130 Yes
8.3E+05 Yes
4.1 Yes
9.8 Yes
2.2 Yes
20 Yes
18 Yes
22 Yes
560 Yes
DCL (pCi/g)
for ;
1 mrem/y : Detectable
1.6 Yes
0.37 Yes
6.9E+04 Yes
21 Yes
0.66 Yes
1.1 Yes
12 Yes
32 Yes
0.86 Yes
1.5E+05 Yes
560 Yes
1.5 Yes
9.4E+04 Yes
0 27 Yes
J.600 Yes
0.60 Yes
0.56 . Yes
35 Yes
0.51 Yes
0.96 Yes
19 Yes
44 Yes
8.4 Yes
5.5E+04 Yes
0 28 Yes
0.65 Yes
0.15 Yes
1.3 Yes
- 1.2 Yes
1.4 Yes
37 Yes
. fot ' '
" 0,1 ttfem)y ,' , Detectable
. 0.16 Yes
0.037 No
6,900 Yes
2.1 Yes
0.066 Yes
0.11 Yes
1.2 Yes
3.2 Yes
0.086 Yes
1.5E+04 Yes
56 - Yes
0.15 Yes
9,400 Yes
0027 No
360 Yes
0.060 Yes
0.056 Yes
3.5 . Yes
0.051 Yes
0.096 Yes
1.9 Yes
4.4 Yes
0.84 Yes
5,500 Yes
0.028 No
0.065 Yes
0.015 No
0.13 Yes
0.12 Yes
0.14 Yes
3.7 Yes

-------
       Table ES-9. Relative Range in MDCs*
Survey Mode
Direct measurement
Scan - small area source
Scan - large area source
Beta
1 -3
1 -7
1 -4
Alpha
1-5
1-14
1 -7
Gamma
1 -7
0.01 - 1
1-8
* The values are multipliers to be applied to the MDCs.
                      ES-24

-------
                                   REFERENCES
DOE 95      U.S. Department of Energy, "Gaseous Diffusion Facilities Decontamination and
             Decommissioning Estimate Report," prepared by G.A. Person, et al,
             Environmental Restoration Division, Oak Ridge, TN for U.S. Department of
             Energy, Office of Environmental Management, ES/ER/TM-171, December 1995.

DOE 96      U.S. Department of Energy, "Taking Stock:  A Look at the Opportunities and
             Challenges Posed by Inventories from the Cold War Era," U.S. Department of
             Energy, Office of Environmental Management, DOE/EM-0275, January 1996.

EPA 95      U.S. Environmental protection Agency, "Analysis of the Potential Recycling of
             Department of Energy Radioactive Scrap Metal," prepared by S. Cohen &
             Associates, Inc. for the U.S. Environmental Protection Agency, Office of
             Radiation and Indoor Air, August 1995.

NRC 79      U.S. Nuclear Regulatory Commission, "Technology, Safety and Costs of
             Decommissioning a Reference Pressurized Water Reactor Power Station," Vol. 1,
             NUREG/CR-0130,1978, prepared by Smith, R.I., et al., Pacific Northwest
             Laboratory for the U.S. Nuclear Regulatory Commission.

NRC 80      U.S. Nuclear Regulatory Commission, "Technology, Safety and Costs of
             Decommissioning a Reference Boiling Water Reactor Power Station," Vol. 2,
             Appendices, NUREG/CR-0672,prepared by Oak, H.D., et al., Pacific Northwest
             Laboratory for the U.S. Nuclear Regulatory Commission, 1980.

NRC 94      U.S. Nuclear Regulatory Commission, "Revised Analyses of Decommissioning
             for the Reference Boiling Water Reactor Power  Station," Vol 2, Appendices,
             NUREG/CR-6174, prepared by Smith, R.I., et al, Pacific Northwest Laboratory
             for the U.S. Nuclear Regulatory Commission, 1994.

NRC 95      U.S. Nuclear Regulatory Commission, "Revised Analyses of Decommissioning
             for the Reference Pressurized Water Reactor Power Station," Vol 1, Main Report,
             NUREG/CR-5884, prepared by Konzek, G.J, et  al., Pacific Northwest Laboratory
             for the U.S. Nuclear Regulatory Commission, 1995.
                                       ES-25

-------
Page Intentionally Blank

-------
                                     CHAPTER 1

                                   INTRODUCTION

1.1    PURPOSE

The Office of Radiation and Indoor Air (ORIA) of the U.S. Environmental Protection Agency
(EPA) is evaluating the potential for recycling scrap metal from nuclear facilities. The clean up
of sites that are contaminated with radioactive material and the decommissioning of nuclear
facilities is expected to generate large amounts of scrap metal. In fact, some sites controlled by
the U.S. Department of Energy (DOE) already have accumulated significant inventories of scrap
metal. EPA is considering the possible impacts of recycling scrap metal from nuclear facilities
as an alternative to disposing of it in a licensed, low-level radioactive waste disposal facility.

As part of its evaluation, EPA has examined current recycle practices, existing regulatory
guidance governing these practices, the inventory of contaminated scrap metal potentially
available for recycle, and the possible radiological impacts associated with recycling such
materials.  This document summarizes the technical information used by EPA in its evaluation.
In a separate document, "Radiation Protection Standards for Scrap Metal: Preliminary Cost-
Benefit Analysis," (EEC 97) EPA describes its analysis of the potential costs and benefits of
recycling scrap metal from nuclear facilities.

Based on the information provided in this Technical Support Document (TSD), as well as the
results of the cost-benefit analyses, EPA will decide whether recycling scrap metal from nuclear
facilities is viable and whether additional regulatory action is necessary to ensure that release of
such materials does not endanger public health and safety.

1.2    SCOPE OF THE ANALYSIS

The information presented in the TSD focuses on scrap metals that are suspected to be
moderately contaminated with radioactivity as a result of deposition or neutron activation
(referred to throughout this document as "scrap metal").  Consequently, scrap metal from nuclear
facilities which is not contaminated, such as that generated from locations outside radiation
control areas, is not considered in EP A's evaluation. Conversely, some metals may be so
contaminated with radioactivity, due to their use during the life of the facility, that they can only
                                          1-1

-------
be disposed of as a radioactive waste and are therefore ineligible for recycling.  Such scrap is
also excluded from this evaluation.

The principal sources of scrap metal evaluated are the DOE complex of weapons-production and
research facilities and commercial nuclear power plants.  Each of these sources has the potential
to produce large quantities of scrap metal. Department of Defense (DoD) and industrial facilities
were not explicitly addressed in the analysis because they are relatively small sources of
recyclable scrap metal.

Although many types of metals (including steels, aluminum, lead, copper, nickel, and precious
metals) are used in nuclear facilities and may be available for recycling, EPA's evaluation of the
impacts of recycling scrap metal to date has been limited to carbon steel. Carbon steel represents
the largest quantity of metal potentially available for recycle from nuclear facilities. In addition,
the annual utilization of carbon steel scrap in the United States is on the order of 70 million tons,
whereas the annual utilization of aluminum, stainless steel, and copper scrap is  about one million
tons of each type of metal. However, the TSD does provide estimates of the amount of these
other metals 'that might be available for recycling from nuclear facilities.

EPA also recognizes that recycling other metals may be economically significant and hence
desirable to the nuclear industry.  Therefore, characterization of other metals and their potential
for recycle is currently being conducted. In addition, the Agency is continuing to improve its
database for certain factors affecting the evaluation of the impacts of recycling carbon steel.

The TSD addresses a number of issues that  are important to assessing the impacts of recycling
scrap metal from nuclear facilities.  The following list represents the steps EPA took in its
analysis of these impacts. The information  resulting from each of these steps is critical to the
Agency's decision making regarding the need for regulatory action concerning release of scrap
metal from nuclear facilities.

•      Assessment of the available information on recycling scrap metal and identification of
       additional information needs.

•      Screening of the diverse and complex information base on scrap metal to select the scope
       and content of information to be used in EPA's analyses.
                                          1-2

-------
•      Characterization of scrap quantities, levels of contamination, and options for disposition.

•      Selection of the conditions for radioactive contamination to be considered in the analyses.
                             i                         i
•      Selection and characterization of scenarios for individual and collective radiation
       exposures in activities including scrap generation and management, steel manufacture,
       and use of products made from recycled metal.

•    ,  Development and use of models and computer codes to evaluate individual and collective
       doses and risks from radiation exposure scenarios.

•      Development of data to numerically characterize parameters used in the radiation
       exposure scenarios!

•      Development of numerical values of normalized doses for each radionuclide of concern.

•      Review of methods and their limitations for detection and measurement-characterization
       of radioactive contamination on or in scrap metal.

•      Identification of factors that affect the potential for dilution and dispersion of
       radionuclides to various material streams, such as metal, slag, and dust.
1.3    ORGANIZATION OF THE TSD

The TSD is organized into three volumes. The first volume, which is comprised of 10 chapters
and an executive summary, closely follows the methodology used in the analysis and described
in Section 1.2 above.  Chapter 2 of the TSD provides an overview of scrap metal operations in
the United States and the characteristics of scrap metal potentially available for recycle from
nuclear facilities. Chapter 3 describes the screening procedures used by EPA to select the scope
of analyses to be conducted. This chapter also discusses the limitations of these analyses.
Chapter 4 describes the principal sources of scrap metal which include the DOE complex and the
commercial nuclear power industry.

Chapters 5, 6, and 7 present the basis for and results of the EPA's risk assessment for
individuals.  Chapter 5 describes the exposure pathways associated with unrestricted recycling of
scrap metal from nuclear facilities. Chapter 6 describes the calculation of the radiological
                                          1-3

-------
impacts associated with unrestricted recycling.  Chapter 7 presents and discusses the results of
the dose and risk calculations for unit radionuclide concentrations and throughput

Chapter 8 evaluates current methods of detecting and measuring radioactivity in or on scrap
metal. Chapter 9 discusses the time integrated collective dose and presents the results of
calculations of potential health effects for unit radionuclide throughput for unrestricted recycle.
Chapter 10 concludes the TSD with a discussion and semi-quantitative analysis of the
sensitivities and uncertainties associated with the Agency's evaluation of the impacts associated
with recycling scrap metal from nuclear facilities.

Volumes 2 and 3 of the TSD are comprised of a number of appendices that provide more detailed
information regarding the potential sources of scrap metal, the models used to  derive individual
and collective doses and risks, the results of the analyses, key modeling assumptions, and the
uncertainties in the data and assessments found in this report.
                                           1-4

-------
                                   REFERENCE
IEC 97       Industrial Economics, Inc., "Radiation Protection Standards for Scrap Metal:
             Preliminary Cost-Benefit Analysis," prepared for the EPA Office of Radiation and
             Indoor Air, under Contract No. 68-DO-0102, Work Assignment Manager Reid
             Harvey, 1997.           "   '
                                        1-5

-------
Page Intentionally Blank

-------
                                     CHAPTER 2

                     OVERVIEW OF SCRAP METAL OPERATIONS

This chapter provides an overview of the types ancLquantities of scrap metal potentially available
for recycle, the operations of the scrap metal industry in the United States, and current practices
for recycling scrap metal from nuclear facilities. It provides .a summary description of the world
of scrap metal generation and utilization; any scrap metal released from nuclear facilities for
recycling becomes a part of this process. This chapter offers a perspective on the data, modeling,
parameter characterizations, and associated technical issues considered by EPA in its evaluation.

A comprehensive discussion of the scrap metal industry in the United States is provided in the
report, "Analysis of the Potential Recycling of Department of Energy Radioactive" Scrap Metal"
(SCA 95). This four-volume report evaluated the potential for recycling scrap metal from the
DOE complex. The discussion in this  chapter presents highlights from that report which are
relevant to EPA's current analysis which addresses the potential for recycling metals from a
broader number of sources.  This chapter also presents updated summary estimates of scrap metal
availability from nuclear facilities.

2.1    CHARACTERISTICS OF SCRAP SOURCES

Scrap metal released from nuclear facilities would become part of the ongoing scrap metal
industry in the United States. The characteristics of the industry, in terms of factors such as
annual production, vary from metal  to  metal.  Comprehensive descriptions of the carbon steel,
stainless steel, aluminum, copper, lead, and nickel scrap metal industries are included in SCA 95.
Discussions of recycling aluminum and copper scrap are presented in Appendices B and C,
respectively, of this TSD.

The DOE complex and decommissioned nuclear power reactors could generate scrap metals of
many types. Possibilities include carbon steel, stainless steel, galvanized iron, copper, Inconel,
lead, bronze, aluminum, brass, nickel,  and precious metals, such as gold and silver. As discussed
hi Chapter 1, EPA has limited its current evaluation to the impacts associated with recycling
carbon steel, although estimates of the available inventories of these other metals were made.
                                          2-1

-------
 Impacts of recycling scrap metal from nuclear facilities must be evaluated for specific
 circumstances. Potential impacts will depend on the quantities, rates, and timing of the recycling
 activity and on the characteristics of the scrap metal industry for the particular metals of concern.
 To support its analysis, EPA has developed estimates of the availability of potential sources of
 scrap from the DOE complex and the commercial nuclear power plants in terms of quantities
 available, contamination characteristics, and timing of the metal's availability for recycle.

 There is at present great uncertainty in the future availability of scrap for recycle from nuclear
 facilities.  In terms of the DOE complex, uncertainty exists as to when and how decom-
 missioning of these facilities and equipment might occur. Uncertainty also exists in terms of the
 timing of commercial reactor shutdowns and the associated decommissioning activities. Reactor
 licenses, most of which are valid for 40 years, might be renewed and dismantlement of the
 reactors might not occur for as long as 60 years after shutdown. Alternatively, some reactors are
 being shut down and decommissioned before their present operating licenses expire.  Given these
 uncertainties, it is possible that the generation of scrap metal that could be recycled from existing
 DOE and reactor sources will span the next century.

•As mentioned above, EPA estimated the potential inventory of scrap metal available for recycle.
 A comprehensive assessment of scrap metal generated by commercial power reactors was
 performed and a full report on this evaluation can be found in Appendix A of the TSD.  An
 assessment of the available information on scrap metal inventories at DOE sites was also
 conducted. Numerous snapshot assessments of scrap metal inventories at DOE sites have been
 done. These assessments are discussed in Chapter 4 of the  TSD. Actual tables of data derived
 from them are included in IEC 97 as part of the cost-benefit analysis.

 DOE recently issued a comprehensive report on materials in inventory, including scrap metal
 (DOE 96a).  This report presents a snapshot of scrap metal  inventories during approximately the
 summer-of-1995 period.  The inventories were characterized as clean (no radiological
 contamination), contaminated (known radiological contamination), and "unspecified"
 (potentially contaminated) scrap metals.  These inventories are subject to large and rapid changes
 as a result of ongoing operations such as the sale of clean scrap, disposal of radioactive wastes,
 and batch production of scrap from decommissioning of structures. This DOE report does not
 include projections of scrap generation and availability since plans for future cleanup and
 decommissioning activities are highly uncertain.
                                          2-2

-------
The characterization of scrap metal inventories at DOE sites is further complicated by the use of
different definitions of "scrap" at various DOE sites. For example, inventory estimates reported
in DOE 96a include approximately 120,000 tons of carbon steel scrap associated with the
uranium enrichment facilities at the Y-12, K-25, Paducah, and Portsmouth facilities. Other DOE
reports note that these same sites contain on the order of 350,000 tons of carbon steel not yet
declared scrap and therefore not included hi current scrap inventories. These reports do not
indicate when this metal might be declared scrap and made available for recycle. Recent
developments in DOE policy concerning re-industrialization suggest that facilities at these sites
might be converted to new uses so that potential scrap production would be limited.

2.2    INDUSTRY PERSPECTIVES

To gain perspective on the potential significance of recycling scrap from nuclear facilities to the
scrap metal industry, EPA compared the annual industry throughput with its  estimates of the
potential inventories of metals available for recycling from DOE sites and commercial reactors.
For carbon steel, SCA 95 shows that the annual production of iron and steel from scrap is on the
order of 68 million tons. Table A4-4, (in Appendix A of the TSD), indicates that the total
amount of carbon steel that could be generated from the decommissioning of all commercial
nuclear reactors is about four million tons.  (This number is derived from the 3.6 million metric
tonnes shown in Table A4-4.)  On the basis of potential for radioactivity contamination under
service conditions, it is estimated thai about 487,000 tons of this total would be scrap carbon
steel that could be available for recycle over a 50-year period.

DOE 96a indicates, in Table 2-9, that the total current inventory of contaminated and
"unspecified" carbon steel at DOE sites is 116,000 tons and the HAZWRAP report (HAZ 95)
provides supplemental information which brings the estimate of the existing  DOE inventory to
about 171,000 tons (see Table 4-3 of the TSD). As noted above, the enrichment facilities contain
another 350,000 tons of carbon steel which may be declared to be scrap. The inventory of
potentially available carbon steel scrap from other DOE sites is small in comparison to the
enrichment facilities since most of the material generated at these other sites will be ineligible for
recycle because of relatively high levels of contamination.

In summary, the upper bound of carbon steel  scrap generated by DOE and  commercial reactor
sources and available for free release and recycle is on the order of a few million tons.  This scrap
                                          2-3

-------
would be released to the scrap metal industry over a period of decades. Therefore, the total
quantity of carbon steel scrap entering the industry is small in comparison to its annual
throughput of 68 million tons.

The annual scrap industry throughput for stainless steel is approximately 1.1 million tons (SCA
95). The amount of stainless steel potentially available for recycle from nuclear reactors is on the
order of 122,000 tons (see Table 5A-4).  DOE 96a indicates an on-hand stainless steel scrap
inventory of about 7,000 tons. Similar to carbon steel, the quantities of stainless steel potentially
available for recycle from nuclear facilities (about 130,000 tons) are therefore also small hi
comparison to the annual throughput of the industry.

Quantities of other types of scrap metals potentially available from nuclear facilities, such as
aluminum and copper, are also small hi comparison with the annual industry throughput.
However, these comparisons do not necessarily indicate that the potential radiological impacts of
recycling this material are small—as noted above, the potential radiological impacts of recycling
scrap metal released from nuclear facilities must be evaluated for specific scenarios within the
scrap metal industry.  As discussed in Chapter 1, the TSD describes hi detail the impacts of
recycling carbon steel generated by nuclear facilities.  EPA will continue to evaluate the
consequences of recycling other types of scrap metals  from these facilities.

2.3    PRINCIPAL SCRAP METAL OPERATIONS CONSIDERED

Operations involved hi the generation, management, and utilization of scrap metals that could
result hi human radiation exposures include:

•      Activities at the scrap-generation site, including demolition of structures or
       decommissioning of equipment, on-site management of scrap piles, characterization of
       contamination, decontamination of scrap, and disposal of radioactive wastes, including
       scrap metal that is not available for recycle.

•      Transport of scrap metal to scrap processor site.

•      Operations at the scrap processor site, such as shredding and preparation for melting.

•      Transport of the prepared scrap to a mill.
                                          2-4

-------
•      Operations at a mill, including management of recycled scrap, furnace operations,
       handling of interim products, such as ingots, bag house operations, and slag disposal.

•      Manufacture, transport, and sale of finished products.

•      End use of finished products, such as automobiles, cookware, and industrial equipment.

As detailed in Chapter 5, human exposure scenarios are defined and characterized for each of
these operations.

There are about ,160 steel mills in the United States; theoretically, any could receive and process
scrap metal generated by nuclear facilities. However, in practice, these mills maintain close
relationships with nearby scrap metal dealers in order to minimize transportation costs. In turn,
the scrap metal dealers receive their materials from close-by sources, again to minimize
transportation costs.

It is to be expected, therefore, that nuclear facility sources of scrap metal, such as DOE sites and
decommissioned commercial power reactors, will send their scrap to nearby scrap dealers who in
turn will respond to orders from the mills they serve. As shown in SCA 95, each of the DOE
sites that could be a significant source of scrap metal has proximate mills with electric arc
furnaces and, by implication, near-by scrap dealers who could receive and process the recycled
scrap.  Scrap dealers and mills are also located in the vicinity of most nuclear power plants.

The capacities of individual electric arc furnaces vary widely, ranging from about 15 to 225 tons
of steel per heat. The steel mills in the United States have in total about 2<60 electric arc
furnaces; virtually all of the metal charged to these furnaces is scrap.

Because of the variations in furnace capacities, the characteristics of the working relationships
between scrap dealers and mills, and the make up of individual charges to a furnace, evaluating
the potential radiation doses and risks must be based on specific scenario parameters. For
example, it is possible that a single charge to a small furnace could be made up entirely of scrap
from a nuclear facility source, in which case any residual radioactivity in the scrap metal would
be undiluted. Alternatively, it is possible that a single charge to a furnace could be a mixture of
scrap from a nuclear facility, and scrap from other sources. In this case, any residual
radioactivity in the scrap from nuclear facilities would be diluted. This variation hi the
                                            2-5

-------
characteristics of the furnace charge becomes significant when considering the characteristics of
the products made from the metal produced.

The levels of residual radioactivity in scrap that become part of the intermediate or finished
products will also depend on partitioning that occurs during furnace operations. As a result of
chemical phase equilibrium phenomena, radioactive and other species will distribute among the
metal-melt, slag, and vapor phases associated with melting operations. The various elements
partition among the phases in different ways; therefore, the evaluations of doses and risks for
various workers, such as bag house and slag pile workers, must account for these differences.
Partitioning during melting of carbon steel is discussed hi Appendix E of the TSD; partitioning
during melting of cast iron is discussed in Appendix F.

In summary, evaluation of radiation doses and risks associated with recycle of scrap from nuclear
facilities must take into account a wide range of scrap management, operations, and practices.  In
addition, such an evaluation must consider a range of specific conditions, such as partitioning of
radionuclide species during melting and the characteristics of an individual furnace charge.
These factors will affect who is exposed and to what level of radioactivity. Given the importance
of these factors, it is inappropriate to base evaluations on averages—the details of specific
scenarios must be examined hi order to rally evaluate the impacts of recycling scrap metal from
nuclear facilities.

2.4    CURRENT RECYCLE PRACTICE FOR NUCLEAR FACILITIES

Recycle of scrap metal from nuclear facilities is currently practiced on a limited and directed
basis. For example, specialty metals, with low levels of contamination, that are generated at one
DOE site are given new use at another site and scrap is converted into containers used for
radioactive waste disposal. Despite these initiatives, the general rule of thumb for management
of scrap for both the DOE complex and facilities licensed by the U.S. Nuclear Regulatory
Commission (NRC) is to dispose of it hi a licensed, low-level waste disposal facility. However,
the interest in recycling scrap metal from the DOE complex is growing. In September 1996,
DOE Assistant Secretary Aim issued a policy encouraging the recycle-of scrap metal from DOE
facilities. Specifically, the policy directs "the release for unrestricted use [of] any material that
meets applicable criteria"  The policy goes on to state that "if decontamination for release for
unrestricted use is not economically feasible, then the [carbon steel] that is recycled shall be
fabricated into one-time-use containers for disposal of low-level wastes...." (DOE 96b).

                                          2-6

-------
Current recycle practices are governed by DOE and NRC.  The NRC established, in Regulatory
Guide (RG) 1.86, residual surface contamination criteria for facilities being decommissioned for
unrestricted release. DOE in large measure adopted the NRC criteria in its basic radiation pro-
tection standards, DOE Order 5400.5. These criteria are discussed hi more detail hi Chapter 8.

It is important to note that RG 1.86 does not take into consideration potential health effects
associated with recycle and re-use of materials released under the criteria. The RG 1.86 criteria
were based on maximum permissible concentrations for ah- and water listed in 10 CFR Part 20
regulations and on the assumption that licensees should not be expected to reduce surface
concentrations to levels below existing environmental background levels caused by fallout from
atmospheric testing of nuclear devices (NRC 94). The estimated average dose for the 24
nuclides evaluated under the RG 1.86 criteria is about 10 millirem/year (mrem/yr).
                                           2-7

-------
                                   REFERENCES
DOE 96a     U.S. Department of Energy, "Taking Stock: A Look at the Opportunities and
             Challenges Posed by Inventories from the Cold War Era," January 1996.

DOE 96b     U.S. Department of Energy Memorandum, A. Aim to Distribution, Subject:
             Policy on Recycling Radioactively Contaminated Carbon Steel, September 20,
             1996.

HAZ 95      "U.S. Department of Energy Scrap Metal Inventory Report for the Office of
             Technology Development, Office of Environmental Management," prepared by
             Hazardous Waste Remedial Actions Program for the Department of Energy,
             DOE/HWP-167, March 1995.

IEC 97       Industrial Economics, Inc., "Radiation Protection Standards for Scrap Metal:
             Preliminary Cost-Benefit Analysis," prepared for the EPA Office of Radiation and
             Indoor Air, under Contract No. 68-DO-0102, Work Assignment Manager Reid
             Harvey, 1997.

NRG 94      U.S. Nuclear Regulatory Commission, Secy-94-145, May 27,1994.

SCA 95      S. Cohen & Associates, "Analysis of the Potential Recycling of Department of
             Energy Radioactive Scrap Metal," prepared for the EPA Office of Radiation and
             Indoor Air under Contract No. 68D20155, Work Assignment 3-19, Work
             Assignment Manager John MacKinney, August 14,1995.
                                        2-8

-------
                                      CHAPTERS

        SCREENING PROCEDURES TO DEFINE THE SCOPE OF THE ANALYSIS

In evaluating the potential for recycling scrap metal from nuclear facilities, EPA was confronted
with a vast array of issues.  How much scrap metal is available? What levels of radiation
contamination is it likely to have? How will it be handled?  Which individuals may be at risk
from exposure to the scrap  as it moves through the recycling process? In order to focus its
evaluation, EPA had to refine its objectives and develop a method of screening potential issues
and data to determine those which were most significant to the analysis.
                                                                                     /

The objectives of EPA's analysis included characterizing the different types of existing and
future sources of scrap metal from nuclear facilities and the levels of radiological contamination
in or on the scrap.  The Agency's analysis also defines the relationship between this
contamination and the radiation doses to individuals and populations that may result from the
free release of the scrap metal. This chapter presents the screening methods that were used to
select the parameters of EPA's evaluation, including: specific scrap metal sources, types of
metals, radionuclides, exposure scenarios and pathways, and potential types of adverse biological
effects that may be associated with the free release of scrap metal from nuclear facilities.

This chapter is divided into seven sections. Section 3.1 provides background information on the
specific areas of inquiry and analysis contained in the TSD.  This section establishes the context
within which the information regarding source, type, and radionuclide composition of scrap
metal is used in the TSD. Section 3.2 describes the overall screening criteria used to select the
specific sources, metal types, and radionuclides explicitly characterized and analyzed in the TSD.
Section 3.3 describes the wide variety of existing and potential sources of scrap metal at nuclear
facilities.  This section also describes the methods used to screen these sources down to a
manageable number without overlooking any potentially important source. Section 3.4 describes
the different types of scrap  metals (e.g., carbon steel, stainless steel, galvanized iron, copper,
aluminum, etc.) potentially available for recycle and the methods used to select the specific types
of metals explicitly addressed in the TSD.  Section 3.5 describes the methods used to select the
radionuclides of primary concern in the TSD. Section 3.6 presents the screening methods used to
select the specific exposure scenarios, pathways, and biological endpoints addressed in this
report. Finally, Section 3.7 summarizes the results of the screening process and the limitations of
the TSD due to the constraints placed on the scope of the analyses provided in this report.
                                          3-1

-------
3.1    PRIMARY PURPOSE OF THE TSD


The TSD provides technical background information for the Agency to use in its evaluation of
the potential for recycling scrap metal from nuclear facilities and the need for regulatory action.
The information contained in this document also supports the cost-benefit analyses conducted by
EPA as part of tMs evaluation. The TSD may also serve as the technical basis for a Regulatory
Impact Analysis (RIA) done by EPA to support a future rulemaking concerning the recycling of
scrap metal from nuclear facilities.


The TSD includes the following background information:


       1.     Characterization of the potential sources of scrap metal that may be affected by a
             future EPA rule addressing the free release of scrap metal from nuclear facilities.

             The management and disposition of scrap metal that has been generated to date,
             and that will be generated in the future by nuclear facilities in the U.S., is
             currently governed primarily by DOE and NRC regulations and guidelines that do
             not apply specifically to the free release of scrap metal from these facilities.  As
             such, decision making regarding the disposition of scrap metal has not been based
             on a single set of comprehensive national standards. Should EPA establish such
             standards, behavior regarding the disposition of scrap metal from nuclear facilities
             may change,  resulting in certain costs and benefits that need to be characterized,
             As a first step in gaining insight into these potential costs and benefits (which are
             addressed in detail in IEC 97), an understanding is needed of the sources,
             quantities, types, and radiological characteristics of the scrap metal that may be
             impacted by national standards governing the free release of scrap metal from
             nuclear facilities.
                                              •%

       2.     Estimates of the potential normalized annual dose and normalized lifetime risk to
             the reasonably maximally exposed individual (RMEJ) associated with the free
             release of scrap metal from nuclear facilities.

             The term "normalized annual dose" to the RMEI refers to the high end annual
             effective dose equivalent (EDE) that may be received by an individual due to the
             release of scrap metal that may contain trace levels of radioactivity. The
             normalized dose is expressed in units of mrem/y EDE per picoCurie per gram
             (pCi/g) of a specific radionuclide in scrap metal.

             The normalized dose is a useful metric because for any free release criterion
             established in unite of mrem/y, the normalized dose can be used to derive the
                                          3-2

-------
              average1 radionuclide concentration level in scrap metal that corresponds to the
              criterion.  For example, Table 7-1 indicates that the normalized dose to the RMEI
              from Co-60 is 0.899 inrem/y per pCi/g.  Assuming a free release criterion of 1
              mrem/y, the average Co-60 contamination level in scrap metal that corresponds to
              1 mrem/y can be estimated by dividing the free release criterion (1 mrem/yr) by
              the normalized dose (0.899 mrem/yr) or 1.1 pCi/g. Alternatively, if the average
              concentration of a given radionuclide in scrap metal is known, the annual dose to
              the RMEI resulting from its  release can be estimated by multiplying the
              normalized dose by the average concentration of the radionuclide in the scrap
              metal.

              The normalized risk is similar to the normalized dose except that it is expressed in
              units of lifetime risk of cancer per year of exposure per pCi/g  of a given
              radionuclide in scrap metal.  As such, it can be used to derive  that concentration
              of a given radionuclide in scrap metal that corresponds to a given risk. It can also
              be used to derive the potential lifetime risk of cancer for the RMEI associated
              with the release of scrap metal containing a known average concentration of a
              given radionuclide.

              Imbedded in both these definitions is the concept of the RMEI. As used in this
              report, the RMEI refers to that individual, within the group of people that have the
              greatest potential for exposure to residual radioactivity contained hi scrap metal
              released from nuclear facilities, who would receive the high end exposure. This
              group of people, which can be referred to as the critical or limiting population
              group for a given radionuclide, have job responsibilities or living habits that result
              hi elevated potential for exposures as compared to other groups. Within the
              group, there is variability among the members with regard to their individual
              potential for exposure. The  RMEI is that individual within the group that has a
              relatively high potential (e.g., ,90th percentile) for exposure. As such, it is unlikely
              that many individuals within or outside the group could receive exposures
              significantly greater than those of the RMEI; most individuals that may be
              exposed are likely to receive exposures that are substantially lower than those
              received by the RMEI.
     1  When deriving the normalized annual dose to the RMEI, the volume of scrap metal over which the
radionuclide concentrations are averaged differs depending on the exposed individual. For example, when deriving
the normalized annual dose to a mill worker, the volume of scrap metal of concern is the entire scrap metal
throughput at the mill over a year. However, when deriving the normalized annual dose to a user of a product made
from scrap metal from a nuclear facility, the averaging volume is the volume of scrap metal required to make the
product.

                                           -3-3

-------
       3.     Estimates of the potential normalized, time-integrated collective dose and
              normalized, time-integrated potential collective public health impacts associated
              •with the release of scrap metal from nuclear facilities.

              The normalized, time-integrated collective dose is expressed in units of the
              collective dose (i.e., person rem) to which a population is committed per unit
              activity contained in released scrap metal.  The units can therefore be simply
              expressed in terms of person rem per Curie, for-example.  The normalized
              collective dose is a convenient metric because, once a determination is made of
              the total quantity and radionuclide inventory that is to be released, the collective
              dose can be determined by simply multiplying the curie inventory associated with
              the scrap metal by the normalized collective dose for each radionuclide.

              The time-integrated normalized collective risk is similar to the time-integrated
              normalized collective dose except that it is expressed in units of the potential
              numbers of adverse health effects per Curie of each radionuclide contained in the
              released scrap metal. As such, it can be used to derive the potential number of
              health effects that may be attributable to the release of a given quantity of scrap
              metal containing a known inventory of radionuclides.

            .  As is the case for the normalized dose for the RMEI, the time-integrated
              normalized collective dose for each radionuclide is derived to ensure that the
              potential collective health impacts are not underestimated; i.e., each represents an
              upper end estimates given their uncertainties.

       4.     Estimates of the minimum detectable concentrations of radionuclides contained
              within or on the surface of scrap metal

              The minimum detectable concentrations (MDC) of various radionuclides are
              needed to assess the feasibility and implementation costs of alternative free
              release criteria. For example, Table 8-6 reveals that the MDC for Co-60 using
              conventional radiation survey techniques is about 470 dpm/100 cm2 for large area
              sources. This value corresponds to about 0.5 mrem/y.

3.2    PRIMARY SCREENING CRITERIA


The following screening criteria were used to focus the scope of the EPA's investigations:


       1.      Sources and types of scrap metal and exposure scenarios, pathways, and
              biological endpoints that are potentially significant or limiting in terms of
              individual and collective doses and risks.
                                          3-4

-------
             For the purposes of its evaluation, EPA defined significance in terms of: the
             quantity of metal, the likelihood that the scrap metal is contaminated, the half-life
             of the contaminants, the concentration of radionuclides hi or on the scrap, the
             magnitude of the potential for exposure based on a person's occupation and living
             habits, and the likelihood of a given adverse health impact occurring.

       2.     The potential market value of the metal.

             The potential market value of a metal may influence the economic impact of a
             future EPA rule governing the release of scrap metal from nuclear facilities. The
             potential market value is determined by the product of the quantity of the scrap
             metal and the unit quantity price of the metal.
3.3    SOURCES OF SCRAP METAL CONSIDERED - ADMINISTRATIVE AND
       FUNCTIONAL CATEGORIES

The potential sources of scrap metal can be categorized by function (e.g., reactors, research
laboratories) and by the administrative authority responsible for their management and
disposition (e.g., DOE, NRC).  Information pertaining to scrap metal from nuclear facilities is
defined and accessible hi terms of these two broad categories. The process used to screen
potential sources of scrap metal was to: (1) review data available within each category of
administrative authority and (2) assess the degree to which the functional categories were
represented. This approach was found to be the most practical method for acquiring scrap metal
data because the needed information was more readily accessible by administrative authority.

3.3.1   Administrative Authorities

The principal administrative authorities responsible for the management of scrap metal from
nuclear facilities are:

       1.     The Department of Energy
       2.     The Nuclear Regulatory Commission
       3.     The Department of Defense
       4.     State or Superfund Authority

Table 3-1 presents an overview of the various administrative categories of sites containing or
contaminated with radioactive materials. A review of data available characterizing contaminated
structures within these administrative categories is provided in EPA 96 (page E4-7). The review

                                          3-5

-------
revealed that there are a total of about 30,000 structures La the major jurisdictional sectors, of
which about 8,000 are contaminated. These structures, along with scrap metal already in storage
at many DOE sites, represent the major potential sources of scrap metal at nuclear facilities in the
U.S.

The Department of Energy (DOE)

DOE is responsible for cleaning up more than 130 contaminated facilities in over 30 States and
territories (DOE 95, BEMR Executive Summary, page iii). These include approximately 45
national laboratories and nuclear weapons production and testing facilities where environmental
restoration and waste management activities are taking place. Many of these are large sites with
facilities that have been used for multiple activities related to nuclear weapons research,
production, and testing over the years and have many areas of contamination. Many of these
facilities also have extensive mixed waste contamination. Several DOE facilities have literally
hundreds of areas that are being investigated and cleaned up separately. For example, DOE's
Hanford facility, which encompasses 570 square miles, is divided into about 1,100 individual
"waste site units" based on their waste characteristics or other factors. EPA 96 estimates that
DOE sites contain a total of 3,179 containment structures and 1,179 buildings. These sites,
which have been grouped into 78 operable structures, are believed to represent the major sources
of scrap metal under the authority of the DOE.

DOE's Environmental Restoration and Waste Management (EM) program is in the process of
characterizing and decontaminating and decommissioning these facilities and restoring the
environmental conditions at these sites.  Information on the status of these programs is provided
in many DOE core documents (DOE95; 95a; 95b). In addition, DOE's "Inventory Report" has
estimated the current and projected inventory of potential scrap metal at many of its facilities
(MIN 96). Table 3-2 siimmari7.es these estimates.
                                          3-6

-------
Table 3-1. Inventory of Sites That are Known to be Contaminated with Radioactivity (from
            EPA94).
-. . -.>•*" ' , f f f f t
f '•' • ••* •''•^ -. %%^ v -V-X" ? ^ •" '"•••• "
Federal
DOE
Major DOE Facilities
National Laboratories
FUSRAP*
UMTRAP**
Other DOE Sites
DOD
Major DOD Facilities
Sites with Burial Areas
Sites with Accident Contamination
Sites withDU*** Contamination
Other DoD Sites
Other Federal Sites
NRC/Agreement State Licensees
Nuclear Power Plants
Test and Research Reactors
Other Fuel Cycle Facilities
Rare Earth Extraction Facilities
Byproduct Material Facilities
Non-Federal NPL Sites
Municipal Landfills
Radium Sites
Other Sites
Other State Sites
~:>-^MMi&.^l


121
72
27
10
31

I3
85
1
15
57
24

125
63
65
22
4401

3
7
11
- (No reliable data)
1.

2.

3.
4.
*
**
***
Fernald, Hanford (4 subsites, including 100,200,300, and 1100 Areas),'ENEL, Mound, Nevada Test Site,
Oak Ridge, Paducah, Pantex, Portsmouth, Rocky Flats, Savannah River, Weldon Spring.
Argonne, Brookhaven, Fermi, Lawrence Berkeley, Lawrence Livermore (main Area and 300 AREA.), Los
Alamos, Sandia.
Aberdeen Proving Ground
Watertown Arsenal (GSA), Fremont National Forest (USDA)
Formerly Utilized Sites Remedial Action Program
Uranium Mill Tailings Remedial Action Program
Depleted Uranium
                                              3-7

-------
Table 3-2. Estimates of Existing and Projected Potential Sources of DOE Scrap Metal Used for
           Screening Purposes (from MIN 96)
• > """'• s "/
DOE Site ,V ^V':

Femald
Hanford
INEL
K-25
ORO
Paducah
Portsmouth
LANL
Rocky Flats
SRS
Pinellas
Weldon Spring
Y-12
Total

••••:••.'>..•">,, > •• •-/%,:.//; '..- > '.',.•>•,;
Current Inventory
23,100
16,263
2317
45,000
1764
32,400
24,600
3407
70
16,461
54
30,623
11,338
207,370

I"-- l^efSe*a£M*Sa1?;^
Projected from D&D
149,000
129,000
37,000
_
727,000
_
_
_
29,000
4000
_
_
_
1,073,335

'',,'• • it? -. . •
' ' ;i Yf ' t' ;•;„* ' '/ ' '
Total
172,000
145,000
39,300
45,000
729,000
32,400
24,600
3407
29,070
20,461
54
30,623
11,338
1,200,000
Based on this understanding of the potential quantities of DOE scrap metal, the DOE sites and
facilities listed in Table 3-2, with the exception of Pinellas (due to the very small quantity), are
explicitly included hi the scope of the TSD2.


The Nuclear Regulatory Commission (NRQ


The NRC and its Agreement States have licensed about 22,000 facilities for the production and
handling of radioactive materials (EPA 93). About one third of these are NRC licensees, while
     2  Other sections of this report present more detailed characterizations of the quantities of scrap metal at these
sites. The results of these more detailed investigations provide volumes that differ from those in Table 3-2.
However, it was the quantities of potential scrap metal summarized in Table 3-2 that were used to identify those
sources of DOE scrap metal that would be included in this TSD. As indicated hi other sections of the TSD, estimates
of scrap metal quantities are continually being revised.
                                             3-8

-------
the remainder are licensed by Agreement States under Section 274 of the Atomic Energy Act.
Licensees include universities, medical institutions, radioactive source manufacturers, and
companies that use radioisotopes for industrial purposes. About 50 percent of NRC's 7,500
licensees use either sealed radioactive sources or only small amounts of short-lived radioactive
materials. Activities at these facilities are not likely to result in residual radioactive
contamination that will need to be cleaned up and disposed because: (1) the radionuclides remain
encased and cause little (if any) contamination and/or (2) because the radionuclides rapidly decay
to non-radioactive elements. A small number of licensees (e.g., radioactive source
manufacturers, radiopharmaceutical producers, and radioactive ore processors) conduct
operations that could result in substantial radioactive contamination in portions of the facility. In
addition, about 250 facilities associated with the production of nuclear power3 maintain large
inventories of radioactive materials; many of these facilities will need to be cleaned up before
their licenses can be terminated.

EPA 96 estimates that NRC licensees are responsible for a total of 4,625 buildings and
structures, including:
                                     F

       •       125 nuclear power plants
       «       63 test reactors
       •       930 sealed source manufacturers
              3,471 medical and R&D facilities
       •       22 rare earth production facilities
       •       14 fuel fabrication plants

Only the scrap metal from commercial nuclear power plants are explicitly included in EPA's
current assessment. The other potential sources are not addressed due to the relatively small
volume of scrap metal generated and/or the short half-lives of the radionuclides involved.

Department of Defense (DoD)

DoD's Installation Restoration Program (IRP) consists of over 17,500 potential hazardous waste
sites located at 1,877 installations (Baca 92).  DoD sites vary widely in function and size.  They
include hospitals, laboratories, proving  grounds, bombing and gunnery practice ranges, missile
     3 These include nuclear power plants, non-power (research and test) reactors, fuel fabrication plants,
uranium hexafluoride production plants, uranium mill facilities, and independent spent fuel storage installations.

                                           3-9

-------
launch sites, weapons manufacturing and storage facilities, and reactors. Only a few of these are
currently known to have radioactive contamination although these sites have not been fully
characterized. Consequently, it is not possible to reliably estimate the number of sites with
radioactive contamination.

DoD sites may contain small enclosed radiation sources, such as radium and tritium instruments.
They may also contain larger sources, such as research reactors, and dispersed sources, such as
laboratory waste storage areas and test ranges contaminated with plutomum and fission products.

EPA 96 estimates that DoD is responsible for 161 contaminated structures, including 43
containment structures and 118 buildings. Due to the relatively limited potential for scrap metal
and the unavailability of data characterizing the scrap metal, DoD sources of potential scrap
metal are not explicitly addressed hi the TSD.

State or Superfund Authority

This administrative category includes sites that are not licensed by NRC or Agreement States but
are under State or Superfund authority. (Sites that are under Superfund authority are those that
are on the National Priority List and are being cleaned up by the Federal government.)  This
category includes  about 1,000 particle accelerator sites that generally contain only small amounts
of short-lived residual radioactivity after shutdown.  Other sites included in this category contain
long-lived naturally-occurring radionuclides varying from small packaged radiation sources to
large areas of mostly low-level dispersed contamination, including mining wastes and materials,
tailings from ore processing, and residues from university or commercial research activities.

The principal sources of scrap metal in this administrative category include metal contaminated
with naturally-occurring radioactive material (NORM), primarily from the oil and gas industry.
This source of scrap metal has been excluded from EPA's current assessment because the
Agency has chosen to limit its evaluation to scrap metal generated by nuclear facilities, thereby
excluding sources contaminated with NORM.

3.3.2  Functional Categories

It is also possible to categorize scrap metal from nuclear facilities according to a variety of
functional categories. Table 3-3 presents the functional categories used to classify nuclear

                                          3-10

-------
facilities in EPA 94. This categorization scheme is useful because it helps to assess the degree to
which the compilation of data characterizing scrap metal represents a full range of functional
categories.
Table 3-3.  Functional Categories for Nuclear Facilities 'and Sites Containing or Contaminated
           with Radioactive Materials
• '•••'' ' • '• ' •&-VV, "•"'•• f f&* %ffff ftfffffff^Jf fff f' .
^•v. '. ,,. ' ' '•'*/ s •"••'•• ^ • y -' '. : ' •$>•>
*'•*'•* ff """"" f V$ ff^. "" *?'• ' ~> *:^ v \ v .',- : w .""
-". ^y*..+ ^sav^V^'^5***^^"'
" f-:- :. !•- Bia<^«m£sl§eg(^?> ..;, -' r^
Mines, Mills, and Rare Earths
Conversion and Enrichment
Fuel Fabrication and Weapons
Assembly
Reprocessing and Extraction
Reactors
Research, Biomedical, and
Analytical Labs
Industrial and Commercial (Non-
sealed sources)
Sealed Source Users
Nuclear Medicine Departments
Accelerators
Fusion Facilities
Nuclear Test Sites
Weapons Accidents and Safety
Tests
DU
Other DoD Facilities
Waste Disposal
Naturally Occurring Radioactive
Material (NORM) Wastes Mixed
with AEA Materials
Entire Multipurpose Facility
' ; '•',' \.f\ , :; 'f ,' •• $ ' >f>- V V'
• s?&W-£ • -' x- • '•'' " '^ v-
DOE and NRC Licenses
DOE
DOE, DoD, and NRC
Licensees
DOE and NRC Licensees
DOE and NRC Licensees
NRC Licensees
NRC Licensees
NRC Licensees
NRC Licensees
DOE
DOE
DOE
DOE and DoD
DoD
DoD
DOE, DoD, NRC
Licensees, and EPA Non-
Federal NPL Sites
DOE and EPA Non-Federal
NPL Sites
DOE and DoD
^1- l^^^ife^ae|^ V,;
33ara$|i£e!EB3 A&&4$^%t'fets&S£t6ft-d
No**
Yes*
Yes*
Yes*
Yes
No
No
No
No
No
No
No
No
No
No
No
No**
Yes*
*  To the extent captured by Table 3-2
** Not within the scope of EPA's current investigations.
                                           3-11

-------
3.4    TYPES OF SCRAP METAL CONSIDERED

Not all scrap metal within the administrative and functional categories evaluated by EPA was
considered in this assessment As mentioned in Chapter 1, the Agency's evaluation focussed on
scrap metal that is moderately contaminated with radioactivity as a result of deposition or neutron
activation. Some metal is so highly contaminated that recycling is not feasible. For example, the
canyons at fuel reprocessing facilities and reactor internals are so contaminated that they are not
considered potential sources of recyclable scrap metal and were therefore not included in this
assessment On the other hand, scrap metal that is not radiologically contaminated, such as scrap
from locations outside the radiation control areas at DOE sites and commercial nuclear power
facilities, is also not included.

The types of metals that EPA explicitly considered have either economic significance and/or
potential public health significance. Economic significance is determined simply by assessing
the potential quantity and unit quantity value of the metal.  The potential public health
significance is a more complex problem.  Public health significance is determined by evaluating
the potential for the metal to result in significant radiation exposures to either a selected group of
individuals or the general population. This means that, though the quantity or value of a
potential type of scrap metal may be small, the radionuclide content and the way in which the
metal is processed, handled, and ultimately used commercially after.being released from a
nuclear facility could result hi the limiting normalized doses for specific radionuclides. As such,
such metals require explicit consideration when developing the normalized dose and risk tables.

3.4.1   Screening Based on Economic Value
                                              =s
Table 3-4, taken from data provided in MIN 96, presents an overview of the potential economic
value of metals that are known to be potential sources of scrap metal from nuclear facilities.
Based on potential value and consideration of the availability of data characterizing each metal
type, the metals evaluated include:

       •      Carbon steel
       •      Stainless steel
       •      Copper
       •      Aluminum
       •     Nickel
                                          3-12

-------
Data characterizing the potential quantities and radiological characteristics of the selected metals
were gathered (where available) for the facilities and sites identified in Section 3.3. However, as
discussed in subsequent sections, normalized doses and risks were developed only for carbon
steel.

Table 3-4. Potential Economic Value of Types of Scrap Metal (MIN 96)
\^S?:"!j

Carbon Steel
Stainless Steel
Gal. Iron
Copper
Inconel
Lead
Bronze
Aluminum
Brass
Nickel
Silver
Depleted Uranium
Others
Total
5^;£p4.: ^::'&tf^i'J:]'?.-
DOE
0.7E6
0.4E5
.
6.1E4
-
2.0E3
_
2.9E4
_
5.9E4
_
1.9E5 Ft3
3.1E5
1.2E6
NRC
3.5E6
1.9E5
1.4E5
7.7E4
1.3E4
5.2E3
2.8E3
0.2E4
1.1E3
1E2
1E2
•%
_
3.9E6
Total
4.2E6
2.3E5
1.4E5
1.38E5
1.3E4
7.2E3
2.8E3
3.1E4
1.1E3
5.91E4
1E2


5.2E6


80
600
80
2,400
about
1000
250
_
450
_
4950
256,000
_
„

tsi^;t'

336E6
138E6
11.2E6
331E6
130E6
1.8E6
„
14E6
_
293E6 ,
25.6E6
.
_
about 1.3 billion
3.4.2   Screening Based on Public Health Considerations

Notwithstanding the value of a given type of metal, a metal may also be of interest if it has the
potential to have normalized doses and risks which are substantively greater than those for
carbon steel.  The mathematical models and assumptions used for deriving normalized doses and
risks, as described in Chapter 6, are based on a reference electric arc furnace used to recycle
                                           3-13

-------
carbon steel. The normalized doses for the Reasonably Maximally Exposed Individual (RMEI)
presented in Table 7-1 are based on assumptions regarding the throughput and dilution of scrap
metal representative of recycling carbon steel. In addition, the assumptions regarding partition
factors, geometries, exposure scenarios and end uses of the slag, baghouse dust and commercial
products resulting from recycling scrap metal are based on carbon steel.  It is possible that
recycling aluminum, copper, or other metals from nuclear facilities could result in normalized
doses and risks that are substantially higher than those derived for carbon steel. In other words, it
is possible that the free release of metals other than carbon steel, based on normalized RME
doses developed for carbon steel, could result in doses to the RMEI which might exceed any
future radiation protection standards. Similarly, it is possible that the normalized collective
doses for other metals are greater than those derived for carbon steel. These issues are discussed
further in the following subsections.

Normalized RME Doses

In order to address the applicability of normalized RMEI doses and risks for carbon steel to other
metals, a review was performed of the processes used to recycle copper and aluminum.
Appendices B and C present the results of this review. The review attempted to ascertain
whether there are aspects to the recycling of aluminum and copper that are significantly different
than those associated with the recycling of carbon steel and which could result in substantively
higher normalized RMEI doses for some radionuclides. It was determined that the processes are
very different and the potential exists for markedly different dilution and partition factors for
some metals which could influence the normalized RMEI doses. Hence, specific investigations
are planned to further assess this issue.

Normalized Collective Doses

The collective doses are derived as the product of the normalized collective  dose (person rem/Ci)
for a given radionuclide and the total quantity of that radionuclide that is contained in the scrap
metal. Table 3-4 reveals that the quantity of metal other than carbon steel that may be released
from nuclear facilities is a very small fraction of the quantity of carbon steel likely to be released.
This is especially true for non-ferrous metals. Consequently, the potential for metals other than
carbon steel to contribute significantly to the collective dose, relative to that of carbon steel, is
extremely small.  It is, therefore, not necessary to derive normalized collective doses for metals
other than carbon steel.

                                          3-14

-------
 3.5    RADIONUCLIDES SELECTED FOR CONSIDERATION

 The criteria used to select the radionuclides for which normalized doses and risks were derived
 are as follows:

       1.     Half-lives greater than 6 months. This cutoff was selected because the time delay
              between contamination and release from the facility would preclude significant
              exposures associated with the shorter lived radionuclides. Most radionuclides
              with half-lives less than 6 months, and which appear to have the potential to be
              present in scrap metal, have very short half-lives.  Hence, 6 months was selected x
              as a convenient break point.,

       2.     Presence in scrap metal.  The relevant radionuclides were identified by a review of
              source documents describing the radionuclides associated with the uranium fuel
             , cycle and found in scrap metal.

       A detailed discussion of this selection process is presented in Appendix D.

 3.6    SCENARIOS, PATHWAYS, AND BIOLOGICAL ENDPOINTS CONSIDERED

 There are about 160 mills hi the United States that may be willing to accept scrap metal from
 nuclear facilities for recycling if radiation protection standards were established. Theoretically,
 hi any given year, some of these mills might handle large quantities of scrap metal from nuclear
 facilities, while others might handle little to none. The number of people that may be exposed to
• radioactivity as a result of these activities could  be numerous. They include the handlers and
 transporters of the scrap metal, the various workers at the mills, other workers downstream from
 the mill operations who handle the steel products, slag, and other materials contaminated as a
 result of recycling scrap metal, members of the public who live near the mills and are exposed to
 the effluents from the mills, consumers who use the products made from scrap metal, and people
 exposed as a result of the final disposal of the products and wastes associated with the recycling
 of scrap metal.

 The workers at the mills are likely to receive the highest doses. Some workers will receive the
 highest doses because of the nature of then: jobs and, perhaps, some unique aspects of the
 operations at specific mills. The individuals that are anticipated to receive the highest individual
 doses as a result of one year of recycling operations are referred to as the RMEI. Hence, out of a
                                          3-15

-------
population that may be exposed as a result of one year of scrap metal recycling operations, the
RMEI is anticipated to receive the highest dose.


A screening process was used to identify those categories of people that have the greatest
potential for individual exposure, the pathways responsible for the exposure, the key assumptions
used to model the exposures, and the biological endpoints of concern. The selection of the
scenarios, exposure pathways, and biological endpoints addressed in this report were selected
based on a review of previous reports addressing the recycle of scrap metal (see Appendix D),
several visits to scrap yards and steel mills, and the insight gained in the preparation of the
precursor documents to this report, primarily SCA 95.


Screening analyses performed in support of SCA 95 derived normalized doses for.about 70
individuals that were grouped according to  14 categories:


       1.     Salvage yard operations
       2.     Transportation to mill
       3.     Scrap yard operations at mill
       4.     Furnace operations
       5.     Foundry operations
       6.     Bag house operations
       7.     Slag operations at mill
       8.     Transporting slag
       9      Transporting baghouse dust
       10.    Slag use in road construction
       11.    Manufacturing products made from recycled scrap metal
       12.    Distribution and use of products made from recycled scrap metal
       13.    Disposal of slag, baghouse, dust and product
       14.    Offsite population exposed to airborne emission

The pathways of exposure included direct radiation, inhalation, ingestion of soot, ingestion of
farm products contaminated as a result of airborne emission from the mill, and contamination of
ground water due to leachate from slag piles.


As a result of these screening analyses, the  17 persons and scenarios described in Section 5
(Table 5-1) were selected for explicit consideration in this report. In addition, offsite doses from
the food ingestion pathways for C-14 and 1-129 were added to the list since these radionuclides
have the potential to be released in the gaseous effluent from a mill and reconcentrate in
vegetable, beef, and milk produce.

                                          3-16

-------
The values selected for the modeling parameters used in the analyses (see Chapters 5 and 6) were
selected to represent the upper end of the range of possible values.4 Among the number of mills
and numerous scrap dealers, there is large variability in working practices, exposure durations,
airborne dust loadings, etc.  As such, by selecting upper end assumptions, there is a level of
assurance that the normalized doses thereby derived reflect the worker and operational practices
that tend to result in the higher normalized doses. However, in selecting the modeling
assumptions, the extreme upper ends of the possible range of the distributions were not selected
in order to avoid over-conservatism. Hence, there may be some facilities, under some
circumstances, and some time periods, that could have normalized doses somewhat higher than
the derived values.

The biological endpoints of potential concern in the recycling of scrap metal from nuclear
facilities could include carcinogenic., genetic, and teratogenic effects.  The methodology used by
EPA in its evaluation was designed 1o specifically quantify the risk of total cancers (not including
non-fatal skin cancer) and fatal cancers. This methodology is consistent with the approach
typically taken by EPA in developing its radiation protection standards. The Agency does not
quantify the potential for non-carcinogenic health effects because they are far less likely to occur
than carcinogenic effects at the dose levels potentially associated with recycling scrap metal from
nuclear facilities (EPA 89). This approach is supported by current international radiation
protection guidance (UN 93).5

Though the models were developed to explicitly quantify the potential number of carcinogenic
effects, the models also present the time integrated collective dose (person rem). This parameter
can be used to estimate potential stochastic effects, other than carcinogenesis, that may be '
averted by future radiation protection standards governing the recycling of scrap metal from
nuclear facilities.

The objective of the screening process was to limit the individuals, scenarios, pathways, and
biological endpoints to a manageable number without excluding any individuals, scenarios,
       ''it is not possible to assign quantitative confidence limits at this time based on the limited data that is currently
available.

       5UN 93 cites a risk coefficient of 5x10"4 per rem for lifetime fatal cancer risk in a nominal population of all ages.
The risk coefficient cited for genetic effects is 1.2x 10^ for a reproductive population for all generations after exposure.
For clinically important disorders for the first generation of offspring of exposed parents, the genetic risk coefficient is
cited as 0.2x10^ to 0.4X10*4 per rem for the reproductive part of the population.

                                           3-17,

-------
pathways, and biological endpoints that could result hi substantially higher normalized doses
than those presented in Table 7-1.  In addition, the modeling assumptions were selected to
provide a level of assurance that the actual normalized doses at real facilities are not significantly
underestimated.

Multiple Pathways
                         %
Consideration was given to the assumption that some individuals could be exposed by multiple
pathways/scenarios. For example, the lathe operator may be exposed to the lathe at work and to
kitchen appliances at home or he may live downwind from a mill recycling scrap metal from
nuclear facilities. Such combinations of limiting pathways and scenarios are unrealistic. In the
first example, not only would the lathe be made entirely from scrap from nuclear facilities, but so
would the lathe operator's kitchen appliances.  The probabilities of this occurring are extremely
small. It must be recognized that, in reality, during the approximate 10 to 50 year tune period
that the approximate 2 to 3 million metric tonnes of scrap metal from nuclear facilities may be
recycled, the metal will be  diluted in a national scrap metal flow of about 68 million tons (62
million metric tonnes) per year. Hence, the scrap from nuclear facilities would experience an
approximate 200-fold dilution hi the finished product. The assumption that the lathe operator is
exposed to a lathe made entirely from undiluted scrap metal from the  nuclear industry is itself
extreme, and the probability that he is also exposed to other undiluted products is even less
likely.

The second example given is the lathe operator who, by chance, happens to live downwind from
the mill that receives the most scrap metal from nuclear facilities in the country.  The limiting
mill was selected as a mill that receives 13 percent of its scrap hi one year from nuclear facilities.
Appendix G indicates that there may be one mill where this can occur, but it is much more likely
that most mills, if not all, will receive a much smaller percent. Hence, the combination of
assumptions that the lathe operator not only has a lathe that is made entirely from scrap from
nuclear facilities, together with the assumption that he happens to live downwind from the
limiting mill, is unrealistic.

Similar arguments can be made regarding other limiting scenarios and pathways. Because of the
unlikelihood that they will  occur, combined pathways/scenarios are not assumed in this analysis.
                                          3-18

-------
Other Pathways

Many scoping analyses were performed to ensure that important scenarios and pathways were not
overlooked in EPA's evaluation. For example, consideration was given to including exposure
from food grown in soil that uses slag as a soil conditioner (liming agent) in determining the
RMEI normalized dose. Scoping calculations revealed that, though this is a realistic pathway, it
could not be limiting for the RMEI because of the large dilution that would be experienced by the
slag when used as a soil conditioner.  For example, assuming 100 pounds of slag are applied per
1000 ft2, down to a plow depth of 15 cm, the combined vegetable/meat/milk pathway normalized
dose for Sr-90 in slag would be less than 0.1 mrem/yr per pCi/g of Sr-90 hi scrap metal. This is
as compared to the limiting normalized dose for Sr-90 of 3.03 reported in Table 7-1. However,
this scenario is significant with regard to the collective impacts and is included in that analysis.

3.7    SUMMARY OF THE SCREENING PROCESS AND ASSOCIATED LIMITATIONS

This section summarizes the results of the screening process and the resulting scope of the
analyses that comprise EPA's evaluation. This section also includes a brief discussion of the
limitations associated with the selected scope. "Limitations" refers to potential misleading or
erroneous conclusions that could result from limiting the scope of the analyses.

3.7.1   Sources of Scrap Metal

Scope

Out of the approximate 17 functional categories (see Table 3-3) and the four major
administrative categories that represent the sources of scrap metal from nuclear facilities, the
EPA's analysis explicitly addresses four functional categories, two administrative categories, and
14 sites that contain or are contaminated with radioactivity. The four functional categories
include DOE conversion and enrichment facilities, fuel fabrication and weapons assembly plants,
reprocessing and extraction facilities, and reactors.  The two administrative categories are DOE
and the NRC. The 14 sites include the 12 DOE sites listed in Table 3-3 (Pinellas is not
included), one reference BWR (Washington Public Power Supply System Nuclear Project No. 2)
and one reference PWR (Trojan Nuclear Plant).
                                         3-19

-------
Limitations

The Agency believes that the important sources in terms of quantity of scrap metal available for
recycle have been captured in the analysis. Hence, a cost/benefit analysis based on the sources
addressed in this report should provide a reasonable national perspective on the potential impacts
of recycling of scrap metal from nuclear facilities. "However, the analysis will provide only
limited insight into the variability of the cost/benefits associated with individual sites and
categories of facilities. As such, unique site-specific issues may need to be addressed on a case-
by-case basis.

       Limitation 1 - The analyses will provide only limited insight into the variability of
       the cost/benefits associated -with individual sites and categories of facilities.

3.7.2  Types of Scrap Metal from Nuclear Facilities

Scope

Of the approximate 12 types of metals and metal alloys that may comprise scrap metal from
nuclear facilities, the Agency has characterized the volume and radionuclide composition of five:
carbon steel, stainless steel, copper, aluminum, and nickel. Of these, normalized doses are only
derived for carbon steel, which may be considered generally representative of ferrous metals.

Limitations

The five types of metal selected for characterization represent the majority of the potential
quantity of scrap metal available from nuclear facilities for recycle.  Hence, there is very little
likelihood that an important type of scrap metal has been overlooked from the perspective of
quantity or economic value. Therefore, the analysis should provide a reliable national
perspective on the costs and benefits of recycling different types of metals from nuclear facilities.
However, the analyses provide limited information pertinent to the assessment of the impacts of
recycling metals not explicitly addressed in this report. As a result, conclusions regarding the
cost/benefit of recycling metals presented in this report cannot necessarily be extended to metals
outside the scope of EPA's analyses.  Hence, the applicability of the results of the Agency's
evaluation to other metals may need to be addressed on a case-by-case basis.
                                           3-20

-------
       Limitation 2 - Concisions regarding the cost/benefit of recycling the metals
       presented in this report cannot necessarily be extended to metals outside the
       scope of EPA's analyses.

The normalized doses presented in the TSD apply specifically to carbon steel. They cannot
necessarily be extended to other metals, nor is there assurance that the doses from other metals
contaminated with specific radiomiclldes may not be more limiting than those from carbon steel.

       Limitation 3 - The normalized doses and risks presented in the TSD for carbon
       steel cannot necessarily be extended to other metals, nor is there assurance that
       normalized doses from other metals contaminated-with specific radionuclides may
       not be more limiting than those for carbon steel.

3.7.3   Radionuclides

Scope

127 radioisotopes could, in theory, be present in scrap metal from nuclear facilities.  Of these, the
TSD explicitly addresses 40.

Limitations                                           h

A formal process was used to screen to the possible radionuclides of concern and to select the 40
that were ultimately evaluated. Hence, the possibility mat one or more significant radionuclides
has been overlooked is small. However, the 6 month half-life cutoff was based on an
engineering judgment that there will be a substantial delay (on the order of years) between the
contamination of scrap metal at nuclear facilities and its release as scrap metal. If the time
between contamination and free release is substantially less than a few years (e.g., months), the
list of radionuclides addressed in the analysis may need to be expanded to assess these special
cases.

       Limitation 4 - The list of radionuclides addressed in the analysis may need to be
       expanded for sites where the time between contamination  and release of scrap
       metal is on the order of months instead of years.
                                          3-21

-------
3.7.4  Scenarios. Pathways. Modeling Assumptions, and Biological Endpoints

Scope

Out of the virtually unlimited number of possible exposure scenarios, 19 scenarios were selected
for analysis. The pathways selected for analysis include external exposure, inhalation of dust,
ingestion of soot, and ingestion of produce. Upper end values for the modeling assumptions
were selected. Of the range of biological endpbints that could be of concern (I.e., individual and
collective dose, individual risk of fatal and non-fatal cancer, total numbers of fatal and non-fatal
cancers in a population, hereditary effects in individuals and populations, and teratogenic effects
in individuals and populations), hereditary and teratogenic effects were not explicitly addressed.

Limitations

There is always a possibility mat some individuals, at some facilities, could have working
practices that could result in higher normalized doses than those identified for the RMEI. There
could also be some pathways, other than those addressed in this report, that could also result in
higher normalized doses. In addition, though a deliberate effort was made to select high end
modeling assumptions, there could be some facilities where the exposure durations are longer
and the dust loadings factors are higher than those considered in the analysis. These limitations
are not unique to this analysis; therefore, they are not identified as limitations specific to EPA's
current evaluation of the impacts of recycling scrap metal from nuclear facilities.
                                          3-22

-------
                                  REFERENCES
Baca 92      Baca, T.E., "DoD Environmental Requirements and Priorities," Federal Facilities
             Environmental Journal, Autumn 1992.

DOE 95      U.S. Department of Energy, "Estimating the Cold War Mortgage, The 1995
             Baseline Environmental Management Report," DOE/EM-0230, March 1995.

DOE 95a     U.S. Department of Energy, "Draft Waste Management Programmatic
             Environmental Impact Statement for Managing Treatment, Storage, and Disposal
             of Radioactive and Hazardous Waste," DOE/EIS-0200-D, August 1995.

DOE 95b     U.S. Department of Energy, "Integrated Data Base -1994: U.S. Spent Fuel and
             Radioactive Waste Inventories, Projections, and Characteristics, DQE/RW-0006,
             Rev. 10.

EPA 89      Environmental Protection Agency, "Risk Assessments Methodology -
             Environmental Impact Statement - NESHAPS for Radionuclides - Background
             Information Document - Volume 1," EPA/520/1-89-005-1, September 1989.

EPA 93      U.S. Environmental Protection Agency, "Issues Paper on Radiation Site Cleanup
             Regulations," EPA 402-R-93-084, September 1993.

EPA 94      U.S. Environmental Protection Agency. Radiation Site Cleanup Regulations:
             Technical Support Document for the Development of Radionuclide Cleanup
             Levels for Soil — Draft for Review; Including Appendices A-K, Appendices L-O,
             EPA Office of Radiation and Indoor Air, 1994.

EPA 96      U.S. Environmental Protection Agency", Radiation Site Cleanup Regulations:
             Technical Support Document for the Development of Radionuclide Cleanup
             Levels for Soil — Addendum, EPA Office of Radiation and Indoor Air, EPA
             402-R-96-01 ID, July 1996.

IEC 97       Industrial Economics, Inc., "Radiation Protection Standards for Scrap Metal:
             Preliminary Cost-Benefit Analysis," prepared for the EPA Office of Radiation and
             Indoor Air, under Contract No. 68-DO-0102, Work Assignment Manager Reid
             Harvey, 1997.

MIN 96      "Taking Stock:  A Look at the Opportunities and Challenges Posed by Inventories
             from the Cold War Era," U.S. Department of Energy, Office of Environmental
             Management, DOE/EM-0275, January 1996.
                                       3-23

-------
SCA 95      Sanford Cohen & Associates, "Analysis of the Potential Recycling of Department
             of Energy Radioactive Scrap Metal," Prepared for the EPA Office of Radiation
             and Indoor Air under Contract No. 68D20155, Work Assignment 3-19, EPA
             Work Assignment Manager John MacKinney, August 14,1995.

UN 93       United Nations Scientific Committee on the Effects of Atomic Radiation,
             "Sources and Effects of Ionizing Radiation," United Nations, New York 1993.
                                       3-24

-------
                                    CHAPTER 4

               QUANTITIES AND CHARACTERISTICS OF POTENTIAL
                SOURCES OF SCRAP METALFROM DOE FACILITIES
                  AND COMMERCIAL NUCLEAR POWER PLANTS

This chapter provides quantitative data on the amount of scrap metal potentially available for
recycle from Department of Energy (DOE) sites and the commercial nuclear power industry.
Scrap metal quantities for DOE sources and the means by which the data were developed are
discussed hi Section 4.1 of this chapter. Because data available for the nuclear power industry
are considerably more detailed, only a summary description of the scrap metal database for
power reactors is presented in Section 4.2 of this chapter. A comprehensive discussion of scrap
metal sources that will be generated from the decommissioning of commercial reactors is
provided hi Appendix A of the TSD.

4.1   EXISTING AND FUTURE SCRAP METAL QUANTITIES AVAILABLE FROM DOE

4.1.1  Background Information

The DOE designs, tests, manufactures, and maintains nuclear weapons that are central to the
national security of the United States. This effort started with the Manhattan Project and the
development of the first nuclear weapons that were employed hi World War II.

Shortly after World War II, deteriorating relations between the United States and the Soviet
Union led to a massive nuclear arms race. In the United States, the nuclear arms race resulted hi
the development of a vast research, production, and testing network of Federal facilities that
came to be known as the "nuclear weapons complex." During half a century of operations, the
complex manufactured tens of thousands of nuclear warheads and test-detonated more than one
thousand.

During its peak, this complex consisted of 16 major facilities each with its own specific mission
(Figure 4-1). Weapons production stopped hi the late 1980's initially to correct environmental
and safety problems. With the end of the Cold War, most of the nuclear weapons activity has
been suspended indefinitely.
                                        4-1

-------
                              The  U. S.   Nuclear Weapons Complex
                              Nuclear weapons  production occured from World War II until the late 1980s
Amchilka
 Island
Nuclear Weapons Production
 Uranium
Mining and  Uranium
 Milling   Refining
                 Uranium
                Enrichment
Uranium   Target
foundry Fobrttalton  "«««<"»

      __/_
            enriehdl cranium,
                 uroniu
                  •
                          coav^fed into is formed mto
                            metal   W w«l krae?
                                 efemtfllt tor
                                  fcodofi
smum     Nuembiy ond
I        K- rsu             ^Sk Df«niantlemenl

tl	fV	^^>-*


      voamstod Inggen
                                                                                   Wofhtaa Trigg*? i, ne«W
                                                                                  generator*, six! orftef etettdc
 former industrial sifei contaminated with
: radioscllvfty, tome but not at! of which
 contributed to nuclear weapons production.


 Number ir.ditat«s how many tiles
 were or ore located in the State,
                                                                                 For n complete filling of rHe sites flvol tlw Dcpartmcni of Energy
                                                                                 Environmcnto! Manogemont program t\ rcijwntibte for see page 7^
                                                                Figure 4-1
                                                           (Source. DOE 95a)

-------
With the end of the Cold War and a diminished threat, the DOE no longer needs many of the
facilities and their components that formerly constituted the DOE nuclear weapons complex.
DOE is now in the process of deciding what should be done with facility structures and materials
that in many instances are radioactive or are radioactively contaminated. Among the materials
that pose significant disposition problems are large quantities of metals that have become
radioactively contaminated in various phases of extracting, testing, and producing nuclear
weapon materials.

Radionuclides Associated with Nuclear Weapons

The principal components of nuclear weapons are highly enriched uranium and plutonium. Early
nuclear weapons were designed to use either highly enriched uranium or plutonium that, when
forced into a "critical mass," would sustain a nuclear chain reaction and result in a nuclear
explosion. As designs for nuclear weapons improved, a new generation of "thermonuclear
weapons" evolved that require both plutonium and uranium highly enriched with the isotope U-
235. Thermonuclear weapons also require a third ingredient:  tritium, a radioactive gas of
hydrogen that boosts the explosive power of the nuclear weapon commonly referred to as the
Hydrogen Bomb.  The process by which these three components are produced is the source of
radioactive contamination of scrap metals at DOE facilities and the subject of this chapter.

Enriched Uranium. In nature, more than 99 percent of uranium atoms have an atomic weight of
238 with less than one 1 percent having an atomic weight of 235.  However, only uranium-235 is
capable of undergoing nuclear fission that is useful in the chain reaction of nuclear weapons. To
make highly enriched uranium-235, DOE facilities at the Oak Ridge Reservation in Tennessee
used two elaborate processes to isolate U-235 from U-238: (1) electromagnetic separation ha the
"Calutron" (California University cyclotron) and (2) gaseous diffusion.

Separated and U-235  enriched uranium hexafluoride gas must be converted into a metal matrix
before it is used in nuclear weapons production.  At the Fernald uranium foundry in Ohio, the
uranium gas was chemically converted into uranium metal. Enriched uranium metal was: (1)
used as fissionable material in nuclear weapons and (2),it was fabricated into nuclear fuel that
operated DOE "production reactors."
                                          4-3

-------
Between 1944 and 1988, DOE operated 14 plutonium-production reactors at the Hanford and the
Savannah River Sites producing about 100 tons of plutonium. Plutonium-239 is produced by
irradiating "depleted" uranium metal targets.

Both irradiated targets and the spent fuel of production reactors are the primary sources for the
production of weapon-grade plutonium.  Unfortunately, both sources also contain hundreds of
different radioactive isotopes that must be chemically separated. Scientists developed elaborate
physical structures and chemical processes to accomplish this separation in a manner that
considered worker and public safety.  A total of eight chemical separation plants, called
"canyons," were operated for the DOE that employed the PUREX process for the separation and
recovery of plutonium and uranium. In total, the eight chemical separation plants (i.e., canyons)
generated more than 100 million gallons of radioactive wastes that are currently cpntained and
stored at DOE facilities.

Sources of Data Used to Quantify and Characterize DOE Scrap

A thorough search for available reports and study data that might contain useful information
regarding  scrap metal inventories and a characterization of those inventories identified only a
very limited quantity. This was not unexpected when viewed in context of the highly
secretive/classified nature of past nuclear weapons activities, the relatively short time since the
end of the Cold War, and the yet-undecided future for many DOE facilities.

For these reasons, the DOE has only in recent years begun to evaluate existing and future
material inventories and their management. Some ofpOE's earliest attempts to assess material
inventories were based on the most cursory of data; data that were further compromised by an
uncertain and continuously revised projection of future needs. Earlier reports are, therefore, of
limited value and data reported therein have been revised and updated to reflect the most current
information, facility status, and future needs.

Currently, the most informative reports pertaining to  existing and future scrap metal inventories
include the following:
                                           4-4

-------
  (1)   DOE/EM-0275,1996, A Report of the Materials in Inventory Initiative. Taking Stock:  A
       Look at the Opportunities and Challenges Posed by Inventories from the Cold War Era,
       U.S. Department of Energy, Office of Environmental Management. (Note:  This report is
       commonly referred to as the "1996 MtN Report.")

  (2)   DOE/HWP-167, 1995, U.S. Department of Energy Scrap Metal Inventory Report for the
       Office of Technology Development, Office of Environmental Management, prepared by
       Hazardous Waste Remedial Actions Program for the U.S. Department of Energy.  (Note:
       This report is commonly referred to as the "HAZWRAP Report.")

  (3)   EPA-1995 Contract Report, Scrap Metal Inventories at U.S. Nuclear Facilities
       Potentially Suitable for Recycling, prepared by S. Cohen & Associates, Inc. for the U.S.
       Environmental Protection Agency, Office of Radiation and Indoor Air.

  (4)   ES/ER/TM-171,1995, Gaseous Diffusion Facilities Decontamination and.
       Decommissioning Estimate Report, prepared by Lockheed Martin Energy Systems, Inc.
       for the U.S. Department of Energy.

Collectively, these four documents identified 13 DOE facilities as principal sources of scrap
metal.  A brief description of each of the thirteen sites is presented below.

  •     Femald. Located on 1,050 acres hi the southwest corner of Ohio, Fernald Environmental
       Management Project (formerly known as the Feed Materials Production Center) was
       constructed in the early 1950s to convert uranium ore to uranium metal targets. Uranium
       targets were subsequently shipped to DOE production reactors, which irradiated targets
       for the production of plutonium used hi nuclear weapons. Over a 36-year period, this
       facility produced over 225 million kilograms of purified uranium. Production of uranium
       targets ceased in 1989. Principal radionuclide contaminants include uranium and its
       radioactive daughter products and technetium-99.

  •     Hanford.  The Hanford reservation encompasses about 560 square miles within the
       Columbia River Basin hi southeastern Washington and borders the Tri-Cities area of
       Richland, Pasco, and Kennewick to the south. Beginning hi the early 1940s, nuclear
       materials were produced at Hanford.  Activities once included plutonium production and
       separation, advanced reactor design and testing, basic scientific research, and renewable
       energy technologies development.
                                          4-5

-------
 Idaho National Engineering Laboratory. The Idaho National Engineering Laboratory
 (INEL) encompasses an area of approximately 890 square miles in southeastern Idaho on
 the edge of the Eastern Snake River Plain.  INEL is a multipurpose laboratory supporting
 the engineering and operations efforts of DOE and other federal agencies in the areas of
 nuclear safety, reactor development, reactor operations and training, waste management
 and technology development, and energy technology/conversion programs. Over 50
 nuclear reactors, most of them small test reactors, have existed at INEL. Some of these
 reactors and their associated support buildings have been decommissioned and
 demolished. Others are planned for decommissioning.

 Los Alamos National Laboratory. Los Alamos National Laboratory (LANL) occupies
 about 43 square miles approximately 25 miles northwest of Santa Fe, New.Mexico.
 LANL was established in 1943 with the specific responsibility of developing the world's
 first nuclear weapon. The Laboratory's original mission rapidly broadened to include
 research programs in nuclear physics, hydrodynamics, conventional explosives,
 chemistry, metallurgy, radiochemistry, and relevant life sciences.  In addition to research,
 a second important mission of the Laboratory between 1945 and 1978 was to process
 plutonium metal and alloys from nitrate solution feedstock provided by other DOE
 production facilities. Other operations included reprocessing of nuclear fuel, processing
,of polonium and actinium, and producing nuclear weapons components. Although the
 Laboratory has retained many of the original research programs dealing with national
 defense, its current mission has been expanded to include research in emerging
 technologies pertaining to biomedicine, space nuclear systems,  materials sciences,
 computational sciences, and environmental management.

Nevada Test Site. The Nevada Test Site (NTS) is 65 miles northwest of Las Vegas and
 occupies 1,350 square miles, making it the largest facility hi the DOE complex. NTS has
 been the primary site for atmospheric and underground nuclear  weapons testing by DOE,
 with more than 300 nuclear tests conducted above and below ground at NTS  and at seven
 other locations outside Nevada. Since 1963, all U.S. nuclear weapons tests at NTS have
 been conducted underground. In addition to weapons testing, NTS is also used for low-
 level radioactive waste disposal.
                                   4-6

-------
Oak Ridge National Laboratory. Founded in 1942, the Oak Ridge National Laboratory
(ORNL) occupies about 2,900 acres within the Oak Ridge Reservation.  The Laboratory's
original mission was to produce and chemically isolate the first gram quantities of
plutonium for use hi nuclear weapons. With time, the scope of ORNL greatly expanded
to include production of other isotopes, fundamental research in a variety of scientific
disciplines, research pertaining to hazardous and radioactive materials, environmental
studies, radioactive waste management/disposal, and a wide range of educational
programs.

Y-12 Plant. Built in 1943 as part of the Manhattan Project, the Oak Ridge Y-12 Plant
occupies approximately 811 acres within the Oak Ridge Reservation. This facility
consists of some 250 buildings that house about seven million square feet of laboratory,
machining, and research and development areas. The initial mission of the Y-12 Plant,
which began operation hi November of 1943, was the separation and enrichment of U-
235 from natural uranium by an electromagnetic separation process. When gaseous
diffusion technology became the accepted process for uranium enrichment, the magnetic
separators were taken out of commission in 1946.  Since that time, the Y-12 Plant's
mission has shifted to the disassembly of returned weapons components, quality
evaluation for the existing stockpile of nuclear weapons, and supportive research in
engineering designs associated with production and fabrication of nuclear weapons
components.

K-25 Site.  The Oak Ridge K-25 Site occupies about 1,500 acres within the Oak Ridge
Reservation. The K-25 Gaseous Diffusion Plant was built as part of the Manhattan
Project during World War II to supply uranium hexafluoride for the production of highly
enriched uranium for nuclear weapons production.  Construction of the primary K-25
building started hi 1943 and was fully operable by 1956. Exclusive production of highly
enriched uranium for weapon production through 1964 was gradually replaced with
commercial-grade, low-enrichment uranium production for the emerging nuclear power
industry.  Because of the declining demand for enriched uranium, the K-25 Plant was
placed on standby in 1985 and was  permanently shutdown in 1987.

Paducah. Located in Kentucky, the Paducah Gaseous Diffusion Plant site occupies
approximately 750 acres  of federally-owned land.  The Paducah Plant was constructed hi
                                   4-7

-------
the early 1950s for the purpose of enriching uranium by the gaseous diffusion process.
Since 1991, the plant has only produced low-enriched uranium for use as fuel in
commercial nuclear power plants.

Portsmouth. The Portsmouth Gaseous Diffusion Plant is located on about 3,700 acres of
federally-owned land in Ohio. In spite of the existing gaseous diffusion program at K-25
facility in Oak Ridge and Paducah in Kentucky, the Portsmouth facility was built to meet
the demand for highly enriched uranium created by the emergence of nuclear submarine
reactors and low-enriched uranium for projected commercial nuclear power reactors.
Since 1991, the plant has produced only low-enriched uranium for use by commercial
nuclear power plants, and, since 1993, production operations were assumed by the United
States Enrichment Corporation, a government corporation formed under the Energy
Policy Act of 1992.

Rocky Flats. The Rocky Flats Environmental Technology Site (RFETS) covers 11
square miles located approximately 16 miles northwest of Denver. Its primary mission
was to produce plutonium and other components for nuclear weapons. Currently,
activities at RFETS include cleaning up contamination and waste from its past activities
and transitioning its facilities for alternative uses.

Savannah River Site. The Savannah River Site (SRS) is located in west-central South
Carolina and comprises approximately 310 square miles; its production facilities occupy
less than 10 percent of the total area. SRS was established by the Atomic Energy
Commission in 1950 for  the purpose of producing plutonium (Pu-239) and tritium for
nuclear weapons. SRS also produced other special isotopes (Cf-252, Pu-238, and Am-
241) to support research in nuclear medicine, space exploration, and commercial
applications. To produce these radioisotopes, metal targets were irradiated at the five
production reactors. Isotopes were recovered from irradiated targets at chemical
separation faculties also located at the SRS. Production reactors have operated during
different time periods. Current operation of chemical process facilities is limited to the
recycling of tritium and the extraction of Pu-238 used in space exploration.

Weldon Spring. The Weldon Spring Site consists of 229 acres, approximately 20 miles
west of St. Louis, Missouri.  The Weldon Spring Chemical Plant and the Weldon Spring
                                   4-8

-------
       Quarry occupy the site. It was part of a site used by the U.S. Army as an ordnance works
       in the 1940s. In the 1950s and 1960s, the Atomic Energy Commission used the site to
       process uranium ore in the Weldon Spring Chemical Plant. The plant was subsequently
       deactivated and no further activities were carried out at the Weldon Spring Site since
       remediation began in 1985.

Relevant data contained hi these four documents are briefly summarized below.  Estimates of
scrap metal quantities and-limited qualitative data are defined in terms of (1) existing scrap metal
inventories and (2) projected scrap metal inventories associated with future decommissioning of
DOE facilities.

Because significant gaps hi quantitalive information remain, an attempt was made to supplement
reported data by direct contact with DOE personnel.  Individuals contacted included key
administrative  staffs at DOE Headquarters and DOE Regional Offices, as well as persons in DOE
field offices. Field personnel included individuals with responsibilities related to scrap metal
decontamination., segregation, storage, environmental monitoring, and salvage and recycling
operations. In  most instances, direct contacts yielded only subjective information that explained
the approach and general methods used to arrive at the reported quantities of scrap metal.

4.1.2  Existing Scrap Inventories at DOE

Data Reported in 1996 MIN Report

DOE's first major undertaking to evaluate its materials management practices dates back to
January 1990 with the establishment of the Mixed Waste and Materials Management
Workgroup. To support the Workgroup effort, an attempt was made to define and inventory
Materials Mot Classified As Waste (MNCAW) and resulted hi the 1994 MIN Report (formerly
known as the MNCAW Report).  This and other reports have been combined and collated with
new data and analysis to provide information presented hi the 1996 MIN Report.

DOE defines "materials in inventory" as materials that are not currently hi use (Le., have not
been used during the past year and are not expected to be used within the coming year) and that
have not been set aside for national defense purposes. The Department identified 10 material
categories with significant quantities. The ten categories are divided into two subcategories;
scrap metal and equipment is  cited among non-nuclear materials (Table 4-1).
                                         4-9

-------
                    Table 4-1.  Groupings in DOE Materials in Inventory
          Spent Nuclear Fuel

          Plutonium and Other NMMSS-
          tracked Materials

          Natural and Enriched Uranium

          Depleted Uranium

          Lithium
° Sodium

o Lead

o Chemicals

° Weapons Components

o Scrap Metal and Equipment
Scrap metal comprises worn or superfluous metal parts or pieces, including but not limited to
structural steel and other metals from decommissioned buildings, scrap metals accumulated from
facility maintenance and renovation in the past, and scrap stored in scrap yards and lay-down
yards. Scrap metal includes metals that are clean and metals contaminated or activated with
radioactivity and/or contaminated with hazardous substances. Equipment considered in the MIN
Report is defined as all equipment used for construction, production, or manufacturing and all
associated spare parts and hand tools.

To estimate scrap material inventories, the Department recruited personnel from each DOE
Operations and Field Office and from designated Headquarters Offices. The MIN Scrap Metal
and Equipment Team sought information by means of site-specific surveys and whenever
possible extracted information contained hi various DOE databases.  MEM data collection was,
therefore, constrained by the need to use existing data sources with no authorization or resources
for new studies and new information. The report acknowledges its limitations and states:

      "... Because of limited data, this report does not attempt to capture the exact amount of
      each material hi inventory. Rather, it attempts to capture the  general magnitude of the
      inventory of each material." (Emphasis added)
                                         4-10

-------
Data Reported.  Despite acknowledged limitations, the 1996 MIN report is regarded as the
principal data source for scrap metal estimates for 10 of the 13 DOE facilities (Table 4-2).
Estimates for several facilities presented in Table 4-2 were not directly extracted from the 1996
MIN Report but represent interpolated values.

Table 4-2.    Existing Contaminated Scrap Inventories at DOE Sites (Source:  1996 MIN
             Report)
'i^^maA^^y^ •,>'•»'•:•
„ ":••••• '.,' /*" -T ..'.. ' /- '- ,,"?< ,
X% »v •'; •"* -; ' ' ,'••;•''".-.
Fernald
Hanford
Idaho National Engineering
Laboratory (INEL)
Los Alamos National Laboratory
(LANL)
Nevada Test Site (NTS)
Oak Ridge National
Laboratory (ORNL)
Y-12
K-25
Paducah
Portsmouth
Rocky Flats
SRS
Weldon Spring
Total
.^ • > « -mf,' *$'.
i'-'^^&ae^,"-!'
: {m^$iib;t<^eS^
4,218
377
727
Not Reported
264
1,129
9,065
29,357
48,374
8,914
Not Reported
13,183
Not Reported
115,508
The need for interpolation was due to the fact that only a few DOE sites provided complete
quantitative estimates that defined existing scrap metal inventories as clean or radiologically
contaminated.  Many facilities either provided only a partial breakdown or no breakdown with
                                          4-11

-------
regard to fractional quantities of contaminated versus uncontaminated scrap metals. In fact, the
largest percentage of DOE scrap metal (~80%) as reported in the 1996 MEN Report is designated
as "unspecified" with regard to radioactive contamination. For scrap metal inventories
designated as "unspecified," it was assumed that 88% of scrap metal was contaminated and 12%
was clean and not considered contaminated. The basis for this assumption is defined in Table
1-6, page  16 of Volume 2 of the 1996 MIN Report.
                                                     i
Of existing scrap stockpiles, Table 4-2 identifies that about 90 percent is currently located at five
sites. In descending order, they include: Paducah, the K-25 facility in Oak Ridge, the Savannah
River Site (SRS), the Y-12 facility in Oak Ridge, and Fernald.

Data Extracted from the 1995 HAZWRAP Report

In 1994, Martin Marietta Energy Systems, Inc., in support of the DOE's Hazardous Waste
Remedial Actions Program (HAZWRAP), conducted a study that assessed scrap metal
inventories and their economic values for 11 DOE facilities.  Collection of information on
amounts and locations of scrap metal within the DOE complex was pursued through three
independent but complementary methods.

A preliminary questionnaire was forwarded to key site personnel, which requested generic
demographic data pertaining to scrap metal management along with a DOE Scrap Metal Data
Sheet (Exhibit 4-1).  Key information sought by the questionnaire included (1) scrap metal type
(e.g., steel, alurninmn, copper, etc.), (2) contamination status, and (3) scrap quantities.
                                            =>p
A second source of information for developing estimates reported in the HAZWRAP Report
came from a thorough review of published reports and DOE databases. A total of 28 documents
were identified as pertinent.

Lastly, the Project Team visited the sites and met with personnel to examine storage areas and
document the locations and amounts of stored scrap metal. Confirmatory estimates of stored
scrap metal quantities were based on physical measurements of size and storage density of piles.
                                        4-12

-------
                                     EXHIBIT 4-1
                         DOE SCRAP METAL DATA-SHEET
D.O.E Site
Location

Description
Material Type

Aluminum
Copper
Monel
Zirconium
Lead
Stainless Sleel
Steel
Nickel



Data


Performed By


Symbol

A)
Cu
M
2
PD
SS
Fe
Hi



Density
*tcv.n.
169
555
550
796
710
431
491
556

X

Density vs Steel
.£.
0.34
1.13
1.12
1.62
1.45 "
1.0
1.0
1.13



Preparation State

Prepared 3x2x5
Unprepared
Mixed
Insulated
25 Box
Drum
Ingot
Sealand
Sheet


Symbol

P
u
M
Is
B
D
Jg
si
s
 Radioactive Contamination
Uncontaminated
Unknown
Material
Tyoe



1


Preparation
State






Radioactivity






Estimated
Volume Yds.


'



Scrap Density
Ton/Yd.






Amount
Tone






Comments


1

i

(Material Preparation
1 Tvoe State

i
i



Radioactivity






Number •>
Units






Unit Weight
(Tons)
"





Amount
Tons






Comments' '
|




i
Preparation Ecuipmenl (shears, shudders, elc.):




Ongoing Activities:






Commervis:
                                          4-13

-------
Scrap metal estimates reported in the 1995 HAZWRAP Report were either directly adopted or
updated and used in the 1996 MUST Report. As indicated in Table 4-2, scrap metal data for
LANL, Rocky Flats, and the Weldon Spring facilities were not fully discussed in the 1996 MIN
Report. A brief description of the management and current inventories of scrap metals at these
three sites, as reported in the HAZWRAP Report, is presented below.

LANL. Located in New Mexico, the DOE facility has an active scrap metal recycling program.
Existing scrap metal inventories are stored at several locations in small piles, the largest of which
                                    in
is about 1,800 metric tonnes.  The total quantity of contaminated scrap metal at LANL is
estimated at 3,099 metric tonnes.

Rocky Flats.  At Rocky Flats, contaminated scrap metal is stored in metal drums and boxes that
were inventoried in the Site Waste Management database. Because the material quantities could
not be examined using the methods previously described, information from the Waste
Management  Program was used to quantify amounts and metal types of scrap inventories. A
total of 24,543 metric tonnes of contaminated scrap metal was estimated.
                                                                                    \
Weldon Spring.  At the Weldon Spring site hi Missouri, scrap metal is located in two storage
areas. Contaminated metal scraps removed in the past from process piping'associated with the
Quarry and Chemical Plants are stored in the Temporary Storage Area (TS A) and in an eight-
acre lay-down area called the Material Storage Area (MSA). A total of 27,839 metric tonnes of
contaminated scrap metal was estimated.

4.1.3  Summary of Existing Scrap Inventories at DOE Sites

Table 4-3 summarizes current best estimates of contaminated scrap metal quantities currently
stored at 13 DOE facilities. Of these estimates, ten were derived from data presented hi the 1996
MIN Report.  The remaining three values were derived from information presented in the 1995
HAZWRAP Report.

Based on the most current data, it is estimated that existing inventories of scrap metal represent
about 171,000 metric tonnes.
                                         4-14

-------
           Table 4-3. Summary Estimates of Existing DOE Scrap Metal Inventories
                                    (Metric Tonnes)

• • f^ »• *" % ^ ;
• >•/ 	 ,,' ; , '.•
j^fe^:^.'l
Fernald
Hanford
INEL
LANL
NTS
ORNL
Y-12
K-25
Paducah
Portsmouth
Rocky Flats
SRS
Weldon Spring
Subtotal
TOTAL

" S fff V^^VT^ifJila^¥' "^Jf^l^lT* Jt/JVM'ClE £\*ri&±*&±i-ii'-*£r*' ''
'KaMjiriiixigi *>vj»p iVJ.^UJi V^UWiKlMV^c . .^
"7'jrtil^brt":.!
4,218
377
727
Not Reported
264
1,129
9,065
29,357
48,374
8,914
Not Reported
13,183
Not Reported
115,608
iiA^ww-^^c



3,099






24,543

27,839
55,481
171,089
4.1.4   Scrap Metal Inventory by Metal Type

Data collected in support of the 1995 HAZWRAP Report provided information regarding the
composition of scrap metal inventories. Quantity estimates were provided for seven forms of
scrap metal classified as: (1) carbon steel, (2) stainless steel, (3) copper and brass, (4) nickel, (5)
aluminum, (6) tin and iron, and (7) miscellaneous, which included lead, monel, and cast iron.
These data were reviewed and updated by the MEN Scrap Metal and Equipment Team. Table 4-4
summarizes data reported in the 1996 MIN Report.
                                         4-15

-------
                   Table 4-4. MEN Scrap Metal Inventory by Metal Type
                                    (Metric Tonnes)
Metal Type ;;'";•
Carbon Steel

Nickel

Stainless Steel

Aluminum

Copper and Brass

Tin and Iron

Miscellaneous

Total
(percent)
'$':'$$&&•;<%;
1,008

0

1,435

27

24

227

782

3,503
(2.5%)
^^iiMpS^''-
,11,437

0

5,392

14

1,483

0

6,537

24,863
(18.0%)
••••-iy^ifc;;
94,472

8,817

959

5,637

147

0

10

110,042
(79.5%)
'•••'••'•*''$&$%;;••'•!"
106,917
(77.2%)
8,817
(6.4%)
7,786
(5.6%)
5,678
(4.1%)
1,654
- (1.2%)
227
(0.2%)
7,330
(5.3%)
138,409
(-100)
Inspection of Table 4-4 identifies the fact that 3,503 metric tonnes of scrap metal were found to
be free of radioactive contamination. Moreover, an estimated 110,042 metric tonnes or about
79.5 percent of existing scrap had not been assessed for radioactive contamination and was
classified as "undetermined."
                                               - a*
                                               r%

To estimate the fractional quantity of "undetermined scrap" that can reasonably be assumed
contaminated and which must be added to the known quantity of contaminated scrap for a total
estimated value of contaminated scrap quantity, the following formula was used:
    Total contaminated =  (contaminated scrap) + (
   contaminated
contaminated + clean
x undetermined)
Table 4-5 provides best estimated values of existing scrap metal inventories that can reasonably
be assumed contaminated. Carbon steel comprises more than 77 percent of total tonnage, with
nickel, stainless steel, and aluminum representing the other major sources.
                                         4-16

-------
Table 4-5. Estimated Scrap Inventories by Metal Type Currently Stored at DOE Facilities
                                    (Metric Tonnes)
&»^r^^---i
Carbon Steel
Nickel
Stainless Steel
Aluminum
Copper and Brass
Miscellaneous
Total
^f/iDiEterala^: _j
11,437
0
5,392
14
1,483
6,537
24,863
^^^(^fSaBtoM
125,537
11,716
1,273
7,490
196
14
146,226
$ .»:•• *&*&<*•. <-?-'\
136,974
11,716
6,665
7,504
1,679
6,551
171,089
4.1.5   Future Scrap Metal Quantities at DOE

During peak periods of activities, the nuclear weapons complex represented more than 120
million square feet of building structures (DOE 95a).  These buildings include 14 large
production reactors with extensive support structures, research reactors and their associated
support structures, eight chemical processing plants (i.e., "canyons") with vast quantities of metal
                                                          v
piping, tanks, valves, motors, duct-work, and structural components, and an array of buildings
used for storage, milling, manufacturing, testing, assembly, and administrative activities.

With the end of the Cold War Era and the questionable need for additional nuclear weapons,
many of these structures will be decommissioned over the next several decades.  As of June
1995, DOE's Office of Environmental Restoration Decommissioning Inventory slated 865
structures for future decommissioning (U.S. DOE, Office of Environmental Restoration
Decommissioning Inventory, June 1995).

Several facilities are still awaiting final notification for deactivation and are not yet designated
for decommissioning.  As a result, assessments aimed at estimating future scrap generation at
some DOE sites have not been conducted for these facilities.
                                          4-17

-------
Site-Specific Estimates

For those DOE sites that have been identified for partial or total decommissioning, scrap
quantities are at best preliminary estimates that are based on limited and incomplete data.
Projected scrap estimates associated with future decommissioning activities were derived from
three reports that include the following facilities:

       EPA 95 Report
              Sites:  Fernald, Hanford, LANL, Rocky Flats

  •    MIN 96 Report
              Sites:  INEL, SRS

  •    DOE 95 Report
              Sites:  K-25, Paducah, Portsmouth

Combined scrap quantities from future decommissioning activities at these sites are estimated at
around 925,000 metric tonnes. Scrap sources and site-specific estimates for the ten facilities are
briefly defined below.

Hanford. To date, only modest attempts have been made to assess future scrap quantities
pertaining to decommissioning activities. However, quantities are expected to be substantial. As
of June 1995,250 buildings at Hanford had been slated for decommissioning. Massive amounts
of structural steel scrap will be produced during the decommissioning of these buildings. Also
included are other structures such as exhaust stacks, storage tanks, and river outfall structures as
well as carbon steel and stainless steel pressure vessels from the Clinch River Breeder Reactor
program.

Approximately 91,798 metric tonnes of scrap are likely to be generated during decommissioning
activities. The vast majority of scrap is expected to be carbon steel with significant amounts of
stainless steel, lead, aad aluminum.

Idaho NationaLEngineering Laboratory. Over the past 50 years, more than 50 nuclear reactors
(mostly small test reactors) have operated at INEL.  While some of these reactors and their
support buildings have already undergone decommissioning., others are targeted for future
decommissioning.  Many published DOE documents that cite scrap estimates were assessed in

                                         4-18

-------
EPA 95 and in the MEM 96 Report. Future decommissioning activities at DSfEL are estimated to
generate 33,486 metric tonnes of scrap metal. At this facility, carbon steel (55.7%) and stainless
steel (44.0%) represent nearly all projected scrap metal quantities.

Los Alamos National Laboratory.  LANL's Metal Inventory Report (INV 96) not only assessed
existing scrap metal inventories but identified future scrap metal quantities associated with
decommissioning activities, as well as for scheduled "upgrade" projects. In combination,
decommissioning and upgrade activities are estimated to generate 2,686 metric tonnes of scrap.

Fernald.  Fernald's production area includes 20 process facilities and supporting structures that
are obsolete and beyond their design life. In total, 128 buildings and 72 miscellaneous structures
have been identified for decontamination and decommissioning. The dismantling.of buildings,
process equipment, and structures is estimated to generate 135,623 metric tonnes of scrap.

Savannah River Site. This site includes five heavy water production reactors that were used in
the production of tritium and other weapon materials. All reactors have been shutdown and are
likely to be disassembled. Scrap associated with the decommissioning of the five production
reactors and support structures/systems is estimated at 3,054 metric tonnes with nearly equal
contributions of carbon steel and stainless steel. The fate of the two SRS chemical separation
plants and the many facilities that support them remains undetermined. The decommissioning of
these facilities would undoubtedly add substantial (but to date undefined) quantities of scrap.

Rocky^Flats. A literature search'in support of the EPA95 revealed the existence of only one
study that estimated future scrap quantities for Rocky Flats. A study by the Manufacturing
Sciences Corporation (MSC 94) stated that the decommissioning of Rocky Flats is expected to
generate about 1,003 metric tonnes of scrap metal from four buildings that will be used for the
National Conversion Pilot Project and an additional 25,300 metric tonnes from the other
buildings and site structures. Most scrap is likely to be contaminated with depleted uranium,
enriched uranium, and/or plutonium.

Oak Ridge. K-25 Facility. The K-25 facility is the first of three DOE gaseous diffusion plants
that are slated for decommissioning. Decommissioning of the K-25 site is estimated to take a
total of eleven years: two years of planning and nine years of decontamination and
                                          4-19

-------
decommissioning. Current projections are that decommissioning activities will be completed in
the year 2006.

Decommissioning will include removal of large quantities of metals associated with process
equipment, piping, and structural components. Principal contaminants include uranium and
daughter products, technetium-99, and trace quantities of neptunium-237 and plutonium-239.
DOE 95 identified a total quantity of 406,372 metric tonnes of recyclable metal but did not
identify the fractions of uncontaminated and contaminated scrap metal.

Personal communications with Gary Person (PERS 96), principal author of DOE 95 Report,
yielded the following estimates: of the total future inventory of 406,273 metric tonnes of scrap
metal, 193,666 are estimated to be free of contamination and about 212,706 are likely to be
moderately contaminated scrap that is considered suitable for recycling.

Portsmouth. Decommissioning of the Portsmouth gaseous diffusion facility is scheduled to
begin in FY 2007 (following completion of decontamination and decommissioning activities at
the K-25 facilityX with a completion date in FY 2015. The decontamination and
decommissioning of the three gaseous diffusion plants are purposely scheduled in series in order
to (1) learn from experience gained, (2) minimize yearly expenditures, and (3) provide a steady
stream of metal for recycle. DOE 95b reported the availability of 312,085 metric tonnes of total
scrap metal. Of this quantity, 189,072 metric tonnes are estimated to be contaminated metal that,
with decontamination, is considered suitable for recycling.

Paducah. The Paducah Gaseous Diffusion Plant is the third and last gaseous diffusion facility to
be decommissioned.  Current projections identify decommissioning to start in 2015 and end in
2023. The first major phase will be the removal and decontamination of major components (i.e.,
motors, cell housing, compressors, compressors and converters, piping and valves, electrical
equipment, and HVAC system) from the process buildings. Personal communication (PERS 96)
identified that of the total projected scrap metal inventory of 331,365 metric tonnes (DOE 95b)
about 230,886 are estimated to represent scrap that is considered suitable for recycling.
                                          4-20

-------
4.1.6   Summary and Conclusions

At its peak, the nuclear weapons complex consisted of 16 major facilities that represented more
than 120 million square feet of buildings. These buildings contain large quantities of equipment,
structural steel, and other metal components. Over the 50-year period, some of these buildings,
their ancillary facilities, and the equipment they housed have been renovated, replaced, and/or
demolished. Currently, more than 170,000 metric tonnes of low-level contaminated scrap metal
is stored at various facilities that are considered suitable for recycling.

Estimates of existing scrap quantities are mostly based on site-specific review of historical
inventory data and physical surveys of scrap piles. Quantity estimates of existing scrap
inventories can, therefore, be viewed with reasonable confidence.

Future scrap quantities are closely linked to projected decommissioning activities at DOE sites
that make up the nuclear weapons complex. At some sites, virtually all structures and their
contents will be dismantled and removed; at other sites decommissioning may be limited, and the
DOE will continue select operations considered crucial to national security or important to
national research. To date, decisions and commitments for decommissioning are not only
incomplete but, in instances where such decisions have been made, they remain both tentative
and are subject to change in scope and schedule. Consequently, estimates of future scrap
quantities are less certain.

In this report, future scrap estimates were based on currently scheduled decommissioning
activities at nine facilities that include Femald, Hanford, INEL, LANL, SRS, Paducah, Y-12 and
K-12.  Decommissioning of these facilities is estimated to yield more than 925,000 metric tonnes
of contaminated scrap metal that is derived from dismantling large production reactors, research
reactors, chemical processing plants, and a vast array of associated support facilities and
structures. With effective decontamination, this scrap metal is potentially available for recycling.

Table 4-6 provides summary estimates that represent existing scrap inventories and future scrap
associated with decommissioning activities. Of the more than one million metric tonnes of
scrap, about 85 percent represents carbon steel with near equal quantities of copper, nickel,
aluminum, and stainless steel representing the remainder. It is possible that these values may
underestimate the total scrap metal quantities due to the fact that current data pertaining to future
decommissioning activities are incomplete.

                                          4-21

-------
                        Table 4-6.  Summary Data for Existing and Future Contaminated Scrap at DOE Facilities*
                                                           (Metric Tonnes)
Site
Name
Fernald
Hanford
Idaho
LANL
NTS
ORNL
Y-12
K-25
Paducah
Portsmouth
Rocky Flats
SRS
Weldon Sp
TOTAL
Percent
of Total
Scrap Metal
Database
Volume
139,841
92,175
34,213
5,785
264
1,129
9,065
242,063
279,260
197,986
50,846
16,237
27,839
1,096,703
100.00
Metal Type
Al
„„
684
30
40
11
18
33
7,988
21,161
6,130
...
14
510
36,619
3.34
C, Steel
101,753
87,020
19,195
5,568
204
992
8,392
232,953
212,917
191,412
33,666
10,403
26,877
931,352
84.92
S, Steel
...
787
14,733
177
15
117
602
753
190
18
2,454
5,809
406
26,061
2.38
Copper
38,088
...
44
...
...
2
38
304
198
408
14,726
11
46
53,865
4.91
Nickel
„_„
24
44
---
17
...
...
— -
44,794
— -
-—
...
_„
44,879
4.09
Monel
...
_-_
	
...
_—
...
___
65
...
18
— -
...
...
83
0.01
Lead
...
291
110
...
...
...
...
...
...
-—
.—
••«.
...
401
0.04
Cast
Iron
...
«...
4
...
...
...
...
...
...
—
...
...
...
4
t
3.6E-6
Black
Iron.
...
...
7
...
-_-
...
...
...
...
-—
...
...
...
7
6.4E-6
Graphite
...
1,632
...

...
...
...
....
...
...
— -
...
...
1,632
0.15
Cu/Bras
s
...
5
8
...
2
...
...
...
...
— -
...
...
...
15
1.4E-5
Tta/Fe
...
1
2
...
1
...
...
...
...
...
.—
...
...
4
3.6E-6
Other
...
1,711
36
...
...
...
...
...
...
...
...
.•.
...
1,747
0.16
Misc.
_
2C
__
__
U
..
...
...
...
...
...
...
...
34
3.1E-5
Includes metal for which decisions regarding its disposition may be affected by an EPA recycling standard.

                                                                4-22

-------
4.2   POTENTIAL SCOURGES AND CHARACTERISTICS OF SCRAP METAL FROM
      THE COMMERCIAL NUCLEAR POWER INDUSTRY

The U.S. commercial nuclear power industry is represented by 123 reactor plants.  At present,
eight reactors have been shutdown; in the next two to three decades, most of the reactors
currently hi operation will have reached their projected forty-year lifetime. A great deal of
information and data has been compiled by the U.S. Nuclear Regulatory Commission (NRC)
and the individual utilities pertinent to the decommissioning of these facilties and the associated
quantities and characteristics of the scrap metal that will be produced.  Appendix A presents a
detailed summary of the pertinent information. This section summarizes the information
provided hi Appendix A.

A crucial factor affecting the quantity of scrap  metal and associated contamination levels is the
basic design of the reactor. The two types of reactors used hi the United States are the
pressurized water reactor (PWR) and the boiling water reactor (BWR). Of the 123 U.S.
reactor units, 40 are BWRs manufactured by General Electric (GE) and 83 are PWRs
manufactured by Westinghouse (W), Combustion Engineering (CE), and Babcock and Wilcox
(B&W).  In the  1976-1980 time frame, two studies were carried out for the U.S. Nuclear
Regulatory Commission by the Pacific Northwest Laboratory (PNL) that examined the
technology, safety, and costs of decommissioning large reference nuclear power reactor plants.
Those studies, NUREG/CR-0130 and NUREG/CR-0672 for a reference PWR and reference
BWR, respectively, reflected the industrial and regulatory situation of the time. To support the
final Decommissioning Rule issued hi 1988, the earlier PNL studies have been updated with
the recent issuance of NUREG/CR-5884, Revised Analyses of Decommissioning for the
Reference Pressurized Water Reactor Station and NUREG/CR-6174, Revised Analyses of
Decommissioning for the Reference Boiling Water Reactor Power Station.  These four
NUREG reports along with several other NRC reports and select decommissioning plans on
file with the Commission represent the primary source of information used to characterize
Reference PWR and BWR facilities and to derive estimates of scrap metal inventories for the
industry at large.
                                        4-23

-------
4.2.1  Summary Estimates of Contaminated Steel for Reference BWR/PWR and the
       Commercial Nuclear Industry

Table 4-7 presents summary data for contaminated steel potentially available for recycling.
Estimates for the Reference BWR and PWR were derived by summing component mass values
previously cited in Table A5-2 and Table A5-3, respectively, of Appendix A.  Estimates for
the entire commercial nuclear industry were derived by taking Reference BWR and Reference
PWR values and applying plant-specific scaling factors for each of the 40 BWRs and 83
PWRs. Approximately 600,000 metric tonnes of contaminated steel may become available
over time for recycling.  About 80% of the  contaminated steel is carbon steel, with stainless
steel representing the balance.
                Table 4-7.  Summary Data for Contaminated Steel Inventories
                              Potentially Suitable for Recycling
J
Contaminated Material*

,
Stainless Steel
• Low-level Contamination
(lx!07 dpm/100 cm2)
Carbon Steel
• Low-level Contamination
(<1 x 10s dpm/100 cm2)
• Medium-level Contamination
(1 x 10s to 1 x 107 dpm/100 cm2)
• High-level Contamination
OlxlO7 dpm/100 cm2)
TOTALS
Quantity (metric tonnes)

Reference
BWR*7
1,688
576

786

326

6,754
2,306

3,146

1,302

8,442

Reference
I»WR
-------
Because past and current regulatory release criteria (i.e., U.S. NRC Regulatory Guide 1.86;
NRC 74) are defined in activity levels per unit surface area, information cited in this section
has been presented in this fashion. However, for risk analysis pertaining to recycling of scrap
metals, a more meaningful approach is to express contamination levels in terms of activity per
unit mass.  This conversion requires the derivation of the average mass thickness (g/cnf) of
metal surfaces by the following equation:

                             ™  ,     ,  ,   i^    / ^ Surface Areas (cm2)
               Average Mass Thickness (g/cm*) =  ^  —	-	
                                                    2_,  Metal Mass (g)

For the contaminated systems/components previously identified for Reference BWR and PWR,
a weighted average mass density of 3.5 g/cnf for contaminated surfaces was estimated.  At a
density of about 8 g/cm3 for steel, this corresponds to an average thickness of about  4,4 mm
(0.17 inches).  This average mass thickness can now be readily applied to estimate the activity
level per unit mass of contaminated steel. For example, under the current interim release
criteria of 5,000 dpm/100 cm2 for beta-gamma emitters, the residual contamination on average
would correspond to about 14 dpm/g (or about 6.5 pCi/g.; or 0.23 Bq/g) of steel.

4.2.2  Contaminated Metal Inventories Other Than Steel

There are  significant quantities of metals and metal alloys other than steel that may be suitable
for recycling, including: (1) galvanized iron, (2) copper, (3) inconel, (4) lead, (5) bronze,  (6)
aluminum, (7) brass, (8) nickel, and (9) silver.  However, there exist no credible data in the
open literature regarding the estimated rraction(s) of jhese metal inventories that are likely to
be contaminated or the extent of their contamination? In the absence of reported data, a
reasonable approach may assume that the contaminated fraction among total plant inventories
of these metals parallels the contaminated fraction of carbon steel for Reference BWR and
Reference PWR.  Justification for this modeling approach is based on the fact that most of
these metals exist as sub-components of larger items consisting primarily of carbon steel.
From data cited in Appendix A, the percent of contaminated carbon steel suitable for recycling
to that of total plant inventory corresponds to 20% and 10% for Reference BWR and Reference
PWR,  respectively. The application of these values-yields-the contaminated metal quantities
(suitable for recycling) cited in Table 4-8. Due to physical differences and chemical properties
that affect corrosion and internal contamination, categorization of the contamination of these
metals using the methods used for steel is not appropriate.

                                          4-25

-------
           Table 4-8.  Summary of Contaminated Metal Quantities Other than Steel
                                     (metric tonnes)
Metal
Type
Galvanized Iron
Copper
Inconel
Lead
Bronze
Aluminum
Brass
Nickel
Silver
Reference' Facility
BWR
258
137
24
9.1
5.0
3.6
2.0
0.2
<0.2
PWR
130
69
12
4.6
2.5
1.8
1.0
0.1
<0.1
" ,/ , liidustry
MlBWRs
8,710
4,625
810
307
169
122
68
7
<7
AllPWRs
10,037
5,327
927
355
193
139
77
8
<8
Total
18,747
9,952
1,737
662
362
261
145
15
<15
4.2.3  Time-Table for the Availability of Scrap Metal from the Decommissioning of Nuclear
       Power Plants

For currently operating nuclear power plants, an operational period of 40 years is assumed.
Following reactor shutdown, a rninimurn of 10 years is assumed before significant quantities of
scrap metal would be available. Thus, for currently operating reactors, the earliest dates for
releasing scrap metal are defined by their startup dates plus 50 years.  Currently, there are
eight reactor units that have been permanently shutdown (Dresden-1 (1984); Indian Point-1
(1980); LaCrosse (1981); TMI-2 (1979); Humboldt Bay (1976); Trojan (1993); Rancho Seco
(1989); San Onofre-1 (1992); and Yankee Rowe (1992)).  A conservative assumption for these
facilities projects the release of scrap metal over a ten-year period between 2000 and 2009.

Table 4-9 summarizes the potential availability of scrap metal for yearly intervals  starting with
the year 2010.  The release of scrap metal based on this time-table should, however, be
considered highly conservative since many, if not most, facilities are likely to delay
decommissioning activities for varying portions of the allowable 50-year SAFSTOR period.
                                          4-26

-------
Table 4-9. Time-Table for Available Contaminated Scrap Metals from
              Decommissioned Nuclear Power Plants
                    Quantities (metric tonnes)
%
Year
2000-
2009
2010
2011
2012
2013
2014
2015
2016
2017
2018
2019
2020
2021
2022
2023
2024
2025
2026
2027
2028
2029
2030
2031
2032
2033
2034
2035
2036
2037
2038
2039
CS
15,377
804
—
-_
—
3,616
3,616
—
, 6,464
14,811
1,763
15,442
2,438
10,328
38,415
41,117
12,927
22,927
5,611
- 9,574
—
9,078
8,961
10,697
11,191
30,466
26,173
32,396
13,192
9,637
8,366
" SS ""
4,107
105
, -
—
—
475
475
—
1,958
2,712
534
2,740
739
1,674
8,571
8,855
3,175
5,326
1,700
2,185
—
2,750
2,714
1,406
2,468
5,672
6,206
6,248
3,996
2,919
1.639
/
, <3alv< ;
Iran
609
31
	
—
—
140
140
—
257
580
70
604
97
370
1,512
1,616
510
903
223
377
—
362
357
415
440
1,193
1,032
1,269
525
384
328
- Copper.
323
17
__
—
—
75
• 75
—
136
308
37
321
51
197
439
859
271
480
118
200
—
192
189
221
234
635
548
675
278
203
174
Jnconel
57
3
	
—
—
13
13
—
24
54
7
56
9
34
141
151
48
84
21
35
—
34
34
38
41
111
96
118
49
36
30
Lead.
21
1
—
...
—
5
5
—
9
20
2
21
3
13
53
57
18
31
8
13
—
13
12
15
16
42
36
45
18
13
12
Bronze
12
<1
	
_
—
3
3
—
5
11
1
12
2
7
29
31
10
17
4
7
. 	
7
7
8
8
23
20
24
10
7
6
Alum.
8
<1
	
—
	
2
2
	
4
8
1
8
1
5
21
22
7
12
3
5
—
5
5
6
6
16
14
17
7
5
4
/Brass
5
<1
	
—
	
1
1
	
2
4
<1
4
<1
3
12
12
4
7
2
3
_-_
3
3
3
3
9
8
10
4
3
2
Nickel
0.5
_—
__-
«.
	
	
— _
	
0.2
0.5
0.1
0.5
0.1
0.3
1.2
1.3
0.4
0.7
0.2
0.3
	
0.3
0.3
0.3
0.3
1.0
0.8
1.0
0.4
0.3
0.2
                             4-27

-------
Table 4-9.  Time-Table for Available Contaminated Scrap Metals from
        Decommissioned Nuclear Power Plants (Continued)
                    Quantities (metric tonnes)
Year
2040
2041
2042
2043
2044
2045
2046+
Total
CS
12,956
mif1.
3,261
„„+.
„_„
2,703
12,868
397,175
SS
3,925
	
988
	
— _
819
3,902
90,983
Gafv,
fcoa H
516
	
130
—
—
107
512
15,609
Copper
273
—
69
—
—
57
271
7,926
Incon&l
48
—
12
—
—
10
48
1,455
Lead
18
—
5
—
—
4
18
547
Bronze
10
—
2
—
—
2
10
299
Alum,
7
—
2
—
—
1
7
212.
(•
Brass
4
	
1
i 	
	
1
4
121
Niefcei
0.4
	
0.1
	
—
0.1
• 0.4
12.2
                             4-28

-------
                                   REFERENCES

DOE 95a     "Closing the Circle on the Splitting of the Atom, "U.S. Department of Energy,
             Office of Environmental Management, January 1995.

DOE 95b     "Gaseous Diffusion Facilities Decontamination and Decommissioning Estimate
             Report," prepared by G. A. Person, et al, Environmental Restoration Division,
             Oak Ridge, TN for U.S. DOE, Office of Environmental Management,
             ES/ER/TM-171, December 1995.

EPA/SCA 95 "Analysis of the Potential Recycling of Department of Energy Radioactive Scrap
             Metal" prepared by S. Cohen & Associates, Inc. for the U.S. Environmental
             Protection Agency, Office of Radiation and Indoor Air, August 1995.

HAZ 95      "U.S. Department of Energy Scrap Metal Inventory Report for the Office of
             Technology Development, Office of Environmental Management," prepared by
             Hazardous Waste Remedial Actions Program for the Department of Energy,
             DOE/HWP-167, March 1995.

INV 96      "Los Alamos National Laboratory (LANL) Metal  Inventory," Los Alamos
             National Laboratory, August 1996.

MUST 96      "Taking Stock: A Look at the Opportunities and Challenges Posed by
             Inventories from the Cold War Era," U.S. Department of Energy, Office of
             Environmental Management, DOE/EM-0275, January 1996.

MSC 94      "National Conversion Pilot Project, Stage I, Preliminary Market Analysis
             Report," Rev. 1, Manufacturing  Sciences Corporation, June 1994.

NUREG/     "Technology,  Safety and Costs of Decommissioning a Reference Pressurized
CR-0130,     Water Reactor Power Station," Vol. 1, prepared by Smith, R.I., et al., Pacific
1978        Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/     "Technology,  Safety and Costs of Decommissioning a Reference Boiling Water
CR-0672,     Reactor Power Station," Vol. 2,  Appendices, prepared by Oak, H.D., et al.,
1980        Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/     "Revised Analyses of Decommissioning for the Reference Boiling Water
CR-6174,     Reactor Power Station," Vol 2, Appendices, prepared by Smith, R.I.,
1994        et al, Pacific Northwest Laboratory for the U.S. Nuclear Regulatory
             Commission.
                                        4-29

-------
NUREG/     "Revised Analyses of Decommissioning for the Reference Pressurized Water
CR-5884,     Reactor Power Station," Vol 1, Main Report, prepared by Konzek, GJ, et al.,
1995          Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

PERS 96      Personal communications with G. A. Person, principal author of the "Gaseous
             Diffusion Facilities Decontamination and Decommissioning Estimate Report,"
             Lockheed Martrin Energy Systems, Inc., October 26, 1996.

NRC 74      U.S. Nuclear Regulatory Commission Regulatory Guide 1.86,  1974,
             "Termination of Operating Licenses for Nuclear Reactors," U.S. Nuclear
             Regulatory Commission.
                                        4-30

-------
                                      CHAPTERS

            DESCRIPTION OF UNRESTRICTED RECYCLING OPERATIONS

5.1    INTRODUCTION

5.1.1   Recycling Scrap Steel—An Overview

Figure 5-1 presents a simplified schematic diagram of the steps that would be involved in
recycling carbon steel scrap into consumer or industrial products. The process starts with steel
scrap that already exists hi scrap piles at various DOE and perhaps at NRC-licensed facilities, or
that will be generated in the course of the decontamination and decommissioning (D&D) of such
facilities. Scrap that has been cleared for release is loaded onto trucks to be transported off site.
As indicated in Figure 5-1, any required decontamination takes place within a radiation
controlled area (RAC). All operations within the RAC are performed by occupationally exposed
workers, who are subject to DOE- or NRC-regulated exposure limits and ALARA procedures.
Therefore, the exposures of these workers are not considered in the present analysis.1

The scrap is transported to a processor where it is unloaded, sorted and possibly cut up or
compacted. The processed scrap is transported to a steel mill where it may be unloaded to a
scrap pile or sent directly to the furnace. In either case, it is loaded into a charging bucket and
charged to an electric arc furnace (EAF), where it is melted.

Certain constituents of the furnace charge are either volatilized or entrained in the air stream as
particulate matter. Most of these emissions are captured by the emission control system and
routed to the baghouse, where the fumes are cooled and filtered.  The filters, which are hi the
form of long bags, are periodically emptied by remotely operated mechanical means. The dust is
transferred to a tanker truck and shipped off site.
    1 Operations indicated by cross-hatched boxes in the diagram are not modeled in the present analysis.

                                           5-1

-------

                                       •ftrea\;.\::^:-;:V:v:j

::fc:v£w&>'L
«. «5* ""A««A'.'A«LVA
          Charge Scrap
           to Furnace
 ^*X"H\vIVAV'V.V.,.,,...-. .W.V."..'.".Jvv,V..Vji f." V.
      f ,*,V.VV.'."VV,V ". "V
 i!KIVJ**S'.v.A1.1

iilMiSs
           Melt in EAF



           Continuous
             Caster
                                        heet Metaf
                               ;;:.:;;;:%:ixSleel.W:.;::..   .  |    v; ScrapPraoes
                                                                         Release to
                                                                        Atmopshere
Kitchen Range
                            Figure 5-1. Operations Analyzed
                                       5-2

-------
After the scrap is melted, first the slag and then the molten steel are poured into ladles. The
molten steel is transferred from the ladle to a tundish from which it is fed to a continuous caster,
where it is made into slabs. These may be sold as such or made into interim mill products, such
as coils of sheet metal.

The slag is transported to a slag pile at the steel mill, where it is stored prior to shipment to a slag
processing facility. The slag processor sells the slag for various uses, such as ballast for road-
building or aggregate which is mixed with cement and used for paving. While the slag is stored
at the mill, various components may leach out and percolate through the soil to an underlying
aquifer, possibly contaminating an underground source of drinking water.

5.1.2   Reference Facility

In the United States, most steel scrap is  melted in either an EAF or a basic oxygen furnace
(BOF). The charge for an EAF usually consists entirely of scrap, while scrap makes up less than
30% of the feedstock of a BOF, the rest being the pig iron output of a blast furnace. It is
therefore possible for a given EAF charge to consist exclusively of scrap from nuclear facilities,
while in the case of a BOF, the scrap would be diluted by pig iron. A steel mill equipped with
EAFs was therefore selected as the reference mill for the present study.

It is unlikely, however, that, for an entire year,2 any steel mill would be exclusively supplied with
scrap resulting from the dismantling of components that had been potentially exposed to
radioactive contamination. To determine the largest fraction of the scrap that could be
potentially contaminated, the anticipated release of scrap metal by various generator sites
nationwide was matched to the scrap processing capacities of nearby steel mills. It was found
that four nuclear power plants in a single locality are scheduled for decommissioning in the same
year. Metal workers would receive the highest likely exposures during that peak year if all the
recyclable scrap metal from these plants were taken to a single scrap processor and melted at a
small nearby steel mill. A detailed report of the study is found in  Appendix G.

The reference steel mill for-the present analysis was based partly on the Calumet Steel Co.
facility in Chicago Heights, IL, which is described in greater detail in Appendix H.  The mill is
   2 The present analysis assesses the radiological impacts during the year of peak exposure.

                                           5-3

-------
equipped with two EAFs, each of which has a 12.5-foot diameter shell and produces a 32-ton
average heat, with a nominal capacity of 75,000 tons per year. Other parameters used in the
analysis are based on data pertaining to other facilities, on engineering judgment, and on
analytical assumptions.  Therefore, the results of the present analysis are not intended, and
should not be used, to predict the radiological impacts of recycling scrap metal on any real
individual or facility, now or hi the future.

5.1.3  Exposure Pathways

The exposure pathways considered in the present analysis fall into two general groups:  external
exposure to direct penetrating radiation and internal exposure from inhaled or ingested
radionuclides.

External Exposure

The external exposures are evaluated by use of the MicroShield™ computer code, which is
described in more detail in Section 6.3.1, or by dose coefficients adapted from Federal Guidance
Report (FOR) No. 12 (Eckerman 93).

MicroShield™ computes the exposure rate from a uniform distribution of one or more
radionuclides within  a specified matrix, such as a solid cylinder of iron, with additional  shielding
material between tine source (i.e., the iron cylinder) and the receptor point. MicroShield™
includes attenuation and build-up factors for nine metallic elements as well as air, concrete and
water. In addition, it is possible for the user to create custom materials by specifying the
densities and elemental compositions of the new material. However, the present analysis uses
iron to represent the various steel alloys in both the source material being processed and in the
components of the furnace that act as radiation shields. Since carbon steel contains over 98%
iron, it is preferable to model it as pure iron, as the build-up factors for iron are based on actual
measurements.

Examples of the MicroShield™ external exposure assessments of all relevant scenarios  are
shown in Appendix H.
                                           5-4

-------
Internal Exposure

The internal exposure pathways consist of the inhalation of radioactively contaminated dust, the
incidental ingestion of contaminated dust, soot or other loose, finely divided material, and the
ingestion of contaminated food or water.

The following sections describe the geometries and the materials used to model the external
exposure from each task, as well as the assumptions regarding the inhalation and ingestion
pathways. A detailed discussion of the last two pathways appears hi Sections 6.3.2 and 6.3.3.

5.2    LIST OF OPERATIONS AND EXPOSURE SCENARIOS

Table 5-1 lists the operations and exposure parameters employed hi the assessment of potential
radiological impacts of recycling scrap metal on the RMEI. These operations and the parameters
used to model the corresponding exposure scenarios were based on an earlier EPA study on the
recycling of DOE scrap metal (SCA 95). That study comprised over 60 exposure scenarios,
which were selected on the basis of reviews of the literature, including work done for the
International Atomic Energy Agency (IAEA 91) and the U.S.  NRC (O'Donnell 78), as well as
observations of current scrap metal recycling practices made during visits to steel mills and scrap
processors, and on the basis of data furnished by personnel of these facilities. The present
analysis encompasses those operations shown to have the maximum potential impacts on the
exposed individuals. The new exposure scenarios utilize information obtained in the course of
additional visits to steel mills and scrap processors, discussions with steel industry personnel,
further research into scrap metal recycling, and the professional judgments and experience of the
project team, whose combined professional experience encompasses health physics, industrial
hygiene and metallurgy.

As seen in Table 5-1, the study also assesses the radiation exposures from several representative
finished products which might be made from recycled scrap.3  These products were selected on
the basis of their wide use and their potential radiological impacts on individuals and/or
   3 Three of these products are made from cast iron, which is produced by a different process than is used to make
carbon steel. Since the radiological impacts of iron founding are not included in the present study, these products are not
represented in Figure 5-1. However, the contaminant distributions characteristic of cast iron are utilized in the impact
assessment of these products (see Section 6.2 and Appendix F).

                                           5-5

-------
population groups—they are comparable to the finished products in the earlier studies.  For many
radionuclides, the impacts on end users would be dominated by exposure to external radiation.
Therefore, the highest impacts would be produced by massive products that are in close
proximity to the exposed individuals for the longest times. Cooking utensils were included to
assess radiation exposures from consumption of food potentially contaminated by radionuclides
leaching from the metal into the food during cooking.

5.2.1  Dilution Factors

As discussed above, potentially contaminated scrap would in most cases be diluted with scrap
that had never been exposed to radioactive contaminants. The ratio of the total amount of metal
to the potentially contaminated scrap—termed the dilution factor—is listed for each of the four
major groups of operations shown in bold-faced upper-case type in Table 5-1.

Scrap Processing

As was previously stated, all the recyclable scrap metal from four commercial nuclear power
plants could plausibly be sent to a single scrap processor hi a single year. However, as  cited in
Appendix G, only 13% of such scrap would be potentially contaminated. Therefore, the average
specific activity of any given radionuclide that in the scrap being processed during the course of
the year would be only 13% of its average specific activity in the potentially contaminated
scrap.4 The dilution factor for the scrap processing operations is thus:
                                         0.13
                                               =  7.7
    4 Throughout this report, the terms "radionuclide concentration" and "specific activity" may appear to be used
interchangeably. Strictly speaking, concentration refers to a given physical quantity, such as mass, per unit volume or unit
mass of the matrix.  The concentration of uranium in slag, for example, might be expressed in micrograms of uranium per
gram of slag. Specific activity is always expressed in units of activity per unit mass, such as pCi/g. For a given
radionuclide, of course, the specific activity is proportional to its concentration. Since radionuclides are usually detected
and assayed in terms of their activities, not in terms of their masses, specific activity is a more useful concept.

                                             5-6

-------
   Table 5-1.  Operations and Exposure Parameters for Radiological Assessments of Individuals
,
Description
Truck driver transporting scrap
' Mnenioaie
SCRPDRVR
Dilution
factor
7.7
, ,". , •, , Expossreg-atlHyays ,„„ „.,,„, ,
External Ex]po|ar-& '
Time
Qxfy)
2000
Distance
8ft
Medium
scrap
Internal
Time
(fer/y)
Medium.:
Dost
Joad

-------
Steel Mill

The two EAFs at the reference steel mill have a total nominal capacity of 150,000 tons per year.
However, as discussed in Appendix G, the postulated decommissioning of the four nuclear
power plants would only yield a total of 132,000 tons of scrap, of which 13%, or about 17,000
   Sj would be potentially contaminated.  These 17,000 tons represent approximately 11% of the
mill's annual capacity. Thus, the dilution factor for the steel mill operations is - =9,1.
                                                                       0.11

Use of Mill Products

It was assumed that the three operations using mill products modeled in this analysis obtained all
their materials from the reference mill. Thus, the materials are assigned the same dilution factor
as the steel mill operations.

End Users

Since any one item could be made from a single heat which could consist only of potentially
contaminated scrap, the dilution factor for the end user scenarios is equal to 1.

5.2,2   Scrap Processing Operations

Assessments were performed on two workers involved in processing scrap. One is a truck driver
who spends eight hours per day in the cab of a truck carrying 21 -ton loads of scrap metal to the
scrap processor. His only exposure would be to external radiation from the load of contaminated
scrap. Another is a worker who spends six hours per day cutting the scrap, but spends a total of
seven hours in canyons surrounded by scrap. He would also inhale and ingest dust which is
assumed to have the same specific activity as the scrap.

5.2.3   Steel Mill

Most mills that process scrap metal receive the scrap via truck or rail. Upon arrival at the mill,
the scrap is unloaded, charged into an EAF and melted. Steel mills typically minimize the
unloading scrap into piles because of the extra cost of reloading the metal for transportation to
the furnace.

                                           5-8

-------
Furnace Operations

The scrap-bearing container is unloaded by means of a large electromagnet and dumped into
charging buckets that move the scrap to the furnace. The exposures of two workers performing
representative tasks involved with furnace operations are assessed in the present analysis.  One is
the crane operator who transfers the charging bucket—he may be exposed to external radiation
from the scrap in the bucket. The other is the furnace operator, who may be exposed to radiation
from the scrap in the furnace while it is melting. They both may be exposed to fugitive furnace
emissions which escape capture by the emission control system. Such emissions may also lead
to radiation exposures of the population living near the facility via the following pathways:

       •     External exposure from immersion in a plume of radioactive emissions
       •     External exposure from radioactive particles deposited on the ground
       •     Inhalation of radioactive emissions
             Consumption of vegetables raised on contaminated soil
       •     Consumption of milk and beef from cattle raised on contaminated forage

Interim Products

Once the steel melts, it is poured either into ingot molds or onto a continuous caster which
produces steel slabs.  At a continuous caster, torches cut the slabs into smaller pieces as the steel
runs down a set of rollers. Cooled slabs are stored, reheated and formed into products such as
coils of sheet metal, or are sold as such. The operator  of the continuous caster may be exposed to
external radiation from the molten steel in the tundish  as well as from the slabs cast in the
continuous caster. He may also be exposed to fugitive furnace emissions.

Baghouse

The baghouse contains rows of filters, suspended from the ceiling, that trap the various effluent
emissions from the melt-refining process.  These bag-like filters are continuously shaken; the
dust settles into collecting hoppers and is fed by a screw mechanism into a tanker trailer. Once
filled to capacity, the trailer is  transported away from the steel mill to a processing facility for
recovery of commercially valuable components, primarily metals such as zinc, cadmium and
lead, and for ultimate disposal.
                                          5-9

-------
Steel mill workers are occasionally assigned to spend a day repairing or changing the baghouse
filters.  Such a worker typically spends four to six hours hi the midst of the suspended filters in
the dust-laden atmosphere of the baghouse,5 wearing a respirator equipped with a half-mask face
piece. At a typical facility, this procedure is carried out an average of seven times per year. The
analysis assumes that the same worker is assigned to this task every time.  While performing
such maintenance, the worker may be exposed to external radiation from the residual dust that is
retained in the filters after they are emptied, as well as to the dust that has settled on the floor of
the baghouse.

In addition, one worker typically spends about one hour per day monitoring the control
mechanisms and performing maintenance that does not require entering the modules containing
the filters.  It is conservatively assumed that the same worker who maintains the filters would be
assigned to this duty on days he were not inside the modules. The rest of the time, he would be
assigned a variety of tasks in the steel mill.  His external exposure rate during that time is
assumed to be the same as that of the crane operator.6 His internal exposure rate is assumed to be
the average of workers inside a typical mill (see Appendix H).

The driver of the tanker truck transporting the dust off site may be exposed to external  radiation
from the dust hi the truck. At one EAF facility which was visited, the dust is shipped to a
processing facility about 60 miles away. Assuming an average speed of 40 miles per hour, one
trip takes VA hours. Since the reference facility operating at full capacity produces about 2,250
tons of dust per year, a truck carrying 25 tons of dust would make 90 trips per year.
Consequently, the driver would spend an average of 135 hours per year transporting this dust.
                                                _i n*
                                                *T
Slag Disposal

After the completion of the melt cycle, the EAF is tilted and the slag is poured into a ladle, which
is moved by overhead crane to a slag yard outside the building. A worker at a typical facility
spends about ha If his time on a platform on the edge of the slag yard and may be  exposed to
    5 Rest periods necessitated by work in a confined area and the need to don and remove protective clothing restrict the
amount of time the worker can spend on this task.

    6 This worker was selected as having the median exposure rate to Co-60, one of the significant radionuclides in the
present analysis.

                                           5-10

-------
external radiation from the slag. Since the rest of his time is in the vicinity of the slag, he would
be exposed to slag dust during the course of the day.

Since the slag pile is exposed to the elements, soluble components of the slag leach out of the
matrix and percolate through the soil until they reach an underlying aquifer. (This process takes
a number of years—see Section 6.4. L) A nearby resident who gets his drinking water from a
well that is down-gradient from the slag pile may, at some tune in the future, be exposed to
contaminated ground water.

5.2.4  Use of Steel Mill Products

All products of the steel mill have industrial uses.  The present analysis deals with, two of these
products:  finished steel and slag. In addition, most baghouse dust is reprocessed to recover zinc
and other valuable metals.

Slag

As shown in Appendix I, slag is primarily used in road building, as fill, or for soil conditioning.
A worker employed in road construction may be exposed to external radiation from the slag hi
the roadbed as well as that in the cement pavement—he may also be exposed to contaminated
slag dust.  It is assumed that the fraction of the slag generated by the melting of residually
contaminated scrap is the same as the fraction of contaminated scrap hi the total scrap melted by
the mill during the peak year (i.e., the dilution factor of slag is 9.1, the same as that of the.steel).

Steel

Steel is used to make a virtually endless variety of finished products. The analysis considers the
four categories of products which are listed below, along with a representative component or
example of each category. These products also represent small, medium and large objects, as
indicated below.
                                          5-11

-------
       * Large home appliance (medium-sized object):  double oven
       « Automotive components (medium-sized objects): engine block, body shell7
       • Large industrial equipment (large object): 8-ton metal-working lathe
       • Cooking utensil (small object): flying pan

Only the the oven and auto body are made from carbon steel, however.  The other three are made
primarily of cast iron, which is produced by a different process. The radiation exposures of
workers producing and assembling two of these products—engine blocks and industrial
lathes—are assessed in the present analysis. In each case, the workers may be exposed to
external radiation from the iron, which is assumed to have the same concentration factor as the
steel. The grinding operations on the lathe bed may also expose the lathe maker to the inhalation
and ingestion of iron dust

End Users of Finished Products

The final group of exposed individuals is composed of people who use the products listed in the
previous section. One representative user of each of the four products is included in the analysis.
A consumer may be exposed to external radiation from the steel in the kitchen range.  A taxicab
driver may be exposed to external radiation from the shell of the auto body, while a lathe
operator may be exposed to radiation from the cast iron lathe bed. Another consumer cooking
food in a cast iron trying pan may be exposed to external radiation from the cast iron, in addition
to eating food which may be contaminated with residual  radioactivity that has leached from the
pan.
   7 The engine block was selected to be the representative automotive component for the manufacturing process
because it comprises a large portion of the total mass and because an assembly worker spends much of his work day in
close proximity to this source. Because of its large area, the sheet metal in the auto bodies has a greater potential for
external exposure of the occupant: of the automobile, and was therefore selected as the representative automotive
component for the end user exposure scenario.

                                           5-12

-------
                                   REFERENCES

Eckennan 93  Eckennan, K. F., and J. C. Ryman, 1993. External Exposure to Radionuclides in
             Air, Water, and Soil, Federal Guidance Report No. 1^, EPA 402-R-93-081. U.S.
             Environmental Protection Agency, Washington, DC.
                             •                             /
IAEA 91     International Atomic Energy Agency, 1991. "Exemption Principles Applied to
             the Recycling and Reuse of Materials from Nuclear Facilities", Draft
             (unpublished).

O'DonnellTS O'Donnell, F. R., et al. 1978. Potential Radiation Dose to Man from Recycle of
             Metals Reclaimed from a Decommissioned Nuclear Power Plant, NUREG/CR-
             0134, Oak Ridge National Laboratory, Oak Ridge, TN.

SCA 95      S. Cohen & Associates, Inc.  Analysis of the Potential Recycling of-Department of
             .Energy Radioactive Scrap Metal. U.S. Environmental Protection Agency, Office
             of Radiation and Indoor Air, Washington, DC.
                                        5-13

-------
Page Intentionally Blank

-------
                                     CHAPTER 6

           CALCULATION OF RADIOLOGICAL IMPACTS ON INDIVIDUALS

Chapter 5 presented the scenarios and modeling parameters used to assess the radiation
exposures of individuals that may result from recycling steel scrap from nuclear facilities.
Chapter 6 discusses how these scenarios, as well as the effluent gaseous emissions from the
facility, are used to perform a radiological assessment of these individuals. For the sake of
clarity in the presentation, the scrap dilution factor presented in Section 5.2 is not discussed in
the present chapter, which assumes that all recycled scrap is potentially contaminated.  The scrap
dilution factor will be explicitly addressed in the discussion of results in Chapter 7.

The concept of the RMEI is central to the assessment—it is discussed here in more detail.  For a
single exposure scenario and a given radionuclide, such as scrap contaminated with a strong y-
emitting nuclide (e.g., cobalt-60 or cesium-137), the choice of the RMEI is relatively
straightforward: it is the individual worker who spends the most time nearest to the scrap. For
the entire life cycle of a given batch of scrap metal—from the time it leaves the custody of a
licensed facility, is transported to a steel mill, is turned into sheet metal, is used to fabricate a
kitchen range and finally is delivered, to a home—there may be several exposed individuals. The
RMEI is not obvious a priori. To determine which individual receives the highest exposure, the
annual doses to the exposed individuals at each stage of production, transportation, distribution
and storage, including the use of the finished product, are compared. The person with the highest
dose rate would become the RMEI for a given radionuclide.

A number of computer codes dealing with recycling and pathways analysis were reviewed for
use in this study but none were found entirely suitable. Initially, a series of computer
spreadsheets were developed to perform the calculations described in this chapter.  As the
analysis progressed, the need for a single integrated computer program became evident Such a
program was therefore developed for this analysis. The program is written in the FORTRAN 90
computer language and can run on an IBM-compatible personal computer.

6.1    RADIOACTIVE CONTAMINANTS

The 40 individual radionuclides studied in this analysis were selected on the basis of a review of
nine published reports which cast light on the raiclides most likely to be present in scrap metal

                                          6-1

-------
that may be a candidate for recycling. A detailed discussion of the selection process is presented
in Appendix D.

Since a period of years is assumed to elapse between the time the metal was contaminated and
the time it would be recycled, short-lived nuclides (i.e. those with half-lives of less than 6
months) would have decayed to insignificant levels and were therefore omitted from the present
analysis.  By the same token, short-lived progeny of long-lived parents are assumed to be in
secular equilibrium, with their parents at the time of recycling. All references to such parent
nuclides in this report include the designation "+D" to indicate that the contributions of this
implicit progeny are included in the calculated annual doses and risks, which are normalized to
unit specific activity of the parent.  The implicit progenies of all nuclides selected for the present
analysis are listed in Table 6-1. The generation number indicates whether the progeny nuclide is
first generation (1), second generation (2), etc.

The analysis also considered steel scrap potentially contaminated with unique combinations of
radionuclides, including long-lived members of natural decay series in secular equilibrium with
their parents. These include: 1) "U-separated"—the three uranium isotopes (in secular
equilibrium with their short-lived progenies but separated from their long-lived progenies) in the
ratios of their natural abundances; 2) "U-depleted"—the same isotopes in ratios characteristic of
depleted uranium;1 3) "U-natural"—natural uranium hi secular equilibrium with the entire U-238
and U-235 radioactive decay series, and 4) "Th-series"—Th-232 in secular equilibrium with its
entire decay series.  The calculated radiological impacts of the mixtures of uranium isotopes, as
well as those of the uranium series, are normalized to unit activities of U-238, while those of the
thorium series are normalized to unit activities of Th-232. The nuclides included in each of these
groupings are listed in Table 6-2.
     Depleted uranium is a byproduct of the uranium enrichment process and contains reduced activities of U-234 and
U-235.

                                           6-2

-------
                 Table 6-1.  Implicit Progenies of Nuclides Selected for Analysis3
' "Pafemt
Nucli.de
Sr-90
Ru-106
Ag-llOm
Sb-125
Cs-137
Ce-144
Pb-210
Ra-226
Ra-228
Ac-227
Th-228
Half-Life
28.6 y
1.01 y
249.8 d
2.77 'y
30.2 y
284 d
22.3 y
V
1600 y
5.75 y
21.8 y
1.91 y
Radiation
P
P
P,Y,e-
P,Y,e-
P
P,Y,e-
P,Y,e-
a,Y,e-
P
cc,p,Y,e-
a,Y,e~
- Progeny' ' '
Generation ;
1
1
1
1
1
1
1
1
2
1
2
3
4
5
1
1
1
2
3
4
5
6
7
7
1
2
3
4
5
6
6
NacJide
Y-90
Rh-106
Ag-1 10
Te-125m
Ba-137m
Pr-144m
Pr-144
Bi-210
Po-210
Rn-222
Po-218
Pb-214
Bi-214
Po-214
Ac-228
Fr-223
Th-227
Ra-223
Rn-219
Po-215
Pb-211
Bi-211
Tl-207
Po-211
Ra-224
Rn-220
Po-216
Pb-212
Bi-212
Tl-208
Po-212
Branching
Ratio ;
100%
100%
• 1.4%
22.8% '
94.6%
1.43%
98.6%
100%
100%
100%
100%
99.98%
100%
99.97%
100%
1.38%
98.6%
100%
100%
100%
100%
100%
99.7%
0.27%
100%
100%
100%
100%
100%
35.9%
64.1%
Half-Life
64.1 h
29.9s
24.6s
58 d
2.52m
7.2m
17.3m
5.01 d
138 d
3.82 d
3.05m
26.8m
19.9m
164 iis
6.13 h
21.8m
18.7 d
11.4d
3.96 d
1.78 ms
36.1m
2.13m
4.77m
0.52s
3.62 d
55.6s
0.15s
10.6 h
60.6m
3.05m
0.30 \is
Radiation
P
P,Y
P,Y
Y,e-
Y,e-
Y,e-
'P,Y
P
a
«,Y
a
P,Y,e-
P,Y,e-
«,Y
P,Y,e-
P,Y,e-
«,Y.e-
a,Y,e-
asy,e-
«,Y
P,Y,e-
a,Y,e-
P,Y
a,v
«,Y,e-
a,Y
a
P,Y,e-
a,P,Y3e-
P,Y,e-
a.
a Only progenies with half-lives of six months or less are included in the implicit progeny of "+D" nuclides.
                                                 6-3

-------
                                  Table 6-1 (continued)
Parent "•- ' -««'
Nuclide
Th-229
U-235
U-238
Np-237
Pu-241
Half-life
7340 y
7.06e+08 y
4.47e+09 y
2.14e+06y
14.4v
Radiation.
e-
a,e-
a,y,e-
p
! ^' /Progeny'' - ",
<3«ttefatJon. j
l
2
3
4
5
6
6
7
1
1
2
3
1
1
NticHde
Ra-225
Ac-225
Fr-221
At-217
Bi-213
Tl-209
Po-213
Pb-209
Th-231
Th-234
Pa-234m
Pa-234
Pa-233
Am-241
BraacMag
Ratio
100%
100%
100%
100%
100%
2.16%
97.8%
100%
100%
100%
100%
0.16%
100%
100%
Half-Life
14.8 d
10.0 d
4.8m
32.3 ms
45.7m
2.20m
4.20 us
3.25 h
25.5 h
24.1 d
1.17m
6.70 h
27.0 d
432 v
Radiatioti
P,Y,e-
cc,Y,e-
a,Y,e-
«,Y
a,P,Y,e-
P,Y,e-
a
.P
P,Y,e-
P>Y,e-
P,Y,e-
P,Y,e-
P,Y,e-
a.y.e"
Because of the variabilities of contamination patterns and storage conditions, it cannot be
assumed that radon isotopes would escape from the surface of the metal.  The contamination
might have been painted over, for instance, or trapped inside a steel component that was crushed
as part of a volume reduction process. Therefore, for example, the assessment of Ra-226+D
assumes that Rn-222 and its short-lived daughter products would remain in the scrap in complete
secular equilibrium with the radium, while that of natural uranium series assumes that both Rn-
222 and Rn-219, as well as then" entire progenies, would be hi secular equilibrium with U-238
and U-235, respectively. Similarly, the assessments of the Th-232 series  and of Th-228+D
assume that Rn-220 would remain in the scrap.

Except for the "+D" nuclides with then- short-lived progeny and the natural uranium and thorium
series,  no ingrowth of progenies was modeled in the radiological assessment of individuals.
Exposures of scrap processors, steel mill workers and slag users are assumed to occur within a
few months of the release of the scrap for recycling, too short a time for any  significant ingrowth
of long-lived progeny. Since the finished products that were among the subjects of the analysis
have useful lives of several years, such ingrowth could potentially occur.  However, as will be
seen in the discussion of such exposure scenarios later in this chapter, such ingrowth would have
no significant impact for the nuclides and the materials considered hi the  analysis.
                                          6-4

-------
           Table 6-2. Nuclides Included in Various Combinations and Decay Series
Series
U-Natural
U-Separated
U-Depleted
Th-Series
Nticlide
U-238+D
U-234
Th-230
Ra-226+D
Pb-210+D
U-235+D
Pa-231
Ac-227+D
U-238+D
U-234
U-235+D
U-238+D
U-234
U-235+D
Th-232
Ra-228+D
Th-228+D
Activity
Fraction
i.
i.
i.
i.
i.
0.047
0.047
0.047
1.
1.
0.047
1.
0.09123
0.01594
1.
1.
1.
6.2    SPECIFIC ACTIVITIES OF VARIOUS MATERIALS
When steel scrap is charged to an electric arc furnace (EAF), chemical agents (fluxes) are added
to control the chemical properties of the molten metal.-.The interactions among the flux, the
refractory brick which lines the furnace, and the molten metal affect the final composition of the
melt and hence the distribution of radionuclides among the several furnace products, such as the
melt, the slag and the off-gas. The melt is subsequently allowed to cool and becomes the
primary output of the mill. Slag is the material not remaining in the metal and includes the
chemical agents,  some of the liner material, and small amounts of the base metal, much of which
is recovered and charged to the furnace for a subsequent melt. Off-gas consists of the fumes and
aerosols evolved  during melting which are captured by the facility's emission control system and,
after cooling, collected in the baghouse in the-form of dust.
             exposure assessment of a given contaminant in the scrap, it is necessary to
             ' that material is distributed in the various media products following the melting of
To perform an exposure assessment of a given contamin
determine how that material is distributed in the various
the scrap.
                                         6-5

-------
The concentration of radionuclide / in medium m is calculated as follows:
                                         CMP
                                  C   = - — i
                                           M
                                             m
      C,m      =     specific activity of radionuclide z in medium m(pCi/g)
      Cfc      =     specific activity of radionuclide i in the scrap (pCi/g)
      MJ      =     mass of radioactive scrap in furnace charge (g)
      Pim      =     partition ratio or vaporization fraction of radionuclide z in medium m
              =     mass of medium m produced from that charge (g)
A literature search as well as thermodynamic calculations were used to develop the partition
ratios and vaporization fractions for EAF melting of carbon steel used in the present analysis
which are listed in Table 6-3. Ranges of partition ratios reflect variability in melting practices.
A detailed report of this study appears in Appendix E.  A similar study was performed for cast
iron production; it is reported in Appendix F.

To calculate the concentration factor, CF (i.e., Cim/Cls), it is also necessary to determine the mass
of each medium as a fraction of the mass of the furnace charge. Based largely on the discussion
in Section E.7 of Appendix E and a comparable discussion hi Appendix F, the following mass
fraction values were adopted for the present analysis:

      •      Imported scrap (scrap from sources outside the mill): 0.95
      •      Home scrap (metal recovered from by-products of previous melts): 0.05
             Finished steel: 0.9
             Steel slag: 0.117                 *
      •      Cast iron slag: 0.065
      •      Baghouse dust: 0.015
             Melt:  0.97 (see below)
                                          6-6

-------
          Table 6-3.  Partition Ratios (PR) and Concentration Factors (CF)a
BretneiJiE
. Ac
Ag
Am
C
Ce
Cm
Co
Cs
Eu
Fe
I
Mn
Mo
Nb
Ni
Np
Pa
Pb
Pm
Po
Pu
Ra
Ru
Sb
Sr
Tc
Th
U
Zn
Imported :
Scrap i
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
1
Famaee
Cbaige* ,
0.95
1
0.95
1
0.95
0.95
1
0.95
0.95
1
0.95
0.98
1
0.95
1
0.95
0.95
0.95
0.95
0.95
0.95
0.95
1
1
0.95
1
0.95
0.95
0.96
s Steel
JPR
(%)
0
99/75

100/27


99


97

24/65
99

99







99
99/80

99


20/0
CF
0
1.02
0
1.03
0
0
1.02
0
0
1
0
0.67
1.02
0
1.02
0
0
0
0
0
0
0
1.02
1.02
0
1.02
0
0
0.2
Cast
Iron
CF'
0
1.01
0
1.01
0
0
1.01
0
0
1
0
0.98
1.01
0
1.01
0
0
0
0
0
0
0
1.0-1-
1.01
0
1.01
0
0
0.02
Slag
PR-
(%)
95

95

95
95

0/5
95
2

72/32

95

95
95

95

95
95


95

95
95

CF
7.79
0
7.79
0
7.79
7.79
0
0.41
7.79
0.19
0
6.15
0
7.79
0
7.79
7.79
0
7.79
0
7.79
7.79
0
0
7.79
0
7.79
7.79
0
Baghouse
PR
(%)
5
1/25
5 "

5
5
1
100/95
5
1

4/3
1
5
1
5
5
100
5
100
5
5
1
1/20
5
1
5
5
80/100
CF
2.6
16.5
2.6
0
2.6
2.6
0.67
63.3
2.6
0.67
0
2.21
0.67
2.6
0.67
2.6
2.6
63.3
2.6
63.3
2.6
2.6
0.67
13.2
2.6
0.67
2.6
2.6
63.3
Volatile
/ PR
<&>

"

0/73






100







,










a        Data is relevant to EAF operations, except cast iron concentration factors, which apply to iron foundries.
b        Refers to the scrap metal charged to the furnace, which consists of 95% imported scrap and 5% recirculating home scrap.
                                             6-7

-------
The iron content of slag can be calculated from the average chemical composition of steel slags,
which is listed in Appendix I. Comparing these data with the composition of baghouse dust,
presented in Appendix E-2, allows us to infer the source of the dust. We thus conclude that:

              Fe content of slag = 19.4 %
       •      Source of baghouse dust:  Vs slag, % steel

In deriving the concentration factors, the home scrap was assumed to have the same
concentration as the melt. In all cases where the partition ratio (PR) for slag was listed as 95, it is
assumed that all the activity first partitions to the slag, which initially has a mass fraction of
0.122, of which 0.005 forms the baghouse dust. The above calculations assumed that all dust is
retained by the baghouse filters—the release of filtered particulates to the atmospheres are not
included in the volatilization factors.

In all cases where a range of partition ratios is listed for a given chemical element in a given
medium, the high end of the range is used to calculate the corresponding concentration factor.
The range results from the variability of melting practices and other factors. Consequently, a
given individual may be exposed to radionuclide concentrations corresponding to the high end of
the range in one medium, while a different individual could be exposed to the high end of the
range for a different medium.  In only one scenario in the present analysis—the operator of the
continuous caster—is the same individual exposed to radioactivity from two different media
(other than scrap).  As shown hi Table 5-1, this individual would be exposed to  external radiation
from the steel while inhaling and ingesting the  furnace emissions (i.e., baghouse dust). The
radiological impacts on this individual of those nuclides that have a range of partition ratios hi
both the steel and the dust—isotopes of silver, manganese, antimony and zinc—are overstated.
(Rather than calculating the partition ratios that would result in the maximum exposures in these
few cases, these conservative estimates were retained in the analysis.) In all other cases,
however, this approach yields a reasonable, maximum exposure assessment.

6.3    EXPOSURE PATHWAYS

6.3.1  External Exposures to Direct Penetrating Radiation

Table 5-1 shows that the external exposure pathway is included in every scenario except the
consumption of ground water contaminated by leachate from a slag pile.  Except for the
assessment of exposure to airborne effluent emissions, which are discussed later hi this chapter,

                                           6-8

-------
external exposures were evaluated either by using the MicroShield™ computer code or by
employing the external exposure dose coefficients hi Federal Guidance Report (FOR) No. 12
(Eckerman 93).                f

Use of MicroShield™ Computer Code

MicroShield™ (Grave 95) is an industry-standard computer program used to perform y-ray
shielding calculations for radioactive sources. Results obtained with MicroShield™ are generally
in good agreement with those performed by photon transport codes employing discrete ordinate
or Monte Carlo methods. At photon energies below about 100 keV, the MicroShield™ results
begin to diverge from those calculated by the more elaborate methods. This limitation, however,
is not of concern in the present analysis.  The primary dose contribution from most of the y-
emitting nuclides is from the high-energy photons. From nuclides that emit only low-energy
photons, the dose is dominated by internal exposure.  In neither case do the low-energy photons
make a significant contribution to the total dose.

MicroShield™ utilizes dose coefficients listed in ICRP Publication 51 (ICRP 87) to calculate the
effective dose equivalent for each of five exposure geometries. For most exposure scenarios, the
present analysis assumes that the radiation is incident in the anterior-posterior direction, which
corresponds to the exposed individual's facing the radiation source. This is a realistic assumption
hi most cases—it also results hi the highest dose. The resulting dose conversion factors (DCFs),
expressed hi millirem per hour, are utilized in the assessments of the external exposure pathways.
An illustrated description of the source and receptor configuration for each scenario analyzed
with Microshield™ are presented hi Appendix H.

The DCFs are used to calculate the normalized doses and risks from external exposure.  The
source-to-receptor distance and the duration of exposure for each scenario are listed hi Table 5-1.
Additional details are presented hi Appendix H. The annual dose to the maximally exposed
individual from a given nuclide hi a given scenario is calculated by multiplying the appropriate
DCF by the exposure duration and by the specific activity hi the source medium, normalized to a
unit specific activity in the scrap.  The concentration factors for the various nuclides in the
different media are listed in Table 6-3. These calculations are shown in Equation 6-2, below.
                                          6-9

-------
Djn*(x)       =     dose from one year of external exposure to radionuclide / in medium m at
                    distance x (mrem/y per pCi/g in scrap)
x             =     distance from source to receptor (m)
t,.             =     annual exposure duration (hr/y)
Fjmx(x)        =     DCF from external exposure at distance x from radionuclide I in medium
                    m in a given source configuration (mrem-g/pCi-hr)

External Exposure over a Varying Distance

In several scenarios, such as the EAF furnace operator, the distance between the source of the
                                                                     i
external radiation and the exposed individual varies over time — Le., the operator is at different
locations during the course of the day. Although the rninimum and maximum distances of a
given individual have been observed or can be inferred, the duration of his or her occupancy of
the various locations within this range is difficult to ascertain. The analysis therefore makes the
simplifying assumption that the individual spends an equal amount of time at each distance. This
is equivalent to assuming that he or she moves uniformly back and forth, like a sentry walking
his post between two points.

To determine the integrated exposure during this time, it is necessary to derive the exposure rate
at some arbitrary distance from the source, given the exposure rates at two fixed distances. To do
this, we first calculate the distance aad strength of a fictitious equivalent point source that would
produce the same exposure rates at the same locations as those calculated for the real source.
Applying the inverse square law, we obtain:
                                        (* -xf
     R(x)          =     exposure rate at distance x from real source
     AO            =     strength of equivalent point source
     x,,             =     distance of equivalent point source from real source

To evaluate the constants A,, and x0, we substitute the calculated values of R(x) at two known
distances:
                                          6-10

-------
                                                                                  (6-4)
                                           A
                                             o
Solving Equations 6-4, we obtain:
                                                                                 (6-5)
Next, we find the mean value of R(x) over the interval, x3 < x < X4
                                             *
                                         (x - xo)2
                               R = __^	                              (6_6)
                                        X4-
Equation (6-6) is used to evaluate the factor Fimx(x) in Equation 6-2 over the range [x3, x4].

Use of FGR 12 Dose Coefficients

MicroSMeld™ is a useful tool for determining dose rates from relatively compact sources. In'
some scenarios, however, the external radiation comes from a planar source whose lateral
dimensions are large in comparison to the source-to-receptor distance, and which has a mass
thickness many times greater than the mean free path of the most penetrating radiation of any of
the nuclides in the analysis. In those cases, the dose coefficients for soil contaminated to an
infinite thickness listed in FGR 12 provide a convenient method of analysis.
                                          6-11

-------
These factors were applied to the slag yard worker standing at the edge of the slag. Since he is
only exposed to one-half of an infinite plane, he would only get half the dose predicted by FGR
12.  Since the average atomic number of slag is somewhat higher than that of soil, the analysis
would tend to overstate the doses.  For the nuclides with the most energetic y-rays, for which
external exposure is a major pathway, the interaction of the radiation with the source material is
primarily by Compton scattering, which is relatively insensitive to the atomic number.

The FGR 12 dose coefficients were also used to evaluate the external exposure of the scrap
cutter.  Since he spends time in alleys surrounded by walls  of scrap, it is reasonable to model the
sources as two vertical half-planes beginning at the ground surface.  The two half-planes together
are equivalent to a single infinite plane.  Again, the scrap has a higher atomic number than the
average for soil, yielding a somewhat conservative but not excessively overstated assessment.

6.3.2  Inhalation of Contaminated Dust

During certain of the operations listed in Table 5-1, some of the radioactively contaminated
material is assumed to be dispersed in the ambient atmosphere in the form of dust.  The radiation
exposure of an individual inhaling  this dust will depend on his breathing rate, the dust loading of
the ambient air, the respirable fraction (i.e., the mass fraction of particles with AMAD^ 10 um)2,
the exposure duration and on whether or not he uses some form of respiratory protection.  The
radiological impacts are modeled by the following equations:
                                                                                   (6-7)
                                   =B C
      Dimh     =      70-year dose commitment from inhalation of radionuclide / in medium m
                     during one year (mrem/y EDE per pCi/g in scrap)
      B       =      breathing rate
                     1.2(m3/hr)
      ff       =      respiratory protection factor (filter factor, dimensionless)
   2 AMAD is the acronym for Activity Median Aerodynamic Diameter, "[which] is the diameter of a unit density
sphere with the same terminal settling velocity in air as that of an aerosol particle whose activity is the median of the
entire aerosol." (Eckerman 88).

                                           6-12

-------
     fj       =     respiiable fraction
     Flh      =     DCF for inhalation of radionuclide 1 ( mrem/pCi—FOR 11)
     %d      —     concentration of dust in air (dust loading, g/m3).
     Rjmh     =     excess lifetime risk of radiogenic cancer from inhalation of radionuclide /
                    in medium m during one year (y"1 per pCi/g in scrap)
     Gih      -     risk factor for inhalation of radionuclide I (pCi"1—EPA 94a)

The dust loading for each exposure scenario is listed in Table 5-1, A discussion of the derivation
of these values appears in Appendix H.

The analysis assumes that all of the airborne dust emanates from the contaminated material being
processed and that the specific activity of a given radionuclide hi the dust is the same as that of
the material.  This assumption is realistic for operations such as handling of baghquse dust or
slag, or the use of a cutting torch on scrap. In these cases, the dust results from the operations
and would contain the same radionuclides as the material in process.

Studies also show that the specific activity in the dust may be either greater or less than the
radionuclide concentrations in the source of the dust due to enhancement and discrimination
processes. For example, the particles that become airborne are usually less than 50 urn in
diameter (Peterson 83). If most of the radioactivity in4he material that is the source of the dust is
found in particles larger than 50 um, then activities hi the dust are likely to be lower than that in
the source of the dust. Conversely, if the activity is primarily on particles smaller than 50 urn,
the specific activity in the dust can be greater than that in the source of the dust The assumption
that the specific activity of a given radionuclide in the dust is the same as that of the material is a
reasonable approximation in most cases. Many other physical and chemical properties besides
particle size can also produce enhancement or discrimination effects.  A discussion of this
subject is provided in Envirosphere 84.

The inhalation DCFs are based on Federal Guidance Report (FOR) No. 11 (Eckerman 88), while
the inhalation risk factors (i.e., slope factors) are based on EPA 94a. These dose and risk factors
are based on the assumption that the airborne particles have an AMAD of 1 urn, which results in
the particles penetrating deeply into the lungs and creating the greatest potential for exposure.
The larger particles that comprise the major part of the dust by mass do not penetrate as deeply
                                          6-13

-------
into the lung and thus have a smaller radiological impact on the exposed individuals3.
Accordingly, the dose and risk factors used in this study are conservative, upper-end values.

The dose factors also depend on the chemical form of the nuclide in question. Since the
chemical form of the radionuclide on the scrap metal is unknown, it is assumed that a nuclide on
or in the scrap has the form corresponding to the highest DCF listed in FOR 11. Similarly, it is
assumed that the nuclides in the ground water will be in their most soluble form. In the case of
baghouse dust, slag dust or vapors from molten metal, the analysis of radionuclide distributions
during the melting of carbon steel indicates that hi almost all cases the nuclides will be present as
oxides. Consequently, the dose conversion factors corresponding to the respective lung
clearance class and fj value, as listed in Table 3 of FOR 11, are adopted for the analysis. For the
ingestion of radionuclides left in the melt—the frying pan scenario in the present analysis—the
nuclides are assumed to be hi the elemental form.

The chemical form of each element with radioactive isotopes that may be found in radioactively
contaminated carbon steel is listed in Table 6-4, along with the appropriate lung clearance class
and f, value.

6.3.3   Incidental Ingestion

Individuals  working in a dusty, sooty environment are likely to inadvertently ingest some of the
contaminated material, which is genetically referred to hi this report as soot. The radiological
Impacts of such incidental ingestion will depend on the soot ingestion rate of the exposed
individual and on the duration of exposure. The  impacts are modeled by the following
equations:

                                  D,-mg  = c/m^'e
                                                                                     (6-8)
                                  Rimg = C^G^I.t.
   3 An additional contribution to the dose from dust inhalation is from large particles that are inhaled, refluxed from
the air passages and then swallowed. In all scenarios where inhalation exposure is modeled, inadvertent ingestion is also
assumed. Since the inadvertent ingestion rate of particulate material is typically several times larger than the inhalation
rate of the same material, the ingestion of these large inhaled particles would not have a significant impact on the total
dose.

                                            6-14

-------
     D.
       img
      img
     G,-0
70-year dose commitment from ingestion of radionuclide /in medium m
during one year (mrem/y EDE per pCi/g in scrap)
DCF for ingestion of radionuclide / ( mrem/pCi)
soot ingestion rate (g/hr)
excess lifetime risk of radiogenic cancer from ingestion of radionuclide /
in medium m during one year (y1 per pCi/g in scrap)
risk factor for ingestion of radionuclide /(pCi"1)
The EPA Exposure Factors Handbook (EPA 89) presents a detailed discussion of soil and soot
ingestion, primarily by children. However, data are also provided for inadvertent soil and soot
ingestion rates by adults working in a dusty environment. For adults, the daily soil ingestion
rates range from 0.56 mg/day for indoor work to 480 mg/d for outdoor work. Given the nature of
the operations at scrap yards and steel mill, a mid-range value of 240 mg/d was assumed.  Since
the EPA values  are assumed to refer to an 8-hour workday, a soot ingestion rate of 30 mg/hr was
adopted for modeling the inadvertent ingestion pathway. The one exception is the lathe
manufacturing operation, where it is likely that only part of the "dirt" in the area would come
from the cast iron that is being ground.  Some would come from the grinder itself, for instance.
Consequently, an ingestion rate of 10 mg/hr was adopted for that operation.

             Table 6-4. Lung Clearance Class and F! Values for Use with FGR 11
Element- 1
Ac
Am
Ag
C
Ce
Cm
Co
Cs
Eu
Fe
I
Mn
Mo
Mb
Ni
' Melt
Forift.


Ag
C


Co


Fe

Mn
Mo

Ni
Class


D



w


w

D
D

D
ft '


0.05
1


0.3


0.1

0.1
0.8

0.05
Dust & Slag
Banft
AcA
Am2O3
Ag
CO2
Ce^C-3
all
CoO
all
all
FeO
all
MnO
MoO3
Nb2O5
NiO
- Class ,
Y
W
*-D

Y
W
Y
D
W
W
D
W
Y
Y
W
f»
E-3
E-3
0.05
. 1
3E-4
E-3
0.05
1
E-3
0.1
1
0.1
0.8
0.01
0.05
Scrap &<3W
£<«m i
max
all
max
organic
max
all
max
all
all
max
all
max
max
max
max
* Class '
D
W
Y

Y
W
W
D
W
D
D
W
. Y
Y
D
•fi
E-3
E-3
0.05
1
3E-4
E-3
0.3
1
E-3
0.1
1
0.1
0.8
0:01
0.05
                                         6-15

-------
        Table 6-4. Lung Clearance Class and F! Values for Use with FOR 11 (Continued)
Element
Np
Pa
Pb
Pm
Po
Pu
Ra
Ru
Sb
Sr
Tc
Th
U
Zn
Melt
Form







Ru
Sb

Tc


all
Class
W






D
D

W


Y
ft







0.05
0.01

0.8


0.5
Dust&Slag
- Fojtra
all
PaO2
all
Pm2O3
PoO2
Pu203
all
RuO4
Sb2O3
SrO
TcO2
ThO2
U02
all
Class
W
Y
D
Y
W
Y
W
Y
W
D
W
Y
Y
Y
fi
E-3
E-3
0.2
3E-4
0.1
E-5
0.2
0.05
0.01
0.3
0.8
2E-4
2E-3
0.5
Jcrap&OW
form
all
231
233/4
all
max
max
max
all
max
max
a
max
231/4
others
max
all
.Class
W
W
Y
D
Y
D
W
W
Y
W -
D
W
Y
W
Y
Y
f,<
E-3
E-3
0.2
3E-4
0.1
E-3
0.2
0.05
0.01
0.3
0.8
2E-4
0.05
0.5
       * All except SrTiO}, an unlikely contaminant of potentially contaminated steel scrap
6.3.4  Radioactive Decay
Sections 6.3.1—6.3.3 present methods of calculating dose rates for all scenarios in which the
source strength is essentially constant during the course of one year. These are the scenarios in
which the source is replaced at frequent intervals. For scenarios in which the source is not
replaced—the end user of finished products scenarios— radioactive decay over the course of a
year must be taken into account. In such cases, the exposure is integrated over a period of one
year, resulting in the following expression:
                                                       -V
                                                    - e   r
                                                                                    (6-9)
                     radioactive decay constant of nuclide /(y1)
                     integration time
                     iy
                                           6-16

-------
The risk from external exposure Is calculated by multiplying the corresponding dose by the risk
factor for doses of low-LET radiation to the whole body:
                               **,M = *>„&)**                             (6-10)
          x)  =     excess lifetime risk of radiogenic cancer from one year of external
                    exposure to radionuclide I In medium m at distance x (y1 per pCi/g in
                    scrap)
     J?RX     =     risk factor for external exposure
                    7.6x10'7 mrenr1 (EPA 94a)

In addition to decay, ingrowth of progeny was also considered. As stated earlier, all progenies
with half-lives of less than six months are included in the exposure assessment of the parent.
Eleven of the elements listed in Table 6-3 significantly partition to steel or iron. Mo-93 is the
only one of isotopes of these elements that were included hi the analysis that has a long-lived
progeny, i.e., Nb-93m, which has a half-life of 16.1 y. The longest lived product is the industrial
lathe, with a useful life of 20 years. Thus, even in the last year of anticipated use, the Nb-93m
activity in the lathe would be less than 60% than that of Mp-93, which has a half-life of 3,500 y
and would thus not decay significantly.  The external dose from the Nb-93m approximately one-
tenth of that from Mo-93, the inhalation dose would be about one-fourth and the ingestion about
one-half. Given the other uncertainties in the analysis, omitting the Nb-93m contribution to the
total Mo-93 dose in the finished product scenarios does not have a significant impact, especially
since, as will be seen later, the individuals exposed to finished products potentially contaminated
with Mo-93 receive a much smaller dose than the RMEI, selected from all scenarios in which
Mo-93 is a potential contaminant.

6.4    UNIQUE SCENARIOS

The exposure assessment of two of the scenarios developed for the analysis required the
development of special sub-models.  These exposure pathways, the consumption of ground water
contaminated by leachate from slag and the consumption of food cooked in a cast iron cookware
made from potentially contaminated scrap, are described hi the following sections, as is the
impact of fugitive airborne emissions from the furnace on nearby residents. A qualitative
discussion of the anticipated impacts of disposition of the recycled scrap in a sanitary landfill is
presented at the end of Chapter 6.

                                          6-17

-------
6.4.1  Ground Water Contaminated by Leachate from Slag Storage Piles

As discussed in Section 5.2.2, an individual residing near the slag storage yard, who gets his
drinking water from a well that is down-gradient from the slag could be exposed to contaminated
ground water. During storage at the steel mill site, slag will be subjected to weathering and
certain components may be leached from the slag and ultimately contaminate the local ground
water. A study, based on a search of available literature, was performed to enable the calculation
of leach rates of various radionuclides. Details of this study are presented in Appendix I. In
addition, EPA is sponsoring an experimental study being conducted at the Brookhaven National
Laboratory to determine the leach rates of constituents of various steel and iron slags.  Some
preliminary results of the Brookhaven study are presented hi Appendix 1-2. This section presents
some of the information obtained from both studies and then develops an interim model of the
leaching of radionuclides which partition to the slag.

Elemental Selection Criteria

Because of the scarcity of data, it was desirable to narrow the scope of the analysis to those
elements that have radioactive isotopes which, if leached from slag produced by the melting of
potentially contaminated steel scrap, would have a significant radiological impact via the ground
water exposure pathway.  The selection criteria include:

       •       potential contamination of steel scrap by one or more isotopes the given element
       •       significant partitioning of the given element to the slag (/. e., concentration factor
              *0.1)
       •       travel time to the aquifer relative td*half-life of longest-lived isotope being
              studied
       •       travel time relative to the 1,000 year period of impact assessment

Travel time. The travel time of the /-th nuclide was determined by Equation E.21 of the
RESRAD user's manual (Yu 93).
                                          6-18

-------
                                               '
                                                   P*K*
                                           =  1  +	
                                                   pt*.
                                             f   \   1
                                                   26+3
      Az      =      thickness of unsaturated zone
              =      4m
      R.      =      retardation factor for 7-th nuclide
       d,
      pe       =      effective porosity of unsaturated zone
                     0.2
      I        =      infiltration rate
                     0.5 ra/y
      pb       =      bulk soil density
              =      1.5g/cm3
      Kd      =      distribution coefficient for /-th nuclide (cm3/g)
      pt       =      total porosity
                     0.485
      KV      =   '   satoated hydraulic conductivity of vadose zone
                     227 ra/y                    V
      b       =      soil-specific exponential parameter
                     5.3
The values listed above, as well as the Kj's for soil, are those used for the generic site analysis in
the TSD for the soil cleanup levels (EPA 94b).
                                           6-19

-------
                      Table 6-5. Potential Contaminants of Groundwater
Element
Ac
Am
Ce
Cm
Cs
Eu
.Fe
Mn
Nb
Np
Pa
Pb
Pm
Pu
Ra
Sr
Th
U
Slag CF.
7.79
7.79
7.79
7.79
0.41
7.79
0.19
6.15
7.79
7.79
7.79
progeny
7.79
7.79
7.79
7.79
7.79
7.79
t,/"'
(y) '
21.8
432.7
0.8
18.11
30.17
13.6
2.7
0.9
20,300
2.14E6
32,760
22.26
2.6
24,131
1600
28.6
1.4E10
4.47E9
?* '
240
1900
500
4000
270
240
170
50
110
5
110
270
240
550
500
15
3200
15
- Atf
' 00
1188.30
9400.32
2474.52
19789.02
1336.71
1188.30
842.01
248.37
545.19
25.76
545.19
1336.71
1188.30
2721.87
2474.52
75.23
15831.42
75.23
Potential
X
X
X'
X
X
X
X
X
•
•
•

X
X
X
•
X
•
Comments
will decay
will decay
will decay
will decay
will decay
will decay
will decay
will decay




will decay
Lt,» l,000y
Af,»l,000y

At,» 1,000 y

              a Half-life of longest lived isotope
                                                                t
Elements with travel times longer than 20 half-lives4 of their longest-lived isotopes or with travel
times much greater than 1,000 years are marked with an "x" in Table 6-5, indicating that they are
eliminated from consideration—it is not likely that any significant radioactivity would reach the
aquifer during the 1,000-year assessment period under any reasonable environmental conditions.
    4 Twenty half-lives was selected as the cutoff criterion since that represents a decay to lO"6 of the initial activity.

                                            6-20

-------
Slag Cement Leaching Studies

The American Nuclear Society has developed and formalized detailed procedures for measuring
the teachability of solidified low-level radioactive wastes (ANS 86).  This procedure involves
testing of controlled geometry specimens in demineralized water at 17,5°C to 27.5°C to
determine releases over successive intervals of time.  Mass transport is assumed to be controlled
by a diffusion process. When the fraction leached from a uniform sample is less than 20% of the
initial activity, the leaching behavior can be approximated by that of a semi-infinite medium
where the "effective diffusivity" is given by the following equation:
                              D, =
                                           a._V
                                            in
(6-11)
     D,      =     effective diffusivity of nuclide / (cmVs)
     Tn      —     mean time of the leaching interval n (d)
                    I    2    )
     a,n      =     activity of nuclide /released during time interval n (g)
     V      =     sample volume (cm3)
     A,0     =     initial activity of nuclide I in sample
     Ant     =     duration of n-th leaching interval (d)
             =     V" tn-l
     S      =     surface area of sample (cm2)

                                      a.
When the cumulative fraction leached, S ——, is greater than 0.2, Equation 6-11 must be
corrected for specimen geometry.     n   lo
Using a model and procedures similar to those described in ANS 86, Japanese investigators have
determined the fractional leaching of Sr-90, Co-60, Cs-137 and H-3 from cement/slag
composites in deionized water and synthetic sea water (Matsuzuru 77,77a, 79),  The duration of
the leaching tests was about 100 days. The radionuclides were incorporated into the cement via a
sodium sulfate solution. The composition of the slag is listed in Table 6-6.
                                         6-21

-------
Leaching data were analyzed using a plane source diffusion model to derive the expression
                                 f,=
                                      2S_
                                       v
(6-12)
  = j&action of nuclide /leached in t days.
Equation 6-12 can be rewritten as
                               f,
                                     2S
                                      V  \
                                 = mtt'
                                                                                 (6-13)
where we have represented the expression in the square brackets by mi5 the slope of the line
obtained by plotting § vs. t'/J. Once m.j is determined, Equations 6-13 can be solved for D;:
                                     1C
                                         IS,
                                                                                (6-14)
Since the actual leaching process involves an initial rapid leaching rate of a few days' duration
(~ 7 d for Sr-90 and ~ 2 d for Co-60), followed by a longer term linear relation between fj and t
the experimental data are fitted to an equation of the form

                                 /,-«,**  +«r
                                                                                 (6-15)
Because of certain limitations and problems such as the initial leach rate, Matsuzuru et al.
defined L, the leaching coefficient, with the same mathematical form as D hi Equation 6-14.
                                          6-22

-------
                   Table 6-6. Composition of Slag Used in Leaching Test
Component .
Si07
A1203
FeA
CaO
MgO
Insoluble Residue
Ignition Loss
Composition
' (Wt%) '
28.7
' 11.5
2.3
50.9
3.2
0.8
0.6
Equation 6-12 can also be used to determine the values off; for various geometries, as follows:
                                f,2  = f,l
(6-16)
where subscript 1 refers to geometry 1 while subscript 2 refers to geometry 2.

Values of L for Sr-90 leached from slag cements ranged from 1.2 x 10'7 to 1.7 x 10'7 cm2/day for
both deionized water and synthetic sea water at 25 °C (Matsuzuru 77a). Using average values of
LSr for samples cured seven days prior to testing in deionized water, and assuming a right circular
cylinder, D = h, V = 70 cm3 (r = 2.233 cm), we have derived values for mSr and aSr in Equation 6-
15, which are listed hi Table 6-7. The leachability of Cs-137 was reported to be about ten times
that of Sr-90; therefore we assumed that m^ =10 mSr and aCs = 10 aSr. We then used Equation 6-
16 to derive values that describe leaching from slag particles that are also right circular cylinders,
but only 1 cm in diameter—a more representative size for EAF slags.
                                          6-23

-------
                          Table 6-7. Leaching Parameters Values
Element
Sr
Cs
Cr*
m(&%)
r = 2.233 cm
5.8E-4
5.8E-3
7E-6
r = {K5em
2.59E-3
2.59E-2
3E-5
a '
r.= 2.233 cm
4.97E-3
4.97E-2
0
r = 0.5cni
2.22E-2
0.222
0
       1 Cr is used as a surrogate for Nb, Np, Pa and U—see discussion below.
Use of the data obtained from slag cement leaching studies is believed to be conservative since
the radionuclides in the cement composites are not dissolved in the slag and therefore not
expected to be as tightly bound in the solid matrix. The strontium data were replaced by the data
obtained from the Brookhaven National Laboratory, as described below.

Preliminary Data from Brookhaven National Laboratory

Preliminary results from leaching experiments on EAF slags performed at the Brookhaven
National Laboratory indicate that the leaching of strontium, the only element checked in Table 6-
5 that was measured in the leachate, was governed by diffusion (Fuhrmann 97). The diffusion
coefficients determined from tests on three monolithic samples of EAF slag are listed in Table 6-
8.  We  calculate the value of mSr for a monolithic cylinder beginning with an inversion of
Equation 6-14:
2S
V \
Dt _
ft
S = 2lt(r2 +
V = icr3
1.99xlO'2d-K
12 Dt
r 'Sj 7C
hr) = 6

                                                          (A =2r)
                    2.5 x ID'11 cm2/s
                    2.16xlO-6cm2/d
                    0.5cm
                                         6-24

-------
             Table 6-8. Diffusion Coefficients for EAF Slag Monolithic Samples"
Slag Sample
AS-1
AS-2
AS-3
Diffasiott'Coefftcient (cmYs)
1.4E-11
2.5E-11
6.2E-12
                           from Fuhrmann 97
Since Fuhrmann did not report any initial releases that were not diffusion-controlled, aSr is
assumed to equal zero.  These new data were used to model the Sr-90 leaching from the slag hi
the present analysis.
                                                          s
Other Slag Leaching Studies

This section describes earlier leaching studies done on pure slags rather than slag/cement
composites.

Australian researchers at CSIRO incorporated the toxic elements As, Sb, Cd, Zn and Cr into
various types of slags by melting at 1300 °C, and subsequently leached the slags according to the
EPA TCLP protocol (Jahanshahi 94).  In the TCLP test, a sample of at least 100 g, which has a
minimum surface area of 3.1 cm2/g or passes through a 9.5 mm sieve, is treated with about
2,000 g of extractant for 18 ± 2 hr at 22 ± 3 °C using rotary agitation.  The extractant has a pH of
either 4.93 or 2.88, depending on the basicity of the sample (40 CFR 261, Appendix II, Method
1311). The pH is achieved by use of acetic acid which is buffered with sodium acetate for the
higher pH level (55 FR 11798).

Slag samples were  prepared by both slow cooling and quenching.  Examination of the slag
samples with an optical microscope showed that interconnecting porosity was present in the slow
cooled and most of the quenched samples.  Slow-cooled slag samples were crushed to either a
"coarse" size (100% minus 10 mm)5 or a "fine" size (100% rninus  1 mm) for the leaching tests.
In generalizing on the results of the TCLP tests, the researchers observed that:
   5  "100% minus 10 mm" means that 100% of the particles passed through a screen with a 10 mm mesh.

                                          6-25

-------
              As and Sb leached more readily than Cd, Cr or Zn
       •      Fine particles generally leached more readily than coarse particles
       •      Slow cooled samples showed similar behavior to quenched samples

Based on the information presented in Jahanshahi 94, we estimated the fraction leached using the
following assumptions:

       •      Slag compositions from Table III of Jahanshahi 94
       •      Sample size = 100 g
       •      Extractant volume — 2 L

The results are presented in Table 6-9. For three of the slags (CaFel, CaFeSil, and FeSil), the
compositions are markedly dissimilar to those expected from EAF melting of carbon steel.  The
other three slags, while not identical to EAF slags, are useful for developing preliminary
modeling parameters.  Unfortunately, of the five elements studied, only Cr is expected to be
found in the slag in any significant quantity. However, in the absence of element-specific
leaching data,  Cr can be considered as a surrogate for the stable oxides expected in slags.
Assuming that the fraction leached is proportional to t'/2, this fraction can be expressed by the
second line of Equation 6-13, where the upper limit of me,, is about 7 x 10"6 d'v', based on Cr in
the BF2 slags and an 18-hr leach test.

Leach Rate

The leach rate from slag is modeled by Equation 6-15, using the parameter values for 1-cm
diameter particles listed hi Table 6-7.

The mill is assumed to produce 150,000 tons of steel per year and 17,600 metric tons of slag as a
by-product.  The slag is assumed to be continuously dumped onto a 1-meter high pile and to be
removed at the same rate, but with a 1-year inventory always remaining in place. The new slag
is mixed uniformly with the old slag—the slag that is removed is thus a representative sample of
this mixture. To model the age-dependent leach rate, we must first determine the age distribution
of the individual particles in the pile. "If there is a constant number, N, of particles in the pile, the
number of particles removed during tune dt is
                                         6-26

-------
                                     dn =  X N dt
             removal rate constant
             1/365  .
             2.74xlO'3d-1
Table 6-9. Fraction of Various Toxic Elements Leached from Slags Using EPA TCLP Protocol
Slag
CaFel
" CaFeSil
CaFeSi2
FeSil
BF1
BF2
Fraction Leached ' . %
As
3.48E-03
• 3.53E-03
5.09E-04
1.54E-04
1.68E-04
9.80E-04
Sb
4.21E-05
2.68E-04
2.37E-04
1.10E-04
1.03E-04
4.29E-04
Cd
3.10E-04
2.40E-04
6.80E-05
1.15E-04
1.10E-04
1.20E-03
Or •''
O.OOE+00
O.OOE+00
5.63E-07
4.82E-07
O.OOE+00
6.00E-06
."' Zn ;
- 3.00E-05
2.70E-05
2.3E-05
2.30E-5
L34E-04
1.23E-03
Assume that v0 particles are added to the pile at some initial time (t = 0). After time t, the
number of these particles left in the pile is given by
                                     v(0 =
By definition, this is the number of particles older than t. The number of particles with ages
between t and t+dt is obtained by differentiating the above expression with respect to t and
changing the sign,

                                   dv =  Xr voe  r' dt

Since this expression is independent of the initial tune, we can generalize it to all the particles hi
the pile:
                                         6-27

-------
                                                                                (6-18)
       dn     =     number of particles in pile with ages between t and t+dt
       N     =     total number of particles in pile

The time-dependent leach rate is derived by differentiating/ in Equation 6-15 with respect to
time:

                                        df,    mtt*
                                     ~~~ ~
Multiplying the above expression by the age distribution function of Equation 6-17 and
integrating over the entire distribution
                           L't =   I  —'—	 dt
                                 m,
                                                                                (6-19)
       LI = leach rate of nuclide i in slag pile

Since the resulting leach rate is time-independent, the fraction leached during one year is
obtained by multiplying the above result by the leaching time and adding the constant term in
Equation 6-15,
                     m, T
                Ft = -!—J£	r- + a,
                                                                                 (6-20)
                     m.
                     -L	 +  a                          = -
                        2         '                     (  r   T
                                         6-28

-------
     F;       = -     fraction of nuclide / in slag leached in time T
     T       =      365d
The concentration in the pore water percolating through the soil (prior to any radioactive decay)
is given by:

                                       C  D F o
                                     =   'S     '  '
     Cip      =      concentration of nuclide z" in pore water (pCi/mL)
     Cig      =      specific activity of nuclide z" in slag
     D       —      depth of slag layer
              =      100 cm
     pg       =      specific gravity of slag
                     2

Dilution. Dilution is modeled by the first of Equations E.27 in the RESRAD manual (Yu 93):
                     dilution factor6 (concentration in groundwater •*• concentration in pore
                     water)
                     0.142
                     length parallel to aquifer flow '•"
                     94m
       A     =      area of contaminated 2»ne
                     V/D
                     8,845 m2
   6  Also called the dilution attenuation factor (DAF)

                                          6-29

-------
       V     =      slag volume
                     M/p
                     8,845 m3
       M     =      one year's slag production
              -      150,000 tons steel/y x 0.13 (wt slag/wt steel)
                     17,690 Mg slag
       p      =      specific gravity of slag
              __.      ^y
       D     =      depth of slag layer
              =      1m
       d,v     =      screened depth of well
                     3m
       Kj     =      saturated hydraulic conductivity of aquifer
                     5500 m/y
       /      =      hydraulic gradient
                     0.02

The values of the last three parameters were also  taken from the generic site analysis in EPA 94b.

Exposure Assessment
                                             3*:
To calculate the dose and risk to an individual drinking the contaminated water, we combine the
expressions in Equations 6-19 and 6-20 with the ground.water dilution factor shown above. The
concentration of a given radionuclide in the ground water is the product of these three
expressions multiplied by an expression for radioactive decay during the travel time to the
aquifer, At,^ which is determined by Equation E.21 of the RESRAD user's manual (Yu 93).  (The
values of At,- of all elements with radioisotopes that are included in the present analysis are listed
in Table 6-5.) The dose or risk to the RMEI is determined by multiplying this concentration by
the appropriate dose or risk factor and the drinking water consumption rate,
                                          6-30

-------
                                                                                (6-22)
                              R  =  C  fG  I e~Xt&t'
                              K,s    ^ipJ **ts * .
Iw     =      annual consumption of water                                   >
     •  =      7.3x105mL/y

Buildup of radioactive progeny

The long travel time of some radionuclides necessitates the consideration of ingrowth of their
long-lived radioactive progeny. Of the five elements listed in Table 6-5 which have
radioisotopes capable of reaching the aquifer, only three—neptunium, protactinium and
uranium—have isotopes which in turn have radioactive progeny with half-lives greater than six
months.

Neptunium.  Table 6-5 shows that it would take neptunium leached from the slag less than 26
years to reach the aquifer. No significant ingrowth of the long-lived progeny of Np-237, the only
neptunium isotope included in the present analysis, would occur during that time.

Protactinium. Pa-231, the parent of Ac-227, which has a half-life of 22.8 years, is the only
nuclide which would have significant ingrowth of long-lived daughter products during its travel
time of 545 years. Actinium has a higher K^ than protactinium and would thus travel more
slowly. Still, its short half-life in comparison with the travel time of the parent indicates that
                                              -I C>i
significant daughter product activity would be found-in the aquifer along with the parent. An
upper bound of the radiological impact of Ac-227 can be calculated by assuming that actinium
has the same K^ as protactinium, so that the two  nuclides would be hi secular equilibrium in the
aquifer.

Uranium.  Table 6-5 shows that it would take uranium leached from the slag 75 years to reach
the aquifer. No appreciable progeny ingrowth from any of the three uranium isotopes included hi
the steel scrap recycling analysis would occur during that time.
                                          6-31

-------
6.4.2  Ingestion of Food Prepared in Contaminated Cookware

One of the finished products examined in this study is cast iron cookware made from potentially
contaminated scrap metal. Radioactivity may leach from such cookware into the food and
subsequently be ingested. The metal content of food cooked in cast iron cookware can be
inferred directly from data in Reilly 85. From the listed concentrations of iron in beef and
cabbage cooked in cast iron utensils, one can derive a weighted average of 13.5 ± 4.7 mg/kg.
The equations used to calculate the dose and risk for this exposure pathway are:
                                Rtg  = Cm CmfIfGig
                                                                               (6-23)
      Cmf     =     concentration of iron in food
                    1.35xlO-sg/g
      Ir       =     amount of food ingested annually
                    Ib + Iv
                    1.45 x 10s g/y
       Ib      =     beef consumption rate
                    7.5xl04g/y
       Iv      =     vegetable consumption rate
                    7.0xl04g/y

6.4.3   Impact of Fugitive Airborne Emissions from the Furnace on Nearby Residents

The impact of fugitive airborne emissions from the furnace on nearby residents was modeled by
means of EPA's Clean Air Assessment Package, using the computer code CAP88-PC.

To calculate the effects of airborne effluent emissions on the RMEI, the map showing locations
of EAF facilities and commercial nuclear power plants and shutdown dates was used to identify
seven EAF facilities which could receive the scrap from two or more nuclear plants in a single
year.  The meteorological data accompanying CAP88-PC was surveyed to identify the
meteorological station nearest to each of these seven facilities. CAP-88 analyses for releases of
C-14 and 1-129 were performed using each of the meteorological data sets — the highest
individual doses from the seven runs were used in the analysis.  The RMEI was assumed to
                                         6-32

-------
reside 1 km from the emission point—default CAP-88 values were used for all other parameters.
Summaries of the CAP-88 analyses can be found in Appendix H.

CAP-88 calculates the risk of cancer fatality, not the risk of cancer incidence, which is the risk
calculated for all other pathways in the present analysis. Furthermore, CAP-88 does not employ
EPA's current risk factors, which are presented in EPA 94a. The CAP-88 dose calculations were
used to estimate the lifetime risk of cancer incidence from one year's exposure to furnace
emissions as follows:
                                         D  G
                                                                                (6-24)
                    lifetime risk of cancer incidence from one year's exposure to nuclide i via
                    pathway y'
                    50-year dose commitment from one year's exposure to nuclide i via
                    pathway7, as calculated by CAP-88 (mrem)
                    risk factor for exposure to nuclide / via pathway j, from EPA 94a
                    dose conversion factor for exposure to nuclide z via pathway j, from FOR
                    11 and 12 (Eckerman 88, 93)
6.4.4  Potential Doses to Individuals Following Disposal of Recycled Metal

The present analysis reflects a broad range of operations and exposure scenarios that could result
from the recycling of steel scrap from nuclear facilities. Not included among these are the
exposures that may be associated with the final disposal of the recycled metal in a municipal
landfill. A quantitative analysis of the potential doses from this stage in the life cycle of recycled
metal is not considered necessary because the individual doses cannot exceed those associated
with the actual handling of the material at the steel mill or the exposure of individual users of the
end products fabricated from recycled metal. As indicated in the discussion of each of the
operations, workers at steel mills and end users of products made from recycled material can
come into close contact with the radionuclides potentially in recycled scrap for extended periods
of time. Recycled metal, if disposed of in a sanitary landfill where it is virtually precluded from
prolonged contact with people, would not likely result in radiation exposures of individuals that
exceed those of the RMEI identified in the present analysis.

                                          6-33

-------
                                   REFERENCES
ANS86
Eckerman 88
Eckerman 93
Envirosphere 84



EPA 89



EPA 94a


EPA94b
Fuhrmann97

Grove 95

ICRP87
American Nuclear Society, 1986.  "Measurement of the Leachability of
Solidified Low-Level Radioactive Wastes by a Short-Term Test
Procedure," ANSI/ANS-16.1-1986.

Eckerman, K. F., A. B. Wolbarst and A. C. B. Richardson; 1988. Limiting
Values ofRadionuclide Intake and Air Concentration and Dose
Conversion Factors for Inhalation, Submersion, and Ingestion, Federal
Guidance Report No. 11,EPA-520/1 -88-020.  U.S. Environmental
Protection Agency, Washington, DC.

Eckerman, K. F., and J. C. Ryman, 1993. External Exposure to
Radionuclides in Air, Water, and Soil, Federal Guidance Report No. 12,
EPA 402-R-93-081. U.S. Environmental Protection Agency, Washington,
DC.

Envirosphere Company, 1984. "Algorithm for Calculating an Availability
Factor for the Inhalation of Radioactive and Chemical Materials," EGG-
2279, EG&G, Idaho.

Exposure Assessment Group, Office of Health and Environmental
Assessment, 1989, Exposure Factors Handbook, EPA/600/8-89/043. U.S.
Environmental Protection Agency, Washington, DC 20460.

U.S. Environmental Protection Agency, 1994. Estimating Radiogenic
Cancer Risk, EPA 402-R-93-076, U.S. EPA, Washington DC 20460.
                                 ^
U.S. Environmental Protection Agency, 1994. Radiation Site Cleanup
Regulations: Technical Support Document for the Development of
Radionuclide Cleanup Levels for Soil, Review Draft, U.S. EPA Office of
Radiation and Indoor Air, Washington, DC 20460.

Fuhrmann, M., 1997. Private communication (see Appendix 1-2.)

Grove Engineering, Inc., 1995. Microshield: Version 4.2.

The International Commission on Radiation Protection, 1987.  Data for
Use in Protection Against External Radiation, ICRP Publication 51,
Pergamon Press, Oxford.
                                        6-34

-------
Jahanshahi 94
Matsuzuru 77
Matsuzuru 77a
Matsuzuru 79
Peterson 83
Reilly 85   -
Yu93
Jahanshahi, S., etaL, 1994."The Safe Disposal of Toxic Elements hi
Slags," in Pyrometallurgyfor Complex Materials and Wastes, pp. 105-
119. -   ,

Matsuzuru, H. etaL, 1977. "Leaching Behavior of Co 60 in Cement
Composites." inAtomkernenergie (ATKE), Bd. 29, Lfg. 4, pp. 287-289.

Matsuzuru, H. and A. Ito, 1977. "Leaching Behavior of Strontium-90 in
Cement Composites," in. Annals of Nuclear Energy, vol. 4, pp. 465-470,
Pergamon Press, Oxford.

Matsuzuru, H. et al., 1979. "Leaching Behavior of Tritium From A
Hardened Cement Paste," in Annals of Nuclear Energy, vol. 4, pp. 417-
423, Pergamon Press, Oxford.

Peterson, H. T., 1983.  "Terrestrial and Aquatic Food Chain Pathways," in:
J. E. Till and H. R. Meyer, eds., Radiological Assessment - A Textbook on
Environmental Dose Analysis, NUREG/CR-3332, ORNL-5968.

C. Reilly, 1985.  "The Dietary Significance of Adventitious Iron, Zinc,
Copper, and Lead in Domestically Prepared Food," Food Additives and
Contaminants, 2:209-215.

Yu, C., et al., 1993.  Manual for Implementing Residual Radioactive
Material Guidelines Using RESRAD, ANL/EAD/LD-2, Argonne National
Laboratory, Argonne, IL.
                                        6-35

-------
Page Intentionally Blank

-------
                                     CHAPTER 7
    RESULTS AND DISCUSSION OF RADIOLOGICAL IMPACTS ON INDIVIDUALS

This chapter presents a summary of the potential impacts of recycling scrap from nuclear
facilities on the RMEI, as well as a brief discussion of the results of these analyses. The dose
and risk from each radionuclide and each pathway are tabulated for every exposure scenario in
Appendix J.  The same results are tabulated more concisely by exposure pathway in Appendix K.
A semi-quantitative evaluation of the uncertainties in the results is found hi Chapter 10, while a
more detailed discussion is presented in Appendix L.

7.1     NORMALIZED DOSES AND RISKS TO THE RMEI

The annual doses and risks for each scenario and for each radionuclide, normalized to unit
specific activities in the scrap, are presented hi Appendix K. Table K-l  lists the normalized
doses to the maximally exposed individual worker from the radionuclides likely to be found on
potentially contaminated steel scrap, calculated for each scenario described hi Chapter 5. (The
mnemonic code for each operation that occurs in this and subsequent tables is identified in Table
5-1.) Tables K-2 to K-4 list the contributions to the dose for each exposure pathway:  external
radiation, inhalation and ingestion.  The corresponding lifetime risks of cancer resulting from one
year's exposure to the same operations are listed in Tables K-5 to K-8.

These tables can be used for two purposes.  First, if the average specific activity of a given
nuclide in scrap is known, the annual dose or lifetime risk of cancer from one year's exposure hi
a given scenario can be calculated by multiplying the specific activity by the appropriate
normalized dose or risk.  If several radionuclides are present, the doses or risks are determined by
summing the contributions of each nuclide. Second, the tables identify the scenario leading to
the maximum dose or risk from each radionuclide. Assume, for example, that the maximum
permissible specific activity of Co-60 hi steel scrap released for recycling that would not cause
any member of the general population to receive a dose greater than 5 mrem EDE1 is to be
determined.  Table K-l shows that the maximum dose, normalized to a unit specific activity (1
pCi/g), is 0.899 millirem hi one year, and that that dose is received by the operator of the
industrial lathe. Thus, the maximum specific activity that would limit the dose of the lathe
    1 This is a purely hypothetical limit for didactic purposes only. No dose limit values are assumed in the present
analysis.

                                          7-1

-------
operator, who has been identified as the RMEI for Co-60, to 5 millirem in any one year is:
                                              Dmax
                                  C(Dmax)  =
                                   Co-60      D
                                              Co-60
       C (D max)     =     maximum specific activity of Co-60 corresponding to maximum
        Co—60
                          permissible dose, Dmax
                    =     5.56 pCi/g
       D""*         -     5 mrem EDE
       Dco-6o        =     maximum annual normalized dose from CO-60
                          0.899 mrem EDE per pCi/g

7.2    MAXIMUM EXPOSURE SCENARIOS

Table 7-1 lists the scenario which would result hi the maximum annual dose to the RMEI in that
scenario for each radionuclide hi the present analysis, as well as the dose and potential risk to ,
that individual.  Several observations can be made about the data hi Table 7-1:

•      The maximum annual normalized dose varies by more than six orders of
       magnitude, from a low of 4 x  IQ-6 mrem EDE per pCi/g from Ni-59 to a high of 8
       mrem from Ac-227+D. Ten of the 44 nuclides and nuclide combinations studied
       would produce maximum doses greater than 1 mrem, while 20 others would be hi
       the range of 0.1 to 1 mrem.

•      Workers are the RMEI for almost all nuclides. The three exceptions are:
                                         7-2

-------
      Table 7-1. Maximum Exposure Scenarios and
Normalized Impacts on the RMEI from One Year of Exposure
Nwslide'
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63 -
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
Ag-llOm+D
Sb-125+D
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
Maximtto jSeeaaiao
Description
Airborne effluent emissions
Lathe operator
Cutting scrap •
Lathe operator
Cutting scrap
Cutting scrap
Cutting scrap
Slag leachate in groundwater
Slag pile worker
Cutting scrap
Cutting scrap
Lathe operator
Lathe operator
Cutting scrap
Airborne effluent emissions
Cutting scrap
Cutting scrap
Slag pile worker
Slag pile worker
Slag pile worker
EAF furnace operator
Slag pile worker
Slag pile worker
Cutting scraip
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Mnemonic
AIRBORNE
OP-LATHE
SCRAPCUT
OP-LATHE
SCRAPCUT
SCRAPCUT
SCRAPCUT
SLGLEACH
SLAGPILE
SCRAPCUT
SCRAPCUT
OP-LATHE
OP-LATHE
SCRAPCUT
AIRBORNE
SCRAPCUT
SCRAPCUT
SLAGPILE
SLAGPILE
SLAGPILE
FURNACE
SLAGPILE
SLAGPILE
SCRAPCUT
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
Anaaal Dose ;
(ribeniEDE
perpCi/g)
8.66e-04
2.02e-01
6.69e-06
8.99e-01
4.39e-06
1.07e-05
9.61e-02
1.52e+00
4.746-01
5.65e-05
2.15e-05
5.16e-02
6.29e-01
6.37e-02
7.91e-01
2.46e-01
8.91e-02
1.77e-02
1.42e-04
3.44e-01
3.08e+00
6.27e-01
3.68e-01
8.00e+00
1.356+00
4.37e+00
6.42e-01
2.84e+00
2.51e+00
3.14e-01
Lifetime* Risk
of Cancer
(perpCi/g)
4.28e-10
1.546-07
2.71e-12
6.84e-07
1.55e-12
4.41e-12
7.31e-08
5.51e-07
3.60e-07
1.176-11
1.416-11
3.93e-08
4.78e-07
4.85e-08
5.04e-07
1.87e-07
6.77e-08
1.36e-08
8.316-11 '
2.61e-07
4.37e-07
4.36e-07
2.36e-07
1.35e-07
6.17e-06
2.32e-07
3.44e-08
3.34e-08
5.20e-08
3.31e-08
                         7-3

-------
                Table 7-1. Maximum Exposure Scenarios and
    Normalized Impacts on the RMEI from One Year of Exposure (Continued)
Nuclide
U-235+D
U-238+D
Np-237-t-D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-241
Cm-244
U-Natural
U-Separated
U-Depleted
Th-Series
Maximum Scenario
#
Description - ,
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
Slag pile worker
EAF furnace operator
Slag pile worker
Slag pile worker
Slagj>ile worker
Mnemonic
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
SLAGPILE
FURNACE
SLAGPILE
SLAGPILE
SLAGPILE
Auaaail>ose
(mterh BDB
perpCt/g)
3.28e-01
2.89e-01
1.53e+00
6.82e-01
7.29e-01
7.29e-01
1.17e-02
6.93e-01
1.21e+00
6.75e-01
3.61e+00
6.18e-01
3.22e-01
4.55e+00
Lifetime Risk
of Cancer
(perpO/g)
5.90e-08
3.55e-08
1.36e-07
4.78e-08
4.73e-08
4.73e-08
4.01e-10
4.46e-08
1.07e-07
6.69e-08
4.78e-07
7.14e-08
3.95e-08
8.86e-07
Sr-90 which, due to its high leachability and low Kj, will readily leach from the slag pile
into the ground water. For this nuclide, the RMEI would be a person living near the slag
storage yard whose drinking water comes from a potentially contaminated well.

C-14, which is potentially volatile (as CO2) and would thus be released to the atmosphere
and be potentially incorporated in food crops and animal fodder grown in the vicinity of
the steel mill. The RMEI would be a person who gets a large portion of his food from
these crops and farm animals.

1-129, which is volatile and would also be released to the atmosphere and potentially
contaminate food crops and animal fodder grown hi the vicinity of the steel mill.  The
RMEI would again be a person who gets a large portion of his food from these crops and
farm animals.

Six scenarios account for the maximum doses from all 44 nuclides and nuclide
combinations.  The ground water potentially contaminated by slag leachate and the
airborne effluent emissions scenarios are discussed above, the remaining four scenarios
are discussed hi the following sections.
                                   7-4

-------
7.2.1   Slag Pile Worker

The slag pile worker would receive the highest doses from many radionuclides that concentrate
in slag, including Nb-94, Ce-144+D, Pm-147, Eu-152, radium and all the actinides except
Ac-227. For the strong Y-emitters—Nb-94, Ce-144+D, Eu-152 and the two radium
isotopes—the primary pathway is external exposure. This results from the worker's spending
four hours per day exposed to slag in the slag yard—a massive source in close proximity. For
the  remaining nuclides, the primary pathway is inhalation of slag dust.

7.2.2   Cutting Scrap

The worker cutting scrap at the scrap yard would receive the highest doses from many of the
nuclides that do not concentrate in the slag:  Fe-55, isotopes of nickel, Zn-65, Mo-93, Tc-99, Sb-
125 and isotopes of cesium. For Zn-65, Sb-125 and cesium, which are  strong Y-emitters, the
primary pathway would be external exposure.  For the other nuclides, the primary pathway is
dust inhalation.

Like the slag pile worker, the scrap cutter would be in close proximity to massive quantities of
potentially contaminated material. His use of a cutting torch causes the metal to volatilize,
                                          *
potentially enhancing the radionuclide concentrations hi the ambient ah".

7.2.3   Lathe Operator

The lathe operator would receive the highest doses 'from four of the nuclides that partition
strongly to cast iron: Mn-54, Co-60, Ru-106+D and Ag-1 lOm+D. His only potential exposure
would be to external radiation from the cast iron hi the lathe, since he would be exposed to
negligible—if any—amounts of ingestible material or respirable particulates from the metal in
this machine.  This individual would receive higher external exposures  from these nuclides than,
say, the scrap cutter because the nuciides hi the lathe were assumed not to be diluted. As was
discussed hi Chapter 5, the scrap yard is assumed to process scrap from the four nuclear plants
that are scheduled for decommissioning in the same year. However, .only 13% of this scrap is
potentially contaminated. Thus, the scrap cutter's exposure is reduced due to the 87%
uncontarninated scrap hi his surroundings. The lathe, however, is assumed to come from a single
furnace heat that happened to be produced from 100% potentially contaminated scrap—a
                                          7-5

-------
reasonable assumption for an occasional occurrence.  Although the radiation source is less
massive, this is more than compensated by the eight-fold higher concentration.

7.2.4  EAF Furnace Operator

The EAF furnace operator would receive the highest doses from Pb-210+D because of his
internal exposure to potentially contaminated dust and soot.  Lead is volatile at steel-melting
temperatures; however, the lead vapors condense to an aerosol dust in the cooler air outside the
furnace.  This dust is inhaled and ingested by the steel mill workers; when it settles and forms
soot it is also inadvertently ingested. According to measured data on dust loadings at various
work stations and our assumptions regarding soot ingestion rates and work assignments, the
furnace operator would have the highest intake of Pb-210 of the workers modeled in our analysis.
Since Pb-210 is a p-emitter with only one low-intensity, low-energy y-Tay, external exposure
would not be a significant pathway.

7.3    EVALUATION OF THE RESULTS OF THE RADIOLOGICAL ASSESSMENT

The analysis was designed to produce a conservative but reasonable assessment of the potential
doses and accompanying risks to individuals resulting from the recycling of steel scrap from
nuclear facilities. This assessment required many assumptions regarding the scenarios and the
physical processes involved.  Several of the assumptions that had a significant effect on the
results are discussed in the following sections.

7.3.1   Dilution of Potentially Contaminated Steel Scrap2

Perhaps the most critical assumptions relate to the dilution of the potentially contaminated steel
scrap by uncontaminated scrap during and after recycling. Relatively little potentially
contaminated steel scrap is being currently released for unrestricted recycling.  Once large-scale
decommissioning of nuclear facilities takes places, it is difficult to predict how much scrap will
in fact be released for recycling, over what period, and with what geographic distribution. The
present analysis made a conservative assumption regarding the maximum likely fraction of
contaminated scrap in the process materials.  Insufficient data is available to determine the
   2 See Appendix G for a comprehensive discussion of the dilution of potentially contaminated scrap.

                                          7-6

-------
probability that the decommissioning scrap from four nuclear plants will be sent to the same
scrap dealer in one year, and that this material will comprise the dealer's entire inventory during
that year. However, due to the plants' geographical proximity to a steel mill (and thus presumably
to a scrap dealer who supplies the mill) and the fact that all four plants are operated by the same
electrical utility, the first assumption is reasonable.  Since data on the distribution and processing
capacity of scrap dealers in that geographical area was not obtainable at the time of this analysis,
the assumption regarding the scrap processing capacity was the only prudent choice to be made.

The choice of the reference steel mill was made ui a similar manner. Although it is based on a
currently operating mill near the four nuclear plants, there is no way of assessing the probability
of the reference mill being the recipient of this scrap.

The assumption regarding the lathe being made entirely of potentially contaminated steel scrap is
conservative but reasonable. Although the average fraction of potentially contaminated steel
scrap in the decommissioning scrap from the four nuclear plants is estimated to be 13%, large
portions of the scrap—the primary coolant piping, for instance—consist entirely of scrap from
service in an environment contaminated with radioactivity. It is reasonable to assume that one or
more rail cars or trucks  hi convoy loaded with potentially contaminated steel scrap would arrive
at an iron foundry, so that one or more heats would be fed entirely by contaminated scrap. The
8-ton lathe could be made from the metal produced from this material.

7.3.2   Exposure Pathways

A number of assumptions were made in modeling the exposure pathways for each scenario.
These will be discussed separately for each pathway.

External Exposure

The external exposures  were modeled using MicroSMeld™ 4.2 for all but two of the scenarios.
MicroShield™ is an mdustry-standard shielding code and produces reliable results for nuclides
with principal y-ray energies greater than 100 keV. Although it is less reliable for assessing
exposures from low-energy v-emitters, this is not a serious drawback in the current analysis. The
nuclides that fall into that category will have their primary impact via the inhalation pathway, so
                                          7-7

-------
that any inaccuracy in assessing the external exposures would have little effect on the limiting
doses listed ha Table 7-1.

FOR 12 (Eckerman 93) provides a highly accurate means of assessing the external exposure from
an idealized source geometry, if the receptor is a person standing on the source and the source
has the same elemental composition as the soil used in the FOR 12 dose calculations.  FGR 12
gives a reasonable approximation to the three scenarios—the slag storage yard, the road built
with slag and the scrap yard—to which it was applied. In both cases, the roughness of the
surface would tend to reduce the actual exposures from the FGR 12 predictions, as would the
higher effective atomic number of the source material.3

Inhalation

The major parameters that affect the dose via the inhalation pathway are the atmospheric
concentration (dust loading) and composition of the dust. The dust loading was, hi most cases,
based on measured values for similar operations. Thus, the dust hi the areas occupied by the
steel mill crane operator, the furnace operator and the operator of the continuous caster were
based on reported measurements for such workers at an actual EAF steel-making facility, albeit
one that primarily produced stainless steel.  Since only analyses for toxic trace constituents hi the
dust had been performed, it was not possible to ascertain the origin of the dust, which would
enable a determination of its hypothetical radioactive contamination. It was therefore assumed
that it came from the furnace emissions—i.e., it had the same composition as baghouse dust.
Since the furnaces are the primary source of airborne emissions in a steel mill, this is a
reasonable assumption.

The dust loading hi the scrap yard was more difficult to  determine, since scrap processors do not
routinely monitor dust levels.  Newton et al. reported that cutting metal with an oxy-acetylene
torch produces aerosol concentrations of 15 mg/m3 (Newton 87) and that these particles were
primarily of respirable size. Thus, a concentration of 15 mg/m3 and a respirable fraction equal to
1 is an upper limit for the scrap cutting operation. However, the cutter works outdoors, so the
aerosols from his torch would have more  of a chance to  disperse.  It was thus assumed that the
total dust is equal to the ACGIH Threshold Limit Value (TLV) of 10 mg/m3 for nuisance dust,
   3 A discussion of the anticipated effect of atomic number on calculations using the FGR 12 dose coefficients can be
found in Section H.2.1 of Appendix H.

                                          7-8

-------
which is also the OSHA PEL.  This is a key assumption, inasmuch as dust inhalation by the scrap
cutter is the major contributor to the maximum annual dose from several radionuclides.

Another key assumption in the scrap cutting scenario is that the dust has the same specific
activity as the scrap. An argument could be made that much of the radioactive contamination
will be on the surface and that it is these surface layers that are the primary sources of the dust.
To counter that argument, we observe that the scrap would have undergone surface
decontamination prior to being released, so that loose surface activity would have been removed.
The cutting operation is assumed to be the major source of the dust.  The aerosols are produced
by the melting and volatilization of the steel; their composition can therefore be assumed to be
the same as the overall composition of the scrap.

Ingestion

Ingestion of radionuclides, either by inadvertently ingesting contaminated dust or soot during the
working day or by eating food contaminated by radioactive cookware, proved to be a major
contributor to the maximum doses of two of the RMEI. One is the furnace operator exposed to
Pb-210. The soot ingestion rate of this worker is based on one-half of the high-end value for soil
ingestion for outdoor workers listed in the EPA Exposure  Factors Handbook (EPA 89), and on
observations of the highly dusty, sooty environment of the steel mill. Thus, the actual ingestion
rate is not likely to be more than twice what was postulated, nor is it likely to be very much
lower. The soot was assumed to have the same composition as baghouse dust, since, like the
dust hi the air, its primary source is the fugitive emissions from the furnace. (See discussion of
the inhalation pathway, above).

Ingestion is the only pathway for the RMEI exposed to Sr-90—the person whose drinking water
well may become contaminated by leachate from the slag  pile. This analysis includes a number
of conservative assumptions, which are discussed in this section.

The first assumption regards the leachability of strontium  from slag. The leach rate was
calculated using diffusion coefficients which were in turn calculated using preliminary data from
on-going EPA-sponsored experimental research being conducted by the Brookhaven National.
Laboratory.  One source of uncertainty in the experimental data—the diffusion coefficients
calculated from measurements on three EAF slag samples showed that the highest of the three
                                          7-9

-------
 values, which was conservatively adopted for the present analysis, was four times greater than
 the lowest value. Another is the assumed size of slag particles. Our calculations assumed an
 average particle size of 1 cm. If the slag consisted primarily of finer particles, the leach rate
 would be correspondingly larger.4

 Another assumption concerned the K$ of the soil layer under the slag.  For all elements
 considered in this study, the lowest reasonable KJs, which had been identified in EPA 94, were
 used to model the transport of the radionuclides through the soil. The metallic elements
 (including strontium) are more mobile in an acidic environment. Slag, however, is basic. Thus,
 the leachate would be basic, causing elements like strontium to be retarded in the soil. This
 would prolong its migration time, leading to more radioactive decay of Sr-90 and perhaps even
 preventing any significant amount from reaching the aquifer. In such a case, of course, another
 scenario would result in the maximum dose.

 7.3.3  Mass Distribution and Partitioning ofJIontaminants
                                         i

 The mass distribution of metal and non-metallic components among the steel or cast iron, slag,
"dust and home scrap was determined from a definitive study of the literature and consultations
                            [      „                    "&
 with other research workers and technical experts. The data on partitioning of contaminants
 among these various media was less definitive.  Nevertheless, a major and largely successful
 effort was made to combine the observed partitioning with thermodynamic calculations to
 produce a set of reasonable and defensible concentration factors. The only conscious
 conservatism that was introduced into this phase of the analysis was the simultaneous use of
 high-end partition fractions for two or more media, wjiich, as was discussed at the end of
 Section 6.2, would overstate the exposure of the operator of the continuous caster to four of the
 radionuclides studied.  Since this individual did not prove to be the RMEI for any nuclide, this
 assumption has no effect  on the maximum doses listed in Table 7-1.

 7.3.4  Scenario Selection

 The scenarios used in the present analysis were selected from a much longer list which had been
 examined in an earlier analysis of recycling residually radioactive steel scrap (SCA 95).
    4 A more detailed discussion of the uncertainties in the slag leaching scenario can be found in Appendix L.

                                          7-10

-------
Scenarios in the previous analysis which were redundant or which had no potential for producing
the maximum doses from any nuclide were dropped from the analysis of the RMEI. (Some of
these were included in the study of the collective impacts described hi Chapter 9). Given the
conservative assumptions used, it is improbable that any plausible scenario would produce
greater impacts than those studied.

One scenario that'was not part of the assessment of the RMEI, but which was included in the
assessment of collective impacts, was the use of slag as a soil conditioning agent. A scoping
analysis was performed to determine if this would be a significant exposure pathway for any of
the radionuclides considered in the present analysis. A brief description of this assessment is
presented below. A more detailed discussion can be found in Appendix H.

Because of its high lime (CaO) content, slag is sometimes used to raise the pH of acidic soils.
According to a vendor of gardening supplies, even highly acidic soils do not require more than
about 100 Ib of liming agent per 1,000 ft2. The liming agent supplied by this vendor contained
48% lime, which is comparable to the CaO content of steel slags listed in  Appendix I. Assuming
that slag were applied to agricultural soil hi the same quantity, and that it were mixed into the top
six inches of soil (the assumed plow depth), the doses from the consumption of agricultural
products grown hi this soil were calculated.

The normalized annual dose to the RMEI via the food ingestion pathway was calculated for each
radionuclide that partitions to the slag by using the dose factors calculated for the agricultural
produce pathway for a generic site with radioactively contaminated soil, as listed in Table 3-1 of
EPA 94.  The doses calculated for this scenario for each radionuclide that would partition to the
slag were one to three orders of magnitude less than the doses to the RMEI for that nuclide.
                                          7-11

-------
                                    REFERENCES


Eckerman 93 Eckerman, K. F., and J. C. Ryman, 1993.  External Exposure to Radionuclides in
             Air, Water, and Soil, Federal Guidance Report No. 12, EPA402-R-93-081. U.S.
             Environmental Protection Agency, Washington, DC.

EPA 89      Exposure Assessment Group, Office of Health and Environmental Assessment,
             1989. Exposure Factors Handbook, EPA/600/8-89/043. U.S. Environmental
             Protection Agency, Washington, DC 20460.

EPA 94      U.S. Environmental Protection Agency, 1994. Radiation Site Cleanup
             Regulations:  Technical Support Document for the Development of Radionuclide
             Cleanup Levels for Soil, Review Draft, U.S. EPA Office of Radiation and Indoor
             Air, Washington, DC 20460.

Newton 87   Newton, G. J., et al, 1987, "Collection and Characterization of Aerosols from
             Metal Cutting Techniques Typically Used in Decommissioning Nuclear
             Facilities," in American Industrial Hygiene J., 48: 922-932.

SCA 95      S. Cohen & Associates, Inc. Analysis of the Potential Recycling of Department of
             Energy Radioactive Scrap Metal. U.S. Environmental Protection Agency, Office
             of Radiation and Indoor Air, Washington, DC.
                                        7-12

-------
                                    CHAPTER 8

              DETECTION AND MEASUREMENT OF CONTAMINATION

8.1     STATEMENT OF PURPOSE

The purpose of this chapter is to address issues relating to the ability to detect and quantify
contamination levels of residually-contaminated scrap metal intended for recycling and in newly
produced steel containing residually-contaminated scrap that was recycled. Of primary concern
are questions regarding the probability of:

  (1)   accepting scrap metal for recycling that is assumed to be non-contaminated when in fact
       it is;

  (2)  , accepting scrap that contains residual contamination well in excess of acceptable release
       limits;

  (3)   rejecting scrap metal that is falsely assumed to contain radioactivity in excess of
       acceptable limits, and

  (4)   unknowingly releasing steel derived from recycled scrap for unrestricted use (i.e.,
       consumer products) at levels of contamination that exceed free-release criteria.

At a minimum., these complex issues require an understanding of the nature of radioactive
contamination, current applicable release limits, release survey measurement methods, and
limitations imposed by counting statistics, instrumentation, and radioanalytical and
radiochemical analyses. These and other topics are briefly discussed in this chapter.
                                            i
8.2     GUIDELINES AND STANDARDS FOR FREE RELEASE OF SCRAP
       ESTABLISHED BY THE NRC AND DOE

Radioactive contamination of scrap metals from commercial nuclear power plants and DOE
facilities that are potentially available for recycling primarily exists in the form of surface
contamination, which is limited to a few-microns-thick layer of radioactivity on the metal
surface. Surfaces may be classified as "internal" and "external."  For a considerably smaller
fraction of potentially releasable scrap metal, the contamination may exist as bulk or volumetric
                                         8-1

-------
  (1)  Contaminated External Surfaces - Plant surfaces become contaminated over the lifetime
       of the plant from leaks, spills and airborne migration of radionuclides.  The specific
       activity is low. However, the contamination is widespread and easily mobilized.

  (2)  Contaminated Internal Surfaces - Activated corrosion and fission products from the fuel
       travel through the reactor coolant water system throughout the radioactive liquid system
       in the plant.  A portion forms a hard metallic oxide scale on the inside surfaces of pipes
       and equipment. This contamination is not easily mobilized.

  (3)  Activated Stainless Steel - Reactor internals, composed of Type 304 stainless steel,
       become activated by neutrons from the core.  Radionuclides have very high specific
       activities and are immobilized inside the corrosion-resistant metal.

  (4)  Activated Carbon Steel - The reactor pressure vessels are made of SA533 carbon steel
       which becomes activated by neutrons bombardment. The specific activities are
       considerably lower than in the stainless steel internals.  However, the binding matrix is
       much less corrosion resistant.

8.2.1  NRC:  Regulatory Guide  1.86 (1974 and 1982")


Criteria for residual contamination levels used to decommission facilities for unrestricted use
have been based on interim guidance contained in Regulatory Guide 1.86, "Termination of
Operating Licenses for Nuclear Reactors," first published hi June 1974, for surface
contamination plus case-by-case considerations for direct radiation.  Limit values from
Regulatory Guide 1,86 (1974)  and (1982) are  shown in Table 8-1.  The only significant
difference between the 1.86 (1974) table and the NRC  (1982) table is that the NRC (1982)
table has a footnote Kf," which states:


       "The average and maximum radiation levels associated with surface contamination
       resulting from beta-gamma emitters should not exceed 0.2 mrad/hr at 1 cm and 1
       mrad/hr at 1 cm, respectively, measured through not more than 7 milHgrams per square
       centimeter of total absorber."
                                          8-2

-------
         Table 8-1   Regulatory Guide 1.86 Acceptable Surface Contamination Levels
Nuclides(B)
U-nat, U-235, U-238, and
associated decay products
Transuranies, Ra-226,
Ra-228, Th-230, Th-228, Pa-
23 l,Ac-227, 1-125, 1-129
Th-nat, Th-232, Sr-90,
Ra-223, Ra-224, U-232,
1-126,1-131,1-133
Beta-gamma emitters (nuclides
with decay modes, other than
alpha emission or spontaneous
fission) except Sr-90 and others
noted above.
Average 
-------
8.2.2  DOE Order 5400.5.

DOE's basic radiation protection standards for cleanup of residual material that include release of
property are defined in Chapter IV of DOE Order 5400.5. Paragraph 4.d. states that:

       "The generic surface contamination guidelines provided in Figure IV-1 are applicable to
       existing structures and equipment.  These guidelines are generally consistent with
       standards of the NRC (NRC 1982) and functionally equivalent to Section 4,
       'Decontamination for Release for Unrestricted Use' of Regulatory Guide  1.86, but apply
       to nonreactor facilities. These limits apply to both ulterior equipment and building
       components that are potentially salvageable or recoverable scrap."

DOE release limits are, therefore, nearly identical to those of Regulatory Guide 1.86, inclusive of
the NRC 1982 footnote on average and maximum dose rates of 0.2 and 1.0 mrad/hr at 1 cm. For
the TRU grouping, however, DOE has held in reserve surface contamination values pending the
development of standards more applicable to DOE facilities. Another difference is DOE's
residual limit for external gamma radiation. Section 4.C. of Chapter IV, DOE Order 5400.4
states that:

       "... the average level of gamma radiation inside a building or habitable structure on a site
       to be released without restriction shall not exceed the background level by more than 20
       uR/h."

8.2.3   Release Criteria for Volumetric Contaminants

Currently, neither the NRC or the DOE have established guidance or criteria with respect to the
release of volumetrically contaminated or activated metals. DOE Order 5400.5, Chapter II,
Section 5.C.(6) states:

       "No guidance is currently available for release of material that has been contaminated in
       depth, such as activated material or smelted contaminated metals (e.g., radioactivity per
       unit volume or per unit mass)."

Such material may be released based on ad hoc criteria and survey techniques approved by the
DOE Office of Envkonment, Safety, and Health (EH).
                                          8-4

-------
For the NRC, volumetric release criteria had been proposed under its 1992 Below Regulatory
Concern (BRC) policy. The BRC policy proposed that the average dose to individuals should be
less than 10 mrera per year for each exempted practice. In addition, an interim dose criterion of
1 mrem per year average annual dose to individuals had been proposed for those practices
involving widespread distribution of radioactive materials in items such as consumer products or
recycled material and equipment. While NRC's BRC.policy has subsequently been withdrawn,
they have issued volumetric release criteria on a case-by-case basis.

8.3    VERIFICATION OF RESIDUAL CONTAMINATION OF MATERIALS RELEASED
       FOR UNRESTRICTED USE  '

Both DOE and NRC require that all materials prior to release must be surveyed to determine
whether both removable and total surface contamination is less than specified limits. DOE and
NRC consider material to be potentially contaminated if it has been used or stored in radiation
areas that could contain unconfined radioactive material or that are exposed to beams of particles
capable of causing activation (i.e., neutrons, protons, etc). Surfaces of potentially contaminated
property must be surveyed using instruments and techniques appropriate for detecting stated
limits.

Standard instrumentation used for a final release survey are broadly categorized as two types:
portable field instruments and laboratory instruments. In general, field instruments are employed
                   i
to assess total surface contamination (i.e., fixed and removable) and are employed in two discrete
modes.  The first involves "fixed point" or "direct" measurements, in which the detector is held
in a defined stationary position above the surface fqr a preselected time period. Direct
measurements do not attempt to evaluate the entire surface but provide a sampling of surface
contamination levels. For relatively large surface areas of suspected contamination, direct
measurements are taken at systematic locations to supplement scan surveys.  These
measurements are collected according to a predetermined pattern without regard to radiation
level. Judgmental direct measurements (i.e., not chosen on a random orsystematic basis) may be
collected at locations where anomalous radiation levels are observed during scan surveys or at
locations suspected of being contaminated.

The second mode of operating field instruments involve surface scans or "frisks." In this survey,
the detector is passed slowly over the surface at a standard scanning speed while maintaining a
constant distance above the surface. Since direct measurements have a low probability of

                                          8-5

-------
identifying small areas of elevated activity, scan surveys are performed to locate such small
areas, which typically represent a very small portion of the item being surveyed. In this survey,
the detector is passed slowly over the surface at a standard scanning speed while maintaining a
constant distance above the surface. Additionally, scan surveys may be performed to determine
general contamination levels over large areas.

Laboratory instrumentation provides complementary data to field survey data.  Wipe or "smear"
samples provide an acceptable method for demonstrating the level of residual surface
contamination that is removable. To assess volumetric sources of contamination, a portion of the
potentially contaminated medium may be collected and analyzed using both chemical and
instrument techniques to quantify the low levels expected in samples. Laboratory  analyses can
also identify individual radionuclides and establish then* relative abundance when a number of
contaminants are present in a mixture.

8.3.1   Total Alpha and Beta-Gamma Direct Measurements

Total alpha contamination measurements are typically performed using a zinc sulfide (ZnS)
alpha scintillation detector or a thin-window gas proportional (GP) counter with a  portable sealer.
The probe is held at a fixed location and a 30  to 60 second count is taken. The distance between
the surface and the probe is maintained at approximately 0.5 cm to 1 cm. The efficiency of the
typical ZnS and GP alpha instrument is between 13% and 22%, which yields an average
efficiency of about 17% for ZnS and 19% for GP counters. Expected background count rates
range between 0 and 4 cpm for both detectors. The window area of the standard ZnS probe is 59
cm2 while that of the GP detector is  126 cm2.

Total beta contamination measurements are likely to be performed using a thin-window (1.4 to 2
mg/cm2) "pancake" Geiger-Mueller  (GM) detector or a GP counter with a portable sealer. The
probe is held at a fixed location and  a 30 to 60 second count is taken. Beta counting efficiencies
for these instruments are extremely energy dependent, ranging from 35 - 40% at higher energies
(>1000 keV maximum beta energy)  to less than 5% as energies drop below 100 keV. The
window area of the standard pancake probe is 20 cm2 while that of the GP detector is 126 cm2.
Background count rates for these detectors may be as low as 50 cpm and 350 cpm, respectively;
however, background may increase depending on ambient radiation fields and/or contamination
levels.
                                          8-6

-------
8.3.2  Surface Scanning for Total Alpha and Total Beta Contamination

The term scan survey is used to describe the process of moving portable radiation detectors
across a suspect surface with the intent of locating areas of contamination. The detectors used in
surface scanning are identical to those used for direct measurements; however, in the scanning
mode, the instrument is equipped with a ratemeter that typically has established tune constants)
settings of 2 and 10 seconds. The probe is passed slowly over the surface at scanning speeds of 5
to 15 cm per second at a relatively constant distance of 1 cm or less. The probability of detecting
localized areas of contamination using scan surveys is not only affected by the sensitivity of the
survey instrumentation, but also by the surveyor's ability.  Personnel conducting scan surveys
must interpret the audible output or visual reading of a portable survey instrument to determine if
the signal being heard or seen exceeds the background level by a sufficient margin to conclude
that contamination is present.

8.3.3 Surveys for Total Gamma Contamination

Sodium iodide, NaI(Tl), scintillation detectors are normally used for scanning areas for gamma
emitters because they are very sensitive to gamma radiation, easily portable, and relatively
inexpensive.  The detector is held close to the surface (~6 cm) and moved in a serpentine (snake
like, "S" shaped) pattern while walking at a speed which allows the investigator to detect the
desired investigation level. A scan rate of approximately 0.5 m/s is typically used for distributed
(large area) gamma emitting contamination; however, this value must be adjusted depending on
the expected detector response and the desired investigation level.  When surveying for small
elevated areas of contamination, much slower scan rates will be required. Nal may also be used
to perform direct measurements. Direct monitoring for alpha- and beta-emitters will usually
result in better MDCs than gamma monitoring. However, for radionuclides that decay by
electron capture, gamma monitoring is the only viable  survey method since, in most cases, no
betas are emitted.

The most sensitive (highest efficiency) portable detector is a 2-inch diameter (surface area - 20
cm2) by 2-inch thick Nal. Specialized detectors are also available that optimize detection of low-
energy (below 100 keV) gamma and x-radiation. Detector efficiency is largely a function of
gamma energy. A  detector's intrinsic efficiency is a measure of the number of counts produced
relative to the number of photons incident upon the detector's surface.  Intrinsic efficiency is

                                           8-7

-------
dependent upon the photon energy, detector wall thickness, and the thickness of the scintiilator.
The source-detector geometry permits calculation of the fluence per unit source strength, based
upon the size of the source and the distance to the detector. The combination of mtrinsic
efficiency, fluence per unit source strength, and detector surface area will determine the overall
detector efficiency.

Background count rates with Nal detectors are dependent upon ambient background dose rate,
the size of the detector, and the ratemeter/scaler setup. More advanced ratemeter/sealers permit
use of threshold and window settings, which permit use of the Nal to count in the specific energy
region of interest and reduce low-energy "noise." Generally, background count rates are in the
range of 3,000 to 12,000 cpm, depending upon detector size.

8.4    LOWER LIMIT OF DETECTION AND MINIMUM DETECTABLE
       CONCENTRATION

Survey measurements of scrap metal for unrestricted release are not without potential error.
There are two types of errors:

  •    Type I Error is the error made in assessing that contamination exists, when in fact it does
       not exist (false detection).

  •    Type II Error is the error made in concluding that contamination dose not exist, when in
       fact it does (false non-detection).

For fixed-point measurements, potential errors may arise due to the fact that residual
                                             &£
contamination may be variable or localized to small areas. Thus, direct measurements and wipe
samples have a statistical probability of measuring contamination levels that are not
representative of overall contamination levels. A more significant contribution to Type I and n
errors is the random nature of radioactive decay and the interference of background radiation
levels.
              e of LLD and MDC .YgjuesJn Survey Measurements. The lower limit of
detection (LLD) is the smallest amount of sample activity that will yield a net count for which
there is a confidence at a predetermined level that activity is present.  The LLD is an a priori
estimate of the detection capabilities of a given measurement process. It is related to the
characteristics of the counting instrument, and is not dependent on other factors involved in the

                                          8-8

-------
measurement method or on the sample characteristics.  For this reason, a statistical value called
the minimum detectable concentration (MDC) has been introduced.  The MDC is a level of
activity concentration which is practically achievable by an overall measurement method. As
distinguished from the LLD, the MDC considers not only the instrument characteristics
(background and efficiency), but all other factors and conditions which influence the
measurement. It is an a priori estimate of the activity concentration that can be practically
achieved under a specified set of typical measurement conditions. These include the sample size
or detector area, counting time, self-absorption and decay corrections, chemical yield, and any
other factors that comprise the activity concentration determination.

MDC values are commonly based on both Type I and Type II errors by specifying a count rate
for which the presence of contamination has a probability p and the probability of falsely
concluding its presence is defined by (1 - p). Both NRC and DOE recommend that MDC levels
be determined by setting the risks of false detection and false non-detection equal, to accept a 5%
chance of incorrectly detecting activity when it is absent, and a 95% confidence that activity will
be detected when it is present. Alternatively, the MDC may be considered to represent the
smallest concentration of radioactive material in a sample that will be detected with 95%
probability, with a 5% probability of falsely concluding that contamination is present when in
fact it is not.

A third factor that profoundly  affects Type I and Type II errors is the mode of instrument
operation. This is especially true for instrumentation used in a scanning mode. For survey
measurements that involve scanning a surface, MDCjalue are dictated by the complex
                                               -*_,
interrelationship of (1) background levels, (2) source-detector geometry, (3) instrument time
constant, (4) variability of surface contamination relative to detector surface area, and (5)
scanning velocity.

8.4.1 MDCs for Surface Scanning for Small Areas of Contamination

The MDC for detection of small areas of contamination using surface scanning is calculated
using the following equation:
                                           8-9

-------
        AfDC =
3
+ 4.65
2 t W
R 60 * V
- 	 * ( Y^ Y * * ^ * 	 * 77F
                                                          (Eq. 8-1)
60 * r
                                          100
where:
             MDC
             BR
             W
             60
             V
             A
             HF
         minimum detectable concentration (dpm/lOOcm2)
         detector background count rate (cpm)
         detector width (cm)
         conversion factor (s/min)
         detector scan rate (cm/s)
         yield for emission I (ptcle-emitted/d)
         detector efficiency for emission I (c/ptcle-emitted)
         detector area (cm2)
         surveyor efficiency (%).
8.4.2 MDC for Surface Scanning for Large Areas of Contamination

The MDC for detection of large areas of contamination using surface scanning is calculated
using the following equation:
        MDC =
•2 4, A «

2 *T ^ ^
60 l^ '
2 *T
R *
•"»
60
A
•Y P ) * . * fTF"
1 100
                                                          (Eq. 8-2)
where:
             MDC
             BR
             t
         minimum detectable concentration (dpm/100cm2)
         detector background count rate (cpm)
         meter time constant (s)
                                        8-10

-------
             60
             Y,
             e,
             A
             HF
conversion factor (s/min)
yield for emission I (ptcle-emitted/d)
detector efficiency for emission I (c/ptcle-emitted)
detector area (cm2)
surveyor efficiency (%).
8.4.3  MDC for Direct Measurements
The MDC for direct measurements is calculated using the following equation:
             MDC =
3 + 4.65
* * fVr
60 (2-F
***-
* 60
A
1 ' 100
                                                  (Eq.8-3)
where:
             MDC
             BR
             t
             60
             Y,
             6,
             A
minimum detectable concentration (dpm/lOOcm2)
detector background count rate (cpm)
count time (s)
conversion factor (s/min)
yield for emission I (ptcle-emitted/d)
detector efficiency for emission I (c/ptcle-emitted)
detector area (cm2).
8.5    RADIONUCLIDE MDCS FOR SURFACE CONTAMINATION

MDCs have been calculated for 40 radionuclides for surface scanning of small areas of
contamination, surface scanning of large areas of contamination, and direct measurements (for a
counting time of one minute). For the small area scan surveys, the MDCs represent the lowest
level of a small but elevated area of surface contamination (e.g., "hot spot") that is likely to be
detected with a high degree of confidence. For the large area scan surveys, the MDCs represent
the lowest level of distributed sources of surface contamination that are likely to be detected.
                                         8-11

-------
The MDCs for direct measurements represent the sensitivity of the detection process at the
location at which the survey is taken.

These MDCs should be considered as realistic, but not conservative. Background count rates,
detector scan rates, and instrument counting efficiencies were chosen as relatively low, slow, and
high values respectively, typical of what may be encountered under controlled, laboratory
conditions. In the field, the background count rate may vary widely, scan rates will vary from
individual to individual and day to day, and instrument efficiencies may be reduced by a number
of factors including the source-detector geometry, source-to-detector distance, the physical
condition of the surface being monitored, and surface coatings such as paint, dust, oil,  or water.
The MDCs for scan surveys have also been adjusted to account for the surveyor's ability in
deciding whether the instrument response represents only background activity, or whether it
represents residual contamination in excess of background.

MDCs have been calculated for two common beta monitoring instruments, GP and GM
detectors, two alpha monitoring instruments, GP and ZnS detectors, and one gamma instrument,
a NaI(Tl) detector.  Detector sizes and areas were selected to represent commercially available
instruments. Efficiencies represent average values compiled from historical instrument
calibration data, and should be considered as the ideal efficiencies obtained under laboratory
conditions.

Tables 8-2, 8-3, and 8-4 present a comparison of MDCs  for small area scan surveys, large area
scan surveys, and direct measurements with the limits given hi NRC Regulatory Guide 1.86.
The NRC's maximum surface contamination limits are used for the small area survey, while the
average limits were used for the large area scan and the direct measurements. For small area
surface scanning, 22 of the 40 radionuclides listed were detectable. Radionuclides which
Regulatory Guide 1.86 places in group 2 (transuranics) are typically not detectable due to the 300
dpm/lOOcm2 limit Most of the radionuclides with a 15,000 dpm/lOOcm2 limit were detectable.
These include the alpha-emitting uranium isotopes, as well as the higher energy beta-gamma
emitting radionuclides such as Co-60, Cs-134, Cs-137+D, Sr-90+D, and Tc-99. Tables 8-3 and
8-4 demonstrate the influence of longer counting times in the large area surface scans and direct
measurement techniques. In these cases, 35 and 3.6 of the 40 radionuclides were detectable
compared to the Regulatory Guide 1.86 limits.  The radionuclides not detectable were  the very
low energy beta emitters due to their low counting efficiency.
                                          8-12

-------
These results indicate that, with respect to Regulatory Guide 1.86 limits, the surface scanning
technique can detect small elevated areas of contamination only for radionuclides with a 15,000
dpm/lOOcm2 limit. However, most radionuclides are detectable when monitoring large areas of
contamination using surface scanning as well as direct measurement, even those radionuclides
with a Regulatory Guide 1.86 limit as low as 100 dpm/lOOcm2.
                                                                     _i
Chapter 7 of this Technical Support Document discusses the derivation of normalized doses for
individuals who may be exposed to radiation as a result of the recycling of scrap. Based upon
these normalized doses, surface concentration limits (in dpm/100 cm2) have been derived for
annual doses of 15,1, and 0.1 mrem. The most challenging survey condition is the ability to
identify small areas of contamination by surface scans. Table 8-5 presents a comparison of
radionuclide MDCs  for small area scan surveys with these derived concentration limits (DCLs)
for these three dose values. For a dose limit of 15 mrem/y, all but one radionuclide is detectable,
and even at 1 mrem/y, most radionuclides (31 of 40) can be detected. The situation reverses
itself at 0.1 mrem/y as only 11 of 40 radionuclides are detectable.

Tables 8-6 and 8-7 compare DCLs with radionuclide MDCs for large area scan surveys and
direct measurements at the three dose values.  The data in  these tables demonstrate that all
radionuclides listed can be detected at the 15 mrem/y and  1 mrem/y dose levels.  Even at 0.1
mrem/y, most radionuclides are detectable (27 of 40 with large area scans and 35 of 40 with
direct measurements).

It should be reiterated that these results are based upon parameter values that represent optimal
monitoring conditions. Variability of the parameters used in MDC calculations can contribute
individually as well  as collectively to cause an inherent variability in MDCs likely to be
encountered under field conditions. Variability of these factors will cause an increase hi MDC
since the MDCs were calculated to represent optimal values, representative of laboratory
conditions.

The factors having the most impact upon MDCs are background count rate, detector scan rate
(for small areas of contamination), human factors efficiency (for surface scanning), source-
detector geometry, source to detector distance, surface material condition, and surface coatings.
The last four factors impact the counting efficiency. Under field conditions, the combined
variability of these parameters could cause actual MDCs to be a factor of 10 higher than shown
                                          8-13

-------
in. Tables 8-2 through 8-7. Under such conditions, the detectability of radionuclides will be
reduced. For instance, in the case of monitoring for small areas of contamination with surface
scans, Table 8-5 showed that 31 of 40 radionuclides could be detected at a DCL of 1 mrem/y. If
MDCs were increased by a factor of 10, only 11 of 40 radionuclides would be detected. For
distributed sources of contamination, this reduction in detectability is not quite as significant.
Table 8-6 shows that for 1 mrem/y, all radionuclides area detectable. If MDCs increase by a
factor of 10, 27 of 40 radionuclides are detectable. The smallest reduction in detectability occurs
in direct monitoring, where the number of radionuclides detectable in Table 8-7 drops from all 40
to 35 of 40.

This reduction in detectability may be offset for materials that are contaminated with multiple
radionuclides. It may be possible to measure just one of these contaminants and still demonstrate
compliance with dose limits  for all of the contaminants present. In using one radionuclide to
measure the presence of others, a sufficient number of high-sensitivity measurements, spatially
separated throughout the material being surveyed, should be made to establish a consistent ratio.
The advantage of surrogate methods is that both time and costs can be saved if the analysis of
one radionuclide is simpler than the analysis of the other. The surrogate method should be used
with caution due to the potential for  shifts or variations in the radionuclide ratios.  Physical or
chemical differences between the radionuclides may cause the ratios to change following
decontamination measures.

In dealing with mixtures of radionuclides, the DCLs for each radionuclide must be taken into
consideration to account for the total dose relative to the appropriate dose limit. One method for
adjusting DCLs is to modify the assumptions made during the exposure pathways modeling to
account for multiple radionuclides. The DCLs shown in the tables in this chapter are based upon
the assumption that no other radionuclides are present. A second method for adjusting DCLs is
to use the unity rule to adjust the individual DCLs. The unity rule is satisfied if the sum of the
ratios of the radionuclide concentration and its DCL is less than or equal to one.
                                           8-14

-------
                                    Table 8-2




Detectability of Radionuclides (Small Area) by Surface Scan* Relative to RG 1.86 Limits
<' '• * ' '• ••
%didkc|lde.

Ac-227+D
Ag-llOm-HD
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237-f-D
Pa-231
Pb-210-f-D
Pm-147
<\-- •->*,-, t: ' !,.,m
<•.•"< - , *-..
*• , . V\. I."" '' ' v >
'$. 'i JJefa. . >;
s • s f ": '*'.•"•' $ ' ' f 'f, *
rv-K ••'• .W; x
440 2,500
3,300 23,000
4,300 78,000
3,900 31,000
650 2,800
24,000 610,000
2,300 13,000
1,800 9,600
1,300 6,600
2,900 15,000
31,000 ND
3,600 30,000
3.9E+06 1.5E+07
46,000 ND
1,600 8,000
700,000 ND
8,800 - ND
1,100 7,000
4,600 220,000
960 5,600
3,300 22,000
SCppMie&cm2)': *• . :/
; ^ *' • •'• '• •} -f '
t ::;;:;JA^:;V^
r::'i^':...;;'\j'^^:v
85 320
ND ND
420 1,600
ND ND
ND ND
420 1,600
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
440 1,600
430 1,600
420 1,600
ND ND
'' "''•', -.-;,
i> 1...,.^-+
. Oamma
t :> v^- ' ';
-•im •
120,000
120,000
340,000
ND
600,000
8.1E+06
200,000
160,000
420,000
139,000
ND
320,000
360,000
1.2E+06
180,000
ND
ND
160,000
860,000
3.9E+06
3.5E+06
^ 'l:'/'Wl-M ••." :l
v'^jN^ni ' ;^1>6tectafeie
^IHiPI..;';.!.! •.-.L^f.^
300 Yes
15,000 Yes
300 No
15,000 Yes
15,000 Yes
300 No
15,000 Tes
15,000 Yes
15,000 Yes
15,000 Yes
15,000 No
300 No
15,000 No
15,000 No
. 15,000 Yes
15,000 No
15,000 Yes
300 No
300 No
15,000 Yes
15,000 Yes

-------
00
                                                       Table 8-2




                   Deteetability of Radionuclides (Small Area) by Surface Scan* Relative to RG 1.86 Limits
«,
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
.,', Bete' i.
19,000 460,000
37,000 750,000
19,000 370,000
18,000 ND
23,000 470,000
77,000 390,000
1,100 5,600
840 3,500
2,100 13,000
550 2,300
2,200 12,000
580 2,700
400 2,200
15,000 120,000
17,000 160,000
16,000 180,000
1,200 10,000
680 3,000
120,000 600,000
!£r.*..:.
420 1,600
420 1,600
420 1,600
1.7E+07 6.5E+07
420 1,600
420 1,600
ND ND
ND ND
ND ND
ND ND
ND ND
85 320
85 320
420 1,600
420 1,600
420 1,600
430 1,600
420 1,600
ND ND
«.;!
8.5E+06
2.2E+06
8.7E+06
6.9E+09
1.1E+07
5.0E+06
370,000
I.OE+06
280,000
8.7E+09
ND
140,000
110,000
1.2E+07
1.5E+07
9.9E+06
160,000
1.2E+Q6
740,000
,. wiM '\ ',;=
MtHloiur8 . Detectable
300 No
300 No
300 No
300 No
300 No
300 No
300 No
15,000 Yes
15,000 Yes
3,000 Yes
15,000 Yes
300 Yes
300 Yes
300 No
3,000 Yes
15,000 Yes
15,000 Yes
15,000 Yes
15,000 No
              * Scan rate = 1/3 detector width per second for alpha & beta, 15 cm/s for gamma

-------
00
                                                       Table 8-3




                   Detectability of Radionuclides (Large Area)'by Surface Scan* Relative to-RG 1.86 Limits
f '• / , \f
' f\ '£ f> *
Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D
Pa-231
Pb-210+D
Pm-147
••^: YY 	 t( -, ;wt

160 790
1,200 7,300
1,500 25,000
1,400 9,800
230 890
8,400 193,000
830 4,100
640 3,000
480 '* 2,100
1,000 4,800
11,000 ND
1,300 9,500
1.4E+06 4.8E+06
16,000 ND
570 2,600
250,000 ND
3,100 ND
370 2,200
1,600 70,000
340 ' 1,800
1,200 6,900
)0ppaiil&cft^)f "'' ; 	 '."'
i':[M^toM
18 68
ND ND
90 340
ND ND
ND ND
89 330
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
92 350
91 340
89 330
ND ND
• ' ,, :.:' J
,, .'. . \. ' $ '
270
270
770
ND
'1,400
19,000
470
370
960
320
ND
730
820
2,800
410
ND
ND
360
2,000
9,000
,7.9E+06
: :P; TisjM;/ ?":
., ' ^ /' jy, >f*jf'f f^ f
-------
                                                      Table 8-3




                   Detectability of Radionuclides (Large Area) by Surface Scan* Relative to RG 1.86 Limits
oo
I
00
Ra4i<«8ieJidfe
•* "* <• :
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
- << - ttC
, Beta -
;>L'<3T*V ' '0M: -
6,800 145,000
13,000 2.4E+05
6,800 1.2E+05
6,400 ND
8,300 150,000
27,000 120,000
380 1,800
300 1,100
730 4,000
190 730
800 3,900
210 860
140 710
5,200 37,000
6,000 50,000
5,600 57,000
410 3,200
240 970
42,000 190,000
)Cpptn/iPcm2)
Alpha
' -OP * - m^
89 340
89 340
89 340
3.7E+06 1.4E+07
89 340
89 330
ND ND
ND ND
ND ND
ND ND
ND ND
18 67
18 67
89 340
89 340
89 340
91 340
89 340
ND ND
Qmfni'
t f m.l ;*
20,000
51,000
20,000
1.6E+07
24,000
12,000
860
2,400
650
2.0E+07
ND
310
250
28,000
34,000
23,000
360
2,800
1,700
	 wijr:": 	 r
Avemge "Peteatafjfe
' " Limit ' " -, , '
100 Yes
100 Yes
100 Yes
100 No
100 Yes
100 Yes
100 No
5,000 Yes
5,000 Yes
1,000 Yes
5,000 Yes
100 Yes
100 Yes
100 Yes
1,000 Yes
5,000 Yes
5,000 Yes
5,000 Yes
5,000 Yes
             * Meter time constant -10s

-------
                                                        Table 8-4
                          Detectability of Radionuclides by Direct Count* Relative to RG 1.86 Limits
00
.::

Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D -
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D
Pa-231
Pb-210+D
Pm-147
* '< ' ' •> > :-' ' ME
• * % '•»•' f S' j • f • ' "•"
70 260
560 2,600
690 8,000
620 3,200
100 290
3,800 63,000
370 1,300
290 990
210 680
460 ''1,600
4,900 ND
580 3,100
620,000 1.6E+06
7,400 ND
260 830
110,000 ND
1,400 ND
170 720
730 23,000
150 580
520 2,300
JC$p*/10(te«#} ' '
llfS:^.!
6 18
ND ND
32 91
ND ND
ND ND
32 90
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
33 93
33 92
32 90
ND ND
"", ' ?•;
"***i-
. 91
93
270
ND
480
6,400
160
130
330
110
ND
250
280
970
• 140
ND
ND
120
680
3,100
2.7E+06
\ / : R£Ht8£ ',''.'
J- Awage. ^ ; ^e&ctabie
100 Yes
5,000 Yes
100 Yes
5,000 Yes
5,000 Yes
100 Yes
5,000 Yes
5,000 Yes
5,000 Yes
5,000 Yes
5,000 Yes
100 No
5,000- Yes {
5,000 Yes
5,000 Yes
5,000 No
5,000 Yes
100 Yes
100 Yes
5,000 Yes
5,000 , Yes

-------
                                                       Table 8-4
                         Detectability of Radionuclides by Direct Count* Relative to RG 1.86 Limits
00
Mteaeiyy
• ". - !>;
1 ~ - . ' , ^j*
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
'Ml
Beta ' "<:*• '
t'.-.'W-. ';.'}?: ilWEv 'f
3,100 47,000
5,900 78,000
3,100 39,000
2,900 ND
3,700 49,000
12,000 41,000
170 580
130 360
330 * 1,300
87 240
360 1,300
93 280
63 230
2,300 12,000
2,700 16,000
2,500 18,000
190 1,000
110 320
19,000 63,000
J0(
-------
                               Table 8-5
Detectability of Radionuclides (Small Area) by Surface Scan* Relative to DCLs
..-..-*.. ;$,*.,.
;i&3kpi$i$&
'' • ' , ** '" '
*. X . . . . r .5 	 /. ..!.."!...!".
Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D '
Pa-231
Pb-210+D
Pm-147
Pu-238
Pu-239
ijf • r ^,'<"$P
;.: V-s&*'i ','*;*
•*m??±:L*MZ:<
440 2,500
3,300 23,000
4,300 78,000
3,900 31,000
650 2,800
24,000 610,000
2,300 13,000
1,800 9,600
1,300 6,600
2,900 15,000
31,000 ND
3,600 30,000
3.9E+06 1.5E+07
46,000 ND
1,600 8,000
700,000 ND
8,800 ND
1,100 7,000
4,600 220,000
960 5,600
3,300 22,000
19,000 460,000
37,000 750,000
a$pi$$fem?3 U'.'
£/J^\J.
ll^iLJlM 	
85 320
ND ND
420 1,600
ND ND
ND ND
420 1,600
ND ND
ND ND
ND ND
ND ND I-
ND ND
ND ND
ND . ND
ND ND
ND ND
ND ND
ND ND
440 1,600
430 1,600
420 1,600
ND ND
420 1,600
420 1,600
. - ... \
, *' ', T
ff ^8frB%
120,000
120,000
340,000
ND
600,000
8.1E+06
200,000
160,000
420,000
140,000
ND
320,000
360,000
1.2E+06
180,000
ND
ND
160,000
860,000
3.9E+06
3.5E+06
8.5E+06
2.2E+06
:> • ./• '-. -'( i'/ .<>' '
.'f "'., • < f, - , • * f
-* f&*. '•:.. . '.= '.{;.
^ , •>,- <>^ • '/•> •• -*;
7,300 Yes
19,000 Yes
9,700 Yes
1.4E+07 Yes
1.3E+06 Yes
17,000 Yes
13,000 Yes
48,000 Yes
130,000 " Yes
34,000 Yes
1.8E+09 Yes
15,000 Yes
58,000 No
2.1E+08 Yes
25,000 Yes
2.7E+09 Yes
1.1E+09 Yes
7,700 Yes
4,700 Yes
3,800 Yes
8.3E+07 Yes
17,000 Yes
16,000 Yes
- 1, . jok®MP^ >
J Jfo«< i '-5 ^ ''". ' \r •' '
:,/. ;^'-'^i ;\>fc *VV^:'?5
•:^LWJtMi%'*^M$$....2
490 Yes
1,200 No
640 Yes
900,000 Yes
89,000 Yes
1,200 Yes
870 No
3,200 Yes
8,800 Yes
2,300 No
1.2E+08 Yes
990 No
3,900 No
1.4E+07 Yes
1, 700 Yes
1.8E+08 Yes
7.3E+07 Yes
510 Yes
310 No
250 No
5.5E+06 'Yes
1,200 Yes
1,100 Yes
-. •- f ^ " A §.. ^'- .' \f ' -.'4
^•\ -.'^Jfifv^^ ^t^^.-y-^
49 No
120 No
64 No '
90,000 Yes
8,900 Yes
120 No
87 No
320 No
880 No
230 No
1.2E+07 Yes
99 No
390 No
1.4E+06 Yes
170 No
1.8E+07 Yes
7.3E+06 Yes
51 No
31 No
25 NO
550,000 Yes
120 No
110 No

-------
                                                                 Table 8-5
                                 Detectability of Radionuclides (Small Area) by Surface Scan* Relative to DCLs
*—
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
^TiJ
19,000 370,000
18,000 ND
23,000 470,000
77,000 390,000
1,100 5,600
840 3,500
2,100 13,000
550 2,300
2,200 12,000
580 2,700
400 2,200
15,000 120,000
17,000 160,000
16,000 180,000
1,200 10,000
680 3,000
120,000 600,000
3(dpm#00cm?3
420' 1,600
1.7E+07 6.5E+07
420 1,600
420 1,600
ND ND
ND ND
ND ND
ND ND
ND ND
85 ,', 320
85 320
420 1,600
420 1,600
420 1,600
430 1,600
420 1,600
ND ND
,<£.
8.7E+06
6.9E+06
1.1E+07
5.0E+06
370,000
l.OE+06
280,000
8.7E+09
ND
140,000
110,000
L2E+07
1.5E+07
9.9E+06
160,000
1.2E+06
740,000
= for
,16,000 Yes
l.OE+06 Yes
17,000 Yes
19,000 Yes
64,000 Yes
460,000 Yes
180,000 Yes
40,000 Yes
5.5E+08 Yes
44,000 Yes
13,000 Yes
18,000 Yes
4,100 Yes
37,000 Yes
36,000 Yes
41,000 Yes
120,000 Yes
•w '
1,100 Yes
67,000 Yes
1,100 Yes
1,300 Yes
4,300 Yes
30,000 Yes
12,000 Yes
2,600 Yes
3.6E+07 Yes
2,900 Yes
900 Yes
1,200 Yes
280 No
2,500 Yes
2,400 Yes
2,700 Yes
8,100 No
'W'^aa/^ »"''«''•
110 No
6,700 No
110 No
130 No
430 No
3,000 Yes
1,200 No
260 No
3.6E+06 Yes
290 Yes
90 Yes
120 No
28 No
250 No
240 No
270 No
810 No
00

tb
to
       * Scan rate = 1/3 detector width per second for beta & alpha, 15 cm per second for gamma

-------
                               Table 8-6
Detectability of Radionuclides (Large Area) by Surface Scan* Relative to DCLs
••• ':*• <•;«*/
k'f' ."' '•',
Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D
Pa-231
Pb-210+D
Pm-147
Pu-238
Pu-239
^'•''-: •.'-;,,';'., MB
£fr$M&3
160 790
1,200 7,300
1,500 25,000
1,400 9,800
230 890
8,400 193,000
830 4,100
640 3,000
480 2,100
1,000 4,800
11,000 ND
1,300 9,500
1.4E+06 4.8E+06
16,000 ND
570 2,600
250,000 ND
3,100 ND
370 2,200
1,600 70,000
340 1,800
1,200 6,900
6,800 150,000
13,000 240,000
.^/ „•/'.< '? •• :-,. .. f . "" y$
~ :• v j- ' £A rttfayi f
18 68
ND ND
90 340
ND ND
ND ND
89 330
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
92 350
91 340'
89 330
ND ND
89 340
89 340
' k :•'' ' '
$8$-
270
270
770
ND
1,400
19,000
470
370
960
320
ND,
730
820
2,800
410
ND
ND
360
2,000
9,000
7.9E+06
20,000
51,000
' " '{•''. '' ^ -' •"'
% ' '•• '. ...'./. '; '',.'; ' '

7,300 Yes
19,000 Yes
9,700 Yes
1.4E+07 Yes
1.3E+06 Yes
17,000 Yes
13,000 Yes
48,000 Yes
130,000 Yes
34,000 Yes
1.8E+09 Yes
15,000 Yes
58,000 Yes
2.1E+08 Yes
25,000 Yes
2.7E+09 Yes
1.1E+09 Yes
7,700 Yes
4,700 Yes
3,800 Yes
8.3E+07 Yes
17,000 Yes
16,000 Yes
^f)GL$pra#9Qrf)'; .
•••$&#—£•;
490 Yes
1,200 Yes
640 Yes
900,000 Yes
89,000 Yes
1,200 Yes
870 Yes
3,200 Yes
8,800 Yes
2,300 Yes
1.2E+08 Yes
990 Yes
3,900 Yes
1.4E+07 Yes
1, 700 Yes
1.8E+08 Yes
7.3E+07 Yes
510 Yes
310 Yes
250 Yes
5.5E+06 Yes
1,200 Yes
1,100 Yes
:; ^ :f f.v;>/ -^ >;j
;*•!. ferr. ^'V' ^': ;:.
49 Yes
120 No
64 No
90,000 Yes
8,900 Yes
120 Yes
87 No
320 No
880 Yes
230 No
1.2E+07 Yes
99 No
390 No
1.4E+06 Yes
170 No
1.8E+07 Yes
7.3E+06 Yes
51 No
31 No
25 No
550,000 Yes
120 Yes
110 Yes

-------
                                                              Table 8-6
                               Detectability of Radionuelides (Large Area) by Surface Scan* Relative to DCLs
s, "* '% ^v
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
; , ,.Beta ^r. *
6,800 120,000
6,400 ND
8,300 150,000
27,000 120,000
380 1,800
300 1,100
730 4,000
190 730
800 3,900
210 860
140 710
5,200 37,000
6,000 50,000
5,600 57,000
410 3,200
240 970
42,457 190,000
DCdfTO/lQffem1)
, ; Alpba , ,. ;;
GF . ZcS '
89 340
3.7E+06 1.4E+07
89 340
89 330
ND ND
ND ND
ND ND
ND ND
ND ND
18 ,, £7
18 67
89 340
89 340
89 340
91 340
89 340
ND ND
• mi \ ,
20,000
1.6E+07
24,000
12,000
860
2,400
650
2.0E+07
ND
310
-250
28,000
34,000
23,000
360
2,800
1,700
for ' ,, ,
. 15njr«tfi/y 'DWctaljJe .*
16,000 Yes
l.OE+06 Yes
17,000 Yes
19,000 Yes
64,000 Yes
460,000 Yes
180,000 Yes
40,000 Yes
5.5E+08 Yes
44,000 Yes
13,000 Yes
18,000 Yes
4,100 Yes
37,000 Yes
36,000 Yes
41,000 Yes
120,000 Yes
_DCL(dpm/lMte,.)
1,100 Yes
67,000 Yes
1,100 Yes
1,300 Yes
4,300 Yes
30,000 Yes
12,000 Yes
2,600 Yes
3.6E+07 Yes
2,900 Yes
900 Yes
1,200 Yes
280 Yes
2,500 Yes
2,400 Yes
2,700 Yes
8,100 Yes
ife£Sv iiLi-
110 Yes
6,700 Yes-
110 Yes
130 Yes
430 Yes
3,000 Yes
1,200 Yes
260 Yes
3.6E+06 Yes
290 Yes
90 Yes
120 Yes
28 No
250 Yes
240 Yes
270 Yes
810 No
oo
      * Meter time constant - 10 s

-------
                         Table 8-7
Detectability of Radionuclides by Direct Count* Relative to DCLs
/ ; '•-' '"':
HlS.. Xf'.'|-
Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni:63
Np-237+D
Pa-231
Pb-210+D
Pm-147
Pu-238
Pu-239
:T X -4" i1 _ SIP
\if.f'., ,'$&'•'?,'.$
.'/&x&'&£f
70 260
560 2,600
690 8,000
620 3,200
100 290
3,800 63,000
370 1,300
290 990
210 680
460 1,600
4,900 ND
580 3,100
620,000 1.6E+06
7,400 ND
260 830
110,000 ND
1,400 ND
170 720
730 23,000
150 580
520 2,300
3,100 47,000
5,900 78,000
£$i«0p^; • ,
-'"•; -r c^jtehix •-,',-•
f, ; -,,.- .ys& • < _ v-
U^:^ML:;
6 18
ND ND
32 91
ND ND
ND ND
32' 90
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
ND ND
33 93
33 92
32 90
ND ND
32 90
32 90
• %; t • ;
i$ai»8*a,'
•C«fe
91
93
270
ND
480
6,400
160
130
330
110
ND
250
280
970
140
ND
ND
120
680
3,100
2.7E+06
6,700
18,000
,. ••-:; j •- .. • .'4 -
r f "i. '", '- - , - > .- •
«'/..
' 'jl^«^y A b$$&W. 1
490 Yes
1,200 Yes
640 Yes
900,000 Yes
89,000 Yes
1,200 Yes
870 Yes
3,200 Yes
8,800 Yes
2,300 Yes
1.2E+08 Yes
990 Yes
3,900 Yes
1.4E+07 Yes
1, 700 Yes
1.8E+08 Yes
7.3E+07 Yes
510 Yes
310 Yes
250 Yes
5.5E+06 Yes
1,200 Yes
1,100 Yes
•> ' : ' >••/ • -f-- ' • ,
••,'f',.\ ^ • .\" ••' ,v.
.. •!•• ^Sj", v $ '••'' •>}'' J
, y. avs'"^* ''"?'/ •',?•'''?'/'
'$f\'i$jm{$i- M«dtall$-
49 Yes
120 Yes
64 Yes
90,000 Yes
8,900 Yes
120 Yes
87 No
320 Yes
880 Yes
230 Yes
1.2E+07 Yes
99 No
390 Yes
1.4E+06 Yes
170 Yes
1.8E+07 Yes
7.3E+06 Yes
51 Yes
31 No
25 No
550,000 Yes
120 Yes
110 Yes

-------
                                                              Table 8-7
                                     Deteetability of Radionuclides by Direct Count* Relative to DCLs
RadienaelWe
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106-i-D
Sb-I25+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
- • =WP
-x \l«a ;' -
-.*wk'. -,' 
-------
8.6   THE POTENTIAL IMPACT OF INTRODUCING RADIOACTTVELY
      CONTAMINATED METAL IN THE PRODUCTION OF STEEL

When surface contaminated scrap metal is introduced in the smelting process, the newly
produced steel can be expected to be volumetrically contaminated.  The potential impact of
surface contaminated scrap metal- on contamination levels in the metal melt can be estimated by
means of the following equation:
                  RSM  RSM
where:
 RSM
 RSM
0.45  =
                   concentration in metal melt (pCi/g),
                   surface contamination of scrap metal (dpm/1 00 cm2 ),
                   thickness of scrap metal (cm),
                   density of scrap metal (g/cm3),
                   mass fraction of contaminated scrap metal in new steel melt,
                   fraction of scrap metal radioactivity transferred to new steel melt,
                   conversion factor (pCi/dpm).
An upper bound estimate of volumetric contamination may be derived from the following
conservative assumptions:

  (1)   all scrap has residual surface contamination at the maximum acceptable release limit of
       5,000 dpm/1 00 cm2,

  (2)   the average thickness of scrap is 1 cm,

  (3)   density of scrap is 7.8 g/cm3.,

  (4)   newly produced steel melt is derived from 40% scrap (60% uncontaminated virgin
       material), and

  (5)   all surface radioactivity is transferred to new steel melt.
                                        8-27

-------
Substituting these conservative values into the above equation yields a bounding estimates for
volumetric contamination in new steel melt of 1.15 pCi/g. Radionuclides for which acceptable
release limits are 1,000 dpm/100 cm2 and 100 dpm/100 cm2 yield bounding values of 0.23 pCi/g
and 0.023 pCi/g, respectively.

8.7    LIMITATIONS OF STANDARD SURVEY MEASUREMENTS WHEN SCRAP IS
       CONTAMINATED VOLUMETRICALLY

When scrap metal is potentially contaminated volumetrically, the limits of detection by standard
release survey measurements are considerably more difficult and less sensitive.  From Equation
8-4, it can be shown that the impact of recycling contaminated scrap metal with a surface
contamination of 5,000 dpm/100 cm2 is equivalent to recycling scrap metal  with a volumetric
contamination of 2.9 pCi/g. Any attempt to establish parity between existing release criteria for
surface contamination and future criteria for volumetric contamination must, therefore, address
the issue of survey measurements of volumetric contaminants. In addition  to all factors which
define detection limits for surface contamination, limits of detection for bulk contamination are
further hampered by the short ranges of alpha and beta particles. Detection  is limited to radiation
emissions occurring at a depth that is less than the range of the particulate radiation.

8.7.1  Limitations When Bulk Contaminant is a Beta Emitter

In effect, such survey measurements are synonymous with counting samples of "infinite"
thickness for which absorption of radiation by the "sample" occurs. An estimate of activity level
per unit volume or weight, therefore, requires that the observed count rate be adjusted for sample
absorption and must further assume that contaminant is distributed uniformly throughout the
sample.  To account for self absorption when the contaminant is a beta emitter, the following
formula is used:


                                                                           (Eq'8"5)
where:

       Fs     =      self absorption factor,
                                        8-28

-------
       RO    —     measured activity,
       R     =     true activity,
       x     =     sample thickness (mg/cm2),
       UM    =     absorption coefficient (cm2/mg) (see NBS Handbook No. 51, pg 26).

Inspection of Equation 8-5 reveals that the self absorption factor is a function of the sample
thickness and UM:


                 (cmz/m   ~     0-693
                M            HVT(mgfcm2)                                   ^'

where HVT is the half-value thickness based on maximum bete energy.

The use of Equations 8-5 and 8-6 and the limits of sensitivity for volumetric contaminated
material can be illustrated by the following example.

Sample Calculation: Calculate the volumetric activity (i.e., pCi/g) of a 1 cm thick slab of steel
uniformly contaminated with the activation nuclide Co-60. A fixed-point surface measurement
with a pancake probe yields a ngt count rate of 50 cpm.

Solution:
  ff
  K. —
where:
                                    0.693        0.693
                            „   - 	 =
 RB=50cpm                M
                                         8-29

-------
               X =     "   .  7800 •,*/«,'  m 7800mgfcm
                    thickness        1 cm
               R  = *•   "  "    '  = 33,783 cpm

To convert count rate (cpm) to sample activity (dpm per unit volume or weight) assume that the
pancake probe has an effective area of 20 cm2 on contact and a 10% counting efficiency.

                           BulkActivity =         R
            n 77 A *•  -^                       33,783 cpm
            BulkActivity  =	——
                          (20cm z)(Q.2cpmldpm)(l cm)(7^g/cm 3~)(2.22dpmlpCi)

                                BulkActivity = 975 pCi/g

Note: The selected R,, value of 50 cpm approximates the lower limits of measurable activity for
the pancake probe in an environment where the background level is 100 cpm. Correspondingly,
the value of 975 pCi/g also approximates the MDC value for any material volumetrically
contaminated with Co-60 having a density thickness of 7,800 mg/cm2, which, for steel,
corresponds to a thickness of 1 cm.

8.7.2  Limitations When Bulk Contaminant is an Alpha Emitter

Owing to the much shorter range of alpha particles, any observed activity by standard alpha
survey instruments must be assumed to have originated within the first micron of metal. This
assumption can be approximated from the formula which defines the range (cm) of alpha
particles in air.

                                   RY=1.24E-2.62

where E is the alpha particle energy in MeV.
                                         8-30

-------
 An alpha particle of 5 MeV is estimated to have a range of about 3.6 cm in air having a density
 of 0.0013 g/cm3. For steel having a density of about 7.8 g/cm3, the range of a 5 MeV alpha
 particle would correspond to less than 1 micron. Assuming that an alpha particle would have to
 have a minimum residual energy of 2.5 MeV in order to trigger a pulse, it can be assumed that
 any detected alpha radiation had to originate within about 1 micron from the surface. For bulk
 contamination with a uniform contamination over a thickness of 1 cm yielding a net alpha count
 rate of 10 cpm, the volumetric contamination can be estimated to be greater than 580 pCi/g.

 Note: A net count rate of about 10 cpm with a ZnS detector operating at a 17% efficiency in a 3
 cpm background environment represents the lower limit of detectable activity. Thus, standard
 alpha survey techniques are quite insensitive to volumetrically distributed conatamination in
 steel.

 8.7.3  Limitations for Gamma Emitting Bulk Contaminants

 Gamma radiation is not subject to the intense self absorption that limits detection of particulate
 radiation under conditions of volumetric contamination. It would appear, therefore, that a Nal or
 GeLi detector might have a suitable application in quantifying volumetrically distributed gamma-
 emitting contaminants in scrap metal. Realistically, however, quantitative measurements for   v
 these detection systems can only be obtained under rigidly controlled conditions hi which the
 assessed sample is essentially identical to a standard calibration source(s) in terms of size, mass,
 configuration, counting geometry, and gamma energy emissions.

 8.8   ASSESSING THE RADIONUCLIDE CONCENTRATION IN STEEL PRODUCED
       FROM SCRAP

 The aforementioned difficulty and irapracticality of assessing volumetric contamination hi steel
 derived from recycled scrap are significantly reduced when such measurements are sought for the
• metal melt. This is due to the fact that the smelting process can be assumed to (1) convert all
 contaminants (surface and volumetric) into volumetric contaminants and (2) distribute
 contamination with relative uniformity hi the metal melt.  Consequently, only a limited number
 of samples are needed to characterize a large mass of metal melt (i.e., for a basic oxygen furnace,
 the mass per charge is 220 tons). Furthermore, samples may be obtained from the metal melt in
 the form of standard ingots-, which provide a suitable basis for developing calibration standards
                                          8-31

-------
for Nal and GeLi systems.  Sub-samples of ingots may also be analyzed by radiochemical
methods.

Radiochemical analysis requires the complete dissolution of the metal sample in acid(s), the
chemical separation of radioisotopes belonging to a common element, and the quantification of
radionuclides on the basis of their emission(s) by means of a suitable detection system.

8.9    MDCS AND ASSOCIATED PARAMETERS FOR LABORATORY ANALYSIS OF
       RADIONUCLIDES

The MDC for laboratory measurements is calculated using the following equation:
              MDC  =
(Eq. 8-7)
where:
             MDC   =   minimum detectable concentration (pCi/g)
             BR     =   detector background count rate (cpm)
             t       5=   count time (min)
             Yr      =   yield for emission I (ptcle/d)
             €j      =   detector efficiency for emission I (c/ptcle)
             M      =   sample mass (g)
             R      =   chemical yield.

For this analysis, MDCs for laboratory analysis of solid samples were obtained from an article by
P.M. Cox and C.F. Guenther (Cox 1995). The authors present a range of MDCs as reported by
24 commercial and government laboratories. The article is quite applicable to this analysis
because it presents state-of-the-art MDCs and associated parameter values for laboratory analysis
of radioactive materials in solids, as well as analysis costs as a function of MDC.

The MDC column contains information that integrates the effects of background levels, detection
efficiencies, count time, and sample size to give a lower limit of what could reasonably be
detected with a given level of confidence. The values given represent the state-of-the-art

                                         8-32

-------
detection capabilities as reported by commercial analytical laboratories. While increasing the
count tune or sample size can lead to the detection of lower concentrations, it should be realized
that there are practical limits on detection that are driven by tune constraints, background levels,
cost, and the desired level of confidence. For most radionuclides, reported background count
rates were below one count per minute and yielded corresponding lower limits of detecting
sample activity that were less than 1 pCi/g.

In most cases under "Mode of detection," the entry was either alpha, beta, or gamma. In several
cases, one could use two different modes of detection for the same isotope.  Examples, are 1-129
and Pb-210 where either beta or gamma analysis could be used. In the case of Th-232, both
alpha and gamma analyses are cited as acceptable. The following provides a brief overview of
instrumentation commonly employed by the reporting analytical laboratories:

  •     Gross alpha measurements were typically made using either scintillation devices (such as
       ZnS  in conjunction with a photomultiplier tube), gas-flow proportional detectors (both
       "windowless" and thin-window types), or thin-window GM tubes. Gas flow counters
       using P-10 gas (a 90/10% blend of argon/methane gases) were the most popular.

  •     For gross beta measurement, gas-flow proportional and thin-window GM tube-type
       detection systems were the most popular.  Some advanced liquid-scintillation detectors
       were also used for general alpha and beta detection.

  •     Both gross gamma and gamma spectrometric measurements were made using the popular
       and very effective sodium iodide [Na(Tl)] and germanium (Ge) detector systems.

Typical background rate, BR, includes the response of a nuclear detector to natural background
radiation.  Such background radiation can be either internal to the detector (as a result of natural
radioactive impurities in the detector materials) or external, such as terrestrial or cosmic that
penetrates into the collection region of the detector. Backgrounds of alpha radiation were
typically low, about 0.1 cpm, vary dfurnally with radon and progeny air concentrations and were
practically invariant with the system used.  Backgrounds of beta radiation varied from 10 to 30
cpm, again basically system invariant. Backgrounds of gamma radiation were totally the result
of external gamma and cosmic-ray backgrounds, and hence increased with altitude/elevation and
varied with building materials.
                                          8-33

-------
Count times, t, and sample mass, M, show the range of counting times and sample sizes
necessary to achieve the MDC ranges indicated. Note that count tunes range from 20 minutes to
almost 17 hours (1,000 minutes), while sample mass ranges from 0.1 to 750 g.

System detection efficiency, e, refers to the fraction of emitted radiation from the sample that
impinged upon a detector per unit time that was converted into a measurable signal, such as
counts per unit time. All variables such as geometry, intrinsic efficiency, etc., were taken into
account  For alpha and beta measurements, the efficiency can reasonably vary from 5% to 40%,
depending on detector type. Other important factors that can affect overall system efficiency
were the distance between the source (sample) and detector surface (particularly for alpha), and
sample area to detector area (surface area to surface area). For example, the best geometry for
detection of alpha and beta radiation was one in which the surface areas of both were the same.
For gamma rays, the efficiency was always low (less than a few %), and for specific energy
peaks representing specific isotopes, the value was even lower (<1%).

Chapter 7 of this Technical Support Document discusses the derivation of normalized doses for
individuals who may be exposed to radiation as a result of the recycling of scrap metal from
nuclear facilities.  Based upon these normalized doses, volumetric concentration limits (in pCi/g)
have been derived for annual doses of 15, 1, and 0.1 mrem.  Table 8-7 presents a comparison of
radionuclide MDCs by laboratory analysis with these derived concentration limits (DCLs) for
these three dose values. These results demonstrate the excellent sensitivity that can be achieved
when samples of the metal melt are obtained for laboratory analysis.  Laboratory MDCs are
adequate to detect all 40 radionuclides for the 15 and 1 mrem/y cases. For 0.1 mrem/y, only 6 of
the radionuclides could not be detected. In each of these cases, the MDC was only slightly
higher than the DCL. If deemed necessary, the laboratory could reach these DCLs by using a
longer count time for samples which must be analyzed for these radionuclides. Note however
(see Table 8-6) that the lowest MDCs for some of these radionuclides, particularly the alpha
emitters, is  already based upon count times ranging from 7 to 17 hours per sample.
                                          8-34

-------
                                                          Table 8-8



                                       Laboratory MDCs, Associated Parameters, and Costs
.];*',...'£*.'••' •• f '
QMkMify;
Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D
Pa-231
Pb-210+D
.•- ?$B&OV.
'•'*&$» i
0.05 - 0.4
0.04
0.05 - 0.5
0.2-37
0.23
6.05 - 0.1
0.01 - 0.3
0.02 - 0.4
0.007 - 0.3
0.02 - 0.9
1-30
0.4-2
0.2 - 0.3
0.2-150
0.2 - 0.3
1-30 ,
1-100
0.05 - 0.5
0.05 - 0.5
0.1-2
•\iMptetf t
'?!&&&&$& t
a
Y
a
P
Y
a
Y
Y
Y
Y
P/Y
P/Y
Y
P
Y
p/Y
P
a
a
a
>;?*• ^c-
• ' ; (tiptft) j ; *
0.03
1
0.008 - 0.01
8-21
1
o.dos - o.oi
1
0.2-1
0.2-1
0.08 - 1
2.5 - 20
0.35-2
1
6-191
1
2.5 - 20
20
0.01
0.01
0.01
;i;f;.ff; >
'/&*% -,
400 - 600
100 - 500
400 - 700
25 - 200
100 - 500
400 - 700
100 - 500
100-500
100-500
100-500
20-100
50-100
100 - 500
25 - 200
100 - 500
20 - 100
20-50
400 - 700
400 - 700
240 - 400
' , &/ • '
/• (&''
14-25
0.6-30
14-35
65-75
2.2
14-35
0.9 - 30
0.12-30
0.6-30
0.18-30
20-35
0.3 - 22
0.5
40-50
0.5
20-35
30-50
14-25
14-25
14-20
. *;;».-.•
; -, •# < •
1-100
500
1-10
0.1-50
500 '
1-10
100-500
100-500
100-175
100-500
1-50
500
100 - 500
0.1-50
100-500
1-50
0.1-50,
, 1-10
1-10
1-50
••>.., >^*%£
80-178
95 - 172
95 - 250
40
95-178
95 - 250
95 - 178
95 - 172
95 - 178
95 - 178
60 - 172
90 - 200
95 - 172
40-110
95-172
60-172
60 - 200
80 - 178
80- 178
70 - 200
$#$l^ty$
2-.i4j£%/
100-210
95 - 220
110-300
90
95 - 220
110-300
95 - 220
95 - 220
95 - 220
95 - 220
80 - 220
100 - 220
95 - 220
• 90-140
95 - 220
80 - 220
75 - 220
100-210
100-210
90 - 240
?fe6i--.'' • "/''
< •<" >* '', 1*.*
' W-V&&
135-270
105-260
145 - 375
100
105-260
145-375
105-260
105-260
105-260
105-260
110-260
125 - 260
105-260
100-170
105-260
110-260
1 10 - 240
135-270
135 - 270
120-300
00

-------
                                                           Table 8-8




                                       Laboratory MDCs, Associated Parameters, and Costs
Radkmueljde
Ptn-147
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
MPC
- ¥%&.
0.5-5
0.02 - 0.4
0.02 - 0.4
0.05 - 0.4
0.02 - 20
0.02 - 0.4
0.02 - 0.7
0.1-2
0.2-1
0.11
0.03 - 5
0.3 - 15
0.05 - 0.4
0.05 - 0.4
0.05 - 0.5
0.05 - 2 -
0.05 - 0.2
0.02 - 0.3
0.02-0.1
0.09 - 0.6
Mode of
detecdoft^
P
a
a
a
P
a
Y
Y
Y
Y
P
P
a
a
a
a/v
a
a
a
Y
BR
(«$»)
2-2.5
0.004-0.01
0.004 - 0.01
0.01
2-10
0.004 - 0.01
0.17-1
0.21 - 1
1
'!
0.5-1
2-30
0.03
0.03
0.01
0.01
0.01
0.004-0.01
0.01
1
t
(ifiift)
100
400 - 1,000
400 - 1,000
400 - 1,000
100-600
400 - 1,000
100 - 400
100-400
100 - 500
100-500
100
50-100
400 - 600
400 - 600
400 - 700
400 - 700
400 - 600
400-1,000
400-1,000
100-500
w
20-40
14-35
14-35
14-35
15-20
14-35
0.7-30
0.35-30
0.7-30
1-30
20-58
20-33
14-25
14-25
14-25
25
14-25
14-25
14-25
0.4-30
M
, CfiS
1-50
1-10
1-10
1-10
1-10
1-10
100-750
100-750
100 - 500
500
1-25
1-10
1-100
1-100
1-10
1-10
1-10
1-10
1-10
100-500
„.- Priee£
'Ifl#pQjfe
70 - 172
95 - 200
95 - 200
95-200
80 - 250
95 - 200
70 - 172
75 - 172
95 - 172
95 - 172
75 - 222
75 - 220
80 - 178
80 - 178
80 - 178
80 - 195
85 - 166
85 - 166
95 - 166
95 - 172
$/*«&{$) to &
- foflCSfe
95 - 220
110-200
110-200
110-200
90 - 300
110-200
80 - 220
90-220
95 - 220
95 - 220
95 - 165
95 - 240
100-210
100-210
100-210
100-210
95 - 200
95 - 200
110-200
95 - 220
ste# 4
cyteffe
120-260
145-300
145 - 300
145 - 300
110-375
145-300
105 - 260
115-260
105-260
105-260
125 - 225
120-300
135-270
135-270
135-270
135 - 270
135-250
135-250
145-250
105-260
00

-------
                                                        Table 8-9
                           Detectability of Radionuclides by Laboratory Analysis Relative to DCLs
00
'*• . t '•
'< ''{ •* :
'la*!$liP^
Ac-227+D
Ag-llOm+D
Am-241
C-14
Ce-144+D
Cm-244
Co-60
Cs-134
Cs-137+D
Eu-152
Fe-55
1-129
Mn-54
Mo-93
Nb-94
Ni-59
Ni-63
Np-237+D
Pa-231
Pb-210+D
'.'•':&$&''>
t::%®j&.t?>
0.05 - 0.4
0.04
0.05 - 0.5
0.2 - 37
0.23
0.05-0.1
0.01-0.3
0.02 - 0.4
0.007 - 0.3
0.02 - 0.9
1-30
0.4-2
0^2 - 0.3
0.2-150
0.2 - 0.3
1-30
1-100
0.05 - 0.5
0.05 - 0.5
0.1 - 2
'^ 	 ;.:v: 	 s •v-
l'imf«m/y, . •• Efe^ai|b--
1.9 Yes
24 Yes
12 Yes
17,000 Yes
850 Yes
22 Yes
17 Yes
61 Yes
170 Yes
44 Yes
2.2E+06 Yes
19 Yes
74 Yes
270,000 Yes
32 Yes
3.4E+06 Yes
1.4E+06 Yes
9.8 Yes
6.0 Yes
4.9 Yes
\t .. ' BCB
-------
                                                       Table 8-9
                           Detectability of Radionuclides by Laboratory Analysis Relative to DCLs
00
w
00
'^MiomiMt \
Pm-147
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Ra-228+D
Ru-106+D
Sb-125+D
Sr-90+D
Tc-99
Th-228+D
Th-229+D
Th-230
Th-232
U-234
U-235+D
U-238+D
Zn-65
' MDC
' Crt)'
0.5-5
0.02 - 0.4
0.02 - 0.4
0.05 - 0.4
0.02 - 20
0.02 - 0.4
0.02 - 0.7
0.1-2
0.2-1
0.11
0.03 - 5
0.3-15
0.05 - 0.4
0.05 - 0.4
0.05 - 0.5
0.05 - 2
0.05 - 0.2
0.02 - 0.3
0.02-0.1
0.09 - 0.6
• iSmrem/y Dfeteetatjle
110,000 Yes
22 Yes
21 Yes
21 Yes
1,300 Yes
22 Yes
24 Yes
41 Yes
290 Yes
240 Yes
25 Yes
700,000 Yes
11 Yes
3.4 Yes
23 Yes
5.3 Yes
48 Yes
46 Yes
52 Yes
160 Yes
DCL(per/$
lmrem/y Detefctabb
7,000 Yes
1.5 Yes
1.4 Yes
1.4 Yes
85 Yes
1.4 Yes
1.6 Yes
2.7 Yes
19 Yes
16 Yes
1.7 Yes
46,000 Yes
0.74 Yes
0.23 Yes
1.6 Yes
0.35 Yes
3.2 Yes .
3.1 Yes
3.5 Yes
10 Yes
0. f smmty Detectable :
700 Yes
0.15 Yes
0.14 Yes
0.14 Yes
8.5 Yes
0.14 Yes
0.16 Yes
0.27 Yes
1.9 Yes
1.6 . Yes
0.17 Yes
4,600 Yes
0.074 Yes
0.023 No
0.16 Yes
0.035 No
0.32 Yes
0.31 Yes
0.35 Yes
1.0 Yes

-------
8.10   SUMMARY  >

Standard scan survey techniques for small areas of contamination are only marginally adequate
relative to Regulatory Guide 1.86 limits, detecting slightly more than one-half of the
radionuclides. Only those radionuclides listed in group 4 of Regulatory Guide 1.86 (with a limit
of 15,000 dpm/100 cm2) can be reliability detected. Significant improvement can be obtained
when surveying for distributed sources of contamination (almost 90% of the radionuclides
detected), even detecting radionuclides as low as 100 dpm/100 cm2.

Relative to the 15 mrem/y DCL, small areas of contamination can be detected for all but one
radionuclide. At 1 mrem/y, deteptability drops to 75%, while only 25% of the radionuclides are
detectable at 0.1 mrem/y. Significant improvement is noted when surveying for distributed
sources of contamination. One hundred percent of the radionuclides are detectable at both 15
mrem/y and 1 mrem/y. At 0.1 mrem/y, almost 70% of the radionuclides are detectable while
scanning for large areas, while almost 90% could be detected using direct measurements.

These results should be considered to represent optimal monitoring conditions. Under field
conditions, levels of detectability would increase, perhaps due to a number of factors most of
which involve a loss of counting efficiency. An increase of a factor of 10 is possible above the
MDCs presented in these tables. Such an increase would have a significant effect  on
detectability, particularly for small areas of contamination.  The extent of the loss of detectability
is less for scanning for large areas of contamination and for direct measurements.  Surrogate
methods may be useful in situations where there are multiple radionuclides present.

Any assessment of volumetrically contaminated metal by standard field survey techniques is
severely restricted by the limited range of particulate radiation. Only those radionuclides with
DCLs greater than a few hundred pCi/g could be detected reliably. However, laboratory analysis
of samples of scrap steel or steel derived from the recycling of scrap metal provides significantly
improved results. State-of-the-art laboratory methods are quite effective at detecting low levels
of volumetric contamination, even down to a few tenths to even hundredths of a pCi/g.  At levels
corresponding to 15 mrem/yr and 1 mrem/yr, 100% of the radionuclides evaluated could be
detected. Even at 0.1 mrem/yr, 85% of the radionuclides are detectable.
                                          8-39

-------
8.11   REFERENCES

Cox, F.M. and C.S. Guenther.  1995. An Industry Survey of Current Lower Limits of Detection
for Various Radionuclides. Health Physics 69,121-129.

U.S. Nuclear Regulatory Commission. 1974. Termination of Operating Licenses for Nuclear
Reactors. NRC Regulatory Guide 1.86, Washington, D.C.

U.S. Nuclear Regulatory Commission. 1982. Termination of Operating Licenses for Nuclear
Reactors. NRC Regulatory Guide 1.86 (Draft), Washington, D.C.
                                         8-40

-------
                                     CHAPTER 9
                   NORMALIZED COLLECTIVE IMPACTS MODELS
This chapter describes the development of total normalized collective doses and risks for each of
the radionuclides considered. These normalized doses and risks can then be used with estimates
of the mass of metal to be recycled, and its radionuclide concentrations, to calculate the total
collective dose and risk received by humans due to the recycle of scrap metal from nuclear
facilities. The environmental pathways that are major contributors to the collective dose and risk
are described in this chapter and exposure models are developed. Figure 9-1 presents a
simplified schematic diagram of some of the potential exposure scenarios that could result from
recycling scrap metal from nuclear facilities into consumer and/or industrial products. In
general, the exposure scenarios can be broken into five categories: air emissions, slag, baghouse
dust, steel (these four categories are associated with the melting process) and transportation.
Figure 9-1: Potential Collective Exposure Scenarios

                                         9-1

-------
Figure 9-2 shows the approach that was used to develop the total normalized collective doses and
risks. As shown on the right side of Figure 9-2, for each of the four resulting media, one or more
initial re-use and/or final disposition scenarios were postulated. Normalized collective doses
were determined for each of these scenarios on the basis of unit activity.  These are referred to as
the unweighted scenario normalized collective doses. Next, the unweighted scenario normalized
collective doses for each media type were combined with the expected usage factor for each
scenario to arrive at media specific normalized collective doses. The calculation of these
normalized collective doses is presented in Sections 9.2 through 9.5.  Finally, as presented in
Section 9.6, the media specific normalized collective doses were combined with the radio-
nuclide specific partitioning factors (a few radionuclides partition to more than one of the media
shown on Figure 9-2; see Table 6-3) and the transportation normalized collective doses to
determine the radionuclide specific total normalized collective doses and risks. (See Table 9-15.)

Scrap

Pi
0

artitioning
able 6-3)
I.C
Mn, Sr, Mb,
Ce, U, Np,
Pu, Am, Cm
Cs, Pb, Zn
Co, Mn, Fe,
Ni,Zn.Tc, Ru
MM •* not Included ki model
W - value varried (see text)

Media
' '; XJr !
t 'Vh 1 1 T ' '

::"**:;-,.

1 Baghoxiss '•

naftM i |
..'- ^ .iiii

Usage
1.0




0246
0145
0.349
0.023
0193
Others
Zn
0.474
0058
0365
0103

Initial Re-Use
t HrstPass

SoilCofldiSonet
ftdad Surfade ',
ftoad Sase
" SSgalfasi •
,(Unc«!fij
: Ha?anJaus ;
. ftufis Bodfefe I
.... . ••!.••
i AssQ?iifl wleS
r^)>){2tl3ClC6S
fy£fipto £ttfif!t l!
Noft-Aceessabfe

Fate
10
00








1
1.0
NM
W

W
Final
Disposition
AS™ |

Mqntdpa!



• • F^J(50fi I
Wvftf) !

Mufijcfpal -
i laQEfiKnsI

S6c(j(f3v 9J-¥

Figure 9-2: Collective Impact Calculational Approach
                                           9-2

-------
The discussion in the following sections focuses on doses and dose models; however, the
potential cancer risks from the recycling of scrap metal from nuclear facilities were also
determined. For external exposures, cancer risks were determined by multiplying the calculated
dose by 7.6x10"4 total cancers per rem and 5.1 xlO"4 fatal cancers per rem dose-to-risk factors, as
recommended by EPA 94. For internal exposure, cancer risks were determined using the same
models presented below for calculating doses, except that slope factors from EPA 94a were used
in place of dose conversion factors.


9.1     TRANSPORTATION

Most of the scrap material potentially recycled from nuclear facilities—either the scrap metal
itself or the products and by-products of melting—would be transported by truck or by rail. The
method used to estimate the radiological impacts on the population residing along the
transportation routes from shipments of contaminated material is the same as that used by
RADTRAN and IMPACTS-BRC (O'Neal 90).  The following equation was used to calculate the
collective dose to the general population from external exposure to a single shipment of a
material:
                      A,,'4''"  "*•/*/*
                                          x.    0
   A,, =  collective dose from exposure to radionuclide i during shipment t (person-rem)
   P  =  population density (rrr2)
   L  =  length of trip (mi)
   Kj  =  source strength of radionuclide z(mrem-m2/hr)
    V =  speed of truck
       =  40mi/hr
    xm =  distance to nearest receptor
       =  30 ft (9. 14m)
   |aa  =  linear attenuation coefficient of air (nr1)
    x  =  distance from roadway to receptor
   y  =  distance along roadway from truck to receptor
   B(u^r), the Berger buildup factor for air, is represented by the following expression:
                                  O = 1  + a^re**'                            (2)

       =  first Berger buildup coefficient for gamma radiation of energy E in air
       =  second Berger buildup coefficient for gamma radiation of energy E in air
                                          9-3

-------
The linear attenuation coefficients in air are calculated from the mass attenuation coefficients for
air presented in the Radiological Health Handbook. The buildup coefficients were taken from
Table A4.9 of the Principles of Radiation Shielding, A. B. Chilton, et al, 1984.

The factor of 4-10"3 accounts for integrating the radiation field of the truck in both directions
(both approaching and receding from the receptor point) and over the population on both sides of
the road, as well as converting from millirem to rem.

The source strength, Kj, is the dose rate at a point one meter from an equivalent point source at
the center of the truck. The source strength was evaluated by calculating the dose rate at a point
100 meters outside the truck without any attenuation by the intervening ah* and extrapolating it to
a point one meter from the center.

In this analysis, six transportation segments were examined, as shown in Table 9-1. Because
most scrap generators, scrap yards, steel mills, etc. are located relatively close together, 100
miles was selected as a representative maximum transport distance. It was assumed that finished
products could be shipped anywhere hi the country; therefore, a longer transport distance of 1000
miles was assumed. For all transportation segments (as well as for all collective impact analyses
in this chapter) a representative population density of 100 people/km2 was assumed.

                   Table 9-1: Major Transportation Pathway Assumptions
Material
Scrap
Scrap
Steel
Steel
Baghouse Dust
Slag
Source :
Generator Site
Scrap Yard
Steel Mill
User (Recycle)
Steel Mill
Steel Mill
. Destination
Scrap Yard
Steel Mm"*
User
Scrap Yard
Processor
User
Distance
100 mi
100 mi
1000 mi
100 mi
100 mi
100 mi
Population Density
100 people/km2
100 people/km2
100 people/km2
100 people/km2
100 people/km2
100 people/km2
Normalized collective doses from the transportation of scrap metals recycled from nuclear
facilities and then* products and by-products are shown hi Table 9-2.
                                           9-4

-------
Table 9-2: Unweighted Transportation Doses
     (person-rem per Ci—transported)
Nuclide
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Mo-93
Nb-94
Tc-99
Ru-106+D
Ag-llOm
Sb-125
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
/ ' Sejeap.
O.OOE+00
2.42E-03
O.OOE+00
7.68E-03
O.OOE+00
O.OOE+00
1.75E-03
O.OOE+00
6.67E-08
4.50E-03
1.74E-10
5.60E-04
7.98E-03
1.06E-03
8.19E-07
4.35E-03
1.55E-03
1.09E-04
2.23E-09
3.16E-03
1.81E-07
5.04E-03
2.60E-03
7.09E-04
4.39E-03
5.58E-04
1.78E-07
5.42E-08
5.30E-05
4.28E-08
: , JStedU,
O.OOE+00
1.77E-02
O.OOE+00
5.62E-02
O.OOE+00
O.OOE+00
1.28E-02
O.OOE+00
4.98E-07
3.29E-02
1.28E-09
4.09E-03
5.83E-02
7.77E-03
6.03E-06
3.18E-02
1.14E-02
7.95E-04
1.64E-08
2.33E-02 ""•
1.33E-06
3.70E-02
1.92E-02
5.14E-03
3.24E-02
4.20E-03
1.31E-06
3.98E-07
3.88E-04
3.I5E-07
, ^kg.
O.OOE+00
3.50E-04
O.OOE+00
1.07E-03
O.OOE+00
O.OOE+00
2.49E-04
O.OOE+00
2.05E-08
6.54E-04
6.00E-11
8.34E-05
1.15E-03
1.64E-04
2.86E-07
6.38E-04
2.29E-04
1.71E-05
6.64E-10
4.46E-04
6.06E-08
7.25E-04
3.69E-04
1.28E-04
6.32E-04
7.96E-05
4.84E-08
1.71E-08
9.54E-06
1.33E-08
Dust
O.OOE+00
6.91E-03
O.OOE+00
2.14E-02
O.OOE+00
O.OOE+00
1.25E-03
O.OOE+00
6.41E-08'
1.29E-02
6.45E-10
1.63E-03
5.71E-03
3.14E-03
8.93E-07
1.26E-02
4.48E-03
7.92E-05
8.09E-09
8.63E-03
7.38E-07
1.42E-02
7.17E-03
2.28E-03
1.22E-02
1.47E-03
6.54E-07
2.07E-07
1.67E-04
1.61E-07
                   9-5

-------
                       Table 9-2: Unweighted Transportation Doses
                            (person-rem per Ci—transported)
Nuclide
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-241
Cm-244
Scrap "
1.79E-04
5.10E-05
3.83E-04
8.67E-09
3.24E-08
8.53E-09
2.79E-09
7.48E-09
2.97E-06
7.56E-09
Steel
1.32E-03
3.93E-04
2.81E-03
6.44E-08
2.38E-07
6.33E-08
2.04E-08
5.55E-08
2.19E-05
5.62E-08
Slag
4.11E-05
5.68E-06
6.96E-05
2.83E-09
l.OOE-08
2.80E-09
6.49E-10
2.47E-09
1.11E-06
2.41E-09
t>ust
6.08E-04
1.11E-04
1.19E-03
8.69E-09
1.18E-07
8.59E-09
2.41E-09
7.56E-09
1.28E-05'
7.45E-09
9.2    AIRBORNE EMISSIONS

Some fraction of each radionuclide contaminating the scrap metal could be released to the
atmosphere when the scrap is melted and converted to steel.  Although, as shown in Figure 9-2,
only iodine (1-129) and carbon (C-14) have airborne emissions as their primary partitioning
medium, other radionuclides could be released in airborne emissions as fugitive dust. Therefore,
airborne emission normalized collective doses were calculated for all 40 radionuclides being
evaluated.

The radiological impacts of these releases were assessed using the EPA's  CAP88-PC computer
code, and CU-POP (EPA 94b), a model developed for use in EPA's assessment of the collective
impacts of radioactively contaminated soils. CAP88-PC was used to calculate the impacts from
the first pass of released radionuclides, while the CU-POP model was used to evaluate long-term
(i.e., 1,000 year) impacts from the ground deposition of airborne emissions for all radionuclides
except C-14 and 1-129.

The CAP88-PC computer program is a desk-top computer version of the earlier CAP-88 code,
used to perform dose and risk assessments for the purpose of demonstrating compliance with 40
CFR 61.93(a). CAP88-PC embodies the CAP-88 methodology for radiological assessments of
                                         9-6

-------
both populations and maximally exposed individuals. The program uses a modified Gaussian
plume equation to estimate the average dispersion of radionuclides and rates of deposition on the
ground. The radionuclide concentrations in produce, leafy vegetables, milk and meat consumed
by humans are estimated by coupling the output of the atmospheric transport models with the
U.S. NRC Regulatory Guide 1.109 terrestrial food chain models. Assessments are done for a
circular grid within a radius of 80 kilometers (50 miles) around the facility. The mathematical
models and equations used in CAP88-PC are presented in the CAP88-PC user's guide.

Time integrated collective doses were calculated from atmospheric data that were provided with
CAP88-PC for seven sites located around the country, along with a constant population density
of 100 people/km2 and the default CAP88-PC usage factors. The seven sites are: Providence, RI;
Harrisburg, PA; Wilmington, DE; Chicago, IL; Los Angeles, CA; Knoxville, TN;  and Moline,
IL. These sites were selected because of then- proximity to nuclear facilities (usually more than
one), and because they represent most areas of the country (two each in the Northeast and Mid-
West, and one each in the South, Mid-Atlantic, and West).

Table 9-3 presents the results of the CAP88-PC calculation of normalized collective doses due to
the initial release of radionuclides into the air. The average collective dose factor from all the
sites was used as the basis for this analysis. However, it should be noted that the doses do not
vary significantly from one site to another, largely because the population density  was kept
constant.
        Table 9-3: Unweighted Airborne Emission Doses (person-rem per Ci—released)
Nuclide
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
First Pass (CAP88-PC) ;
Maximum
2.54E-01
1.04E+00
3.76E-02
5.59E+00
1.29E-02
3.26E-02
3.25E+00
5.82E+00
7.95E+00
7.80E-03
Minirmiso,
1.21E-01
7.22E-01
2.65E-0'2
3.88E+00
8.98E-03
2.28E-02
2.31E+00
4.10E+00
5.54E+00
5.42E-03
Average ;
1.81E-01
8.64E-01
3.16E-02
4.66E+00
1.08E-02
2.74E-02
2.73E+00
4.89E+00
6.65E+00
6.48E-03
Soil Deposition (CU-P'OF)
Maximara
NC
9.79E-01
5.03E-04
2.36E+01
1.90E-01
1.19E+00
3.63E+00
4.84E+02
1.46E+03
2.27E+01
Mttli*™*™

NC
9.30E-01
4.99E-04
1.68E+01
1.49E-01
7.81E-01
3.29E+00
1.67E+02
3.46E+00
4.61E-01
Average'
6.48E+01
9.57E-01
5.01E-04
1.97E+01
1.65E-01
9.43E-01
3.37E+00
2.51E+02
3.14E+02
5.41E+00
                                          9-7

-------
Table 9-3: Unweighted Airborne Emission Doses (person-rem per Ci—released)
Nuclide
Tc-99
Ru-106+D
Ag-llOm
Sb-125
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-241
Cm-244
PifStPassCCAW-SC)
Maximum?
2.23E-01
2.89E+00
4.53E+00
7.23E-01
2.63E+01
8.23E+00
4.56E+00
2.25E+00
2.11E-01
2.72E+00
4.80E+02
1.14E+02
4.99E+01
3.24E+03
1.19E+03
3.29E+03
1.16E+03
1.67E+03
2.34E+03
6.34E+02
5.88E+02
5.64E+02
2.34E+03
1.55E+03
1.67E+03
1.67E+03
2.62E+03
L59E+03
2.59E+03
1.37E+03
iVflmmnm •
1.57E-01
1.96E+00
3.16E+00
5.02E-01
2.00E+01
5.82E+00
3.25E+00
1.52E+00
1.42E-01
1.86E+00
2.71E+02
6.62E+01
2.82E+01
2.13E+03
7.90E+02
2.20E+03
7.74E+02
1.11E+03
1.56E+03
4.23E+02
3.92E+02
3.77E+02
1.56E+03
1.03E+03
1.11E+03
1.11E+03
1.75E+03
1.06E+03
1.73E+03
9.13E+02
. Average
1.87E-01
2.41E+00
3.78E+00
6.02E-01
2.33E+01
6.90E+00
3.84E+00
1.88E+00
1.76E-01
2.26E+00
3.66E+02
8.80E+01
3.81E+01
2.67E+03
9.87E+02
2.74E+03
9.66E+02
1.39E+03
1.95E+03
5.27E+02^
4.89E+02
4.70E+02
1.95E+03
1.29E+03
1.39E+03
1.39E+03
2.17E+03
1.32E+03
2.15E+03
1.14E+03
Soil Depo$it|.ob (CU^OF)
Maximom
4.27E+02
5.78E-01
3.56E+00
8.64E-02
NC
5.78E+00
3.74E+01
1.47E-01
4.01E-04
1.89E+01
4.93E+02
1.26E+04
2.16E+01
8.04E+01
3.55E+00
6.63E+02
3.25E+03
2.92E+03
5.31E+04
1.75E+03
1.74E+03
1.61E+03
3.19E+04
1.26E+01
1.08E+02
1.04E+02
3.14E-01
1.04E+02
1.40E+02
1.30E+00
Minimum
2.58E-01
5.03E-01
3.17E+00
7.80E-02
NC
5.67E+00
3.03E+01
1.46E-01
3.96E-04
1.61E+01
4.29E+02
8.24E+03
2.10E+01
6.84E+01
3.55E+00
4.95E+02
2.36E+03
2.16E+03
1.72E+02
2.52E+01
2.39E+01
2.20E+01
1.30E+02
9.30E+00
4.22E+01
4.11E+01
2.91E-01
4.07E+01
1.19E+02
1.29E+00
Average
2.34E+01
5.28E-01
3.28E+00
8.02E-02
1.14E+02
5.73E+00
3.38E+01
1.46E-01
3.98E-04
1.75E+01
4.50E+02
1.03E+04
2.13E+01
7.44E+01
3.55E+00
5.80E+02
2.80E+03
2.54E+03
4.11E+03
6.64E+02
6.43E+02
6.16E+02
1.81E+04
1.09E+01
6.85E+01
6.62E+01
3.03E-01
6.61E+01
1.30E+02
1.29E+00
                                  9-8

-------
A brief mathematical description of CU-POP is found in Exhibit 9-A.  A more detailed
description is found in Sections 2.2.5 and E.I of the TSD for clean up radiologically
contaminated soil (EPA 94b). In EPA 94b, a Generic Site and 27 Reference Sites were defined
and evaluated. As with the sites selected for the CAP88-PC analysis, the 28 sites evaluated with
CU-POP represent most areas of the country. Also, a constant population density of 100
people/km2 and the EPA 94b usage factors were used.  Again, Table 9-3 presents the results of
the GU-POP calculation of normalized collective doses due to the ground deposition of '
radionuclides released into the air. The .average collective dose from all 28 sites is used as the
basis for the collective dose factor. The total normalized collective doses from the air emission
of radionuclides is the sum of the initial release and the subsequent ground deposition
normalized collective doses.

For C-14, a global normalized dose of 399 person-rem/Ci over 10,000  years was taken from
Section 9.2.6 of NRC 95 and adjusted to the 1,000 year evaluation period being used in this
analysis—giving a value of 64.8 person-rem/Ci.  For 1-129, a global normalized dose of 3,800
rem/Ci to the thyroid over a 1,000 year period is reported in Section 4.8.2 of NCRP  83. This
value was converted to  an effective dose equivalent by multiplying by  0.03—giving a value of
114 person-rem/Ci.

9.3    SLAG

Slag—material that is a by-product of steel production—accumulates large fractions of many of
the radionuclides considered in the present analysis. Consequently, exposure to the  slag will
account for most of the collective impacts of these nuclides. Slag is usually sold by the steel
manufacturer to slag processing companies that in turn re-sell it for a wide range of various uses,
including road building, fill material, railroad ballast, soil conditioning, ice control,  etc. (see
Figure 9-1). The following sub-sections develop exposure scenarios for most of these slag uses,
and unweighted normalized collective doses are calculated.
                                                                                   i
Alternatively, if there is no slag market, the steel manufacturer will stockpile the slag on-site
until the slag market returns. In this case, collective exposures during the operation of the steel
plant would result from the leaching of contamination from the slag and the off-site  consumption
of ground water. If a slag pile remains following the closure of the steel plant, collective
exposures could result from people living or farming on the abandoned slag piles. Since the
                                           9-9

-------
current slag market is robust and virtually all of the slag generated is sold for re-use, unweighted
normalized collective doses have not been calculated for this stockpiling scenario.

9.3.1  Road-Building

A significant portion of the slag is currently used in road construction, as described below.
However, it should be pointed out that in the future this may not be the case. There are problems
with expansion from hydration of the slags which causes alligator cracking of surfaces and
buckling of substructures and significantly reduces the road's life span.  Also, there are some
locations (e.g., Maryland) where leachates from the slag have a very high pH and could be
considered to be too hazardous to use.  Nonetheless, since this analysis is based on current
practices, the use of slag in road building was evaluated. (See Chapter 10 for the effect of
eliminating this use of slag.)

Road Base

Slag from electric arc furnaces is currently used as the under-layment in road construction. The
information in this paragraph is based on a survey of the Federal and seven state departments of
transportation. The design life of highways is normally 35-40 years, although primary state
highways can be left in place for closer to 60 years. However, highways are seldom simply
abandoned at the end of their design life. They are repaired, resurfaced, and/or enlarged and used
indefinitely. Furthermore, most states try to recycle all materials used in road
building—concrete is crushed and used as aggregate in highway bases, wrdle old asphalt is
mixed with new in proportions as high as 50-60%. Materials not used are stockpiled, and if still
not used, are given to lower levels of government for use on local roads.

Therefore, it was assumed that slag used hi road base construction would remain within the road
base for the entire 1,000 year evaluation period. During this time, strong Y-emitting nuclides in
this slag expose the motoring public to external radiation.

Assessment of External Exposure

The external exposures of members of the general population traveling on roads built with slag
that is a by-product of recycling scrap metal from nuclear facilities are assessed in the following
manner:

                                          9-10

-------
                                  103F  (x)/. F
                             As N       IX^- * JI  O
                             I VI —	
                             IJt J —	    "•  	
                                       md
   A,dx(x)     =  normalized cumulative collective dose to occupants of vehicles from external
                 exposure to nuclide i in the road at distance x (person-rem per Ci in scrap)
   Fix(x)  =   road b386 dose conversion factor (mrem/hr per pCi/g)
   f,      =   fraction of radionuciide in scrap which partitions to slag
   F0     =   roadway occupancy factor (see Equation 4)
          =   20.56 (hrnvV)
   nij    =   mass of slag in road per unit length (g/m)
   t,.      =   evaluation tune
          =   l.OOOy

The roadway occupancy factor is callculated from the total highway travel in the U.S., as reported
by the Federal Highway Administration; the total length of U.S. highways, measured along the
centerline of each road (referred to as "racetrack miles"); and an assumed average speed of
passenger vehicles.
                                      F. = —p-
D
                                        (4)
                                            L v
   Dp =  total highway travel in U.S.
       =  3.87653 x 10,12 person-mi/y
   Lr  =  total length of U.S. highways
       =  3,904,721 "racetrack" miles
   v , =  Average speed of passenger vehicles
       -  30mphxl,609m/mi
       =  48,280 m/hr

The road base dose conversion factor [Fu/x)] was calculated with the MicroSbield™ computer
code, and includes the shielding effect of the road pavement. The normalized collective doses
from the use of slag as road base are presented in Table 9-4.
                                          9-11

-------
Table 9-4: Slag Pathway Normalized Collective Doses (person-rem per Ci—in
                              pathway)
Nuclide
C-14
Mh-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
Ag-llOm
Sb-125
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
Soil
O.OOe+00
9.57e-01
5.01e-04
1.97e+01
1.65e-01
9.43e-01
3.37e+00
2.51e+02
3.14e+02
5.41e+00
2.34e+01
5.28e-01
3.28e+00
8.02e-02
1.41e+03
5.73e+00
3.38e+01
1.46e-01
3.98e-04
1.75e+01
4.50e+02
1.03e+04
2.13e+01
7.44e+01.
3.55e+00
5.80e+02
2.80e+03
2.54e+03
4.11e+03
6.64e+02
Concrete
3.31e-03
1.65e+00
O.OOe+00
3.21e+01
O.OOe+00
O.OOe+00
9.32e-01
O.OOe+00
2.48e+03
1.396-01
3.28e-02
5.01e-01
4.41 e+00
1.06e-01
3.37e+00
7.32e+00
3.85e+01
9.45e-02
4.95e-05
3.59e+01
5.11e-02
2.36e+03
3.48e+01
1.65e+01
7.31e+00
3.98e+02
5.48e+02
4.17e+03
4.91e+01
l.Sle+00
Road Base:
O.OOe+00
4.40e-02
O.OOe+00
1.61e+00
O.OOe+00
O.OOe+00
3.94e-02
O.OOe+00
6.086+01
3.78e-24
8.74e-10
8.84e-03
1.46e-01
1.21e-03
3.50e-20
1.55e-01
6.68e-01
4.38e-03
6.01e-10
1.34e+00
1.09e-05
1.13e+02
2.17e+00
9.25e-02
5.83e-01
5.81e+00
2.61e+01
2.60e+02
9.99e-02
8.14e-02
RR Ballast
2.54e-03
1.27e+00
O.OOe+00
2.46e+01
O.OOe+00
O.OOe+00
7.15e-01
O.OOe+00
1.90e+03
1.07e-01
2.51e-02
3.85e-01
3.38e+00
8.16e-02
2.58e+00
5.62e+00
2.95e+01
7.24e-02
3.80e-05
2.75e+01
3.92e-02
1.81e+03
2.67e+01
1.27e+01
5.61e+00
3.05e+02
4.20e+02
3.20e+03
3.77e+01
1.39e+00
River
O.OOe+00
l.Ole+00
1.59e-01
7.11e+00
9.28e-02
3.94e-01
1.54e+01
1.78e+02
7.39e+01
1.31e+00
3.66e+00
6.21e+00
4.36e+00
8.13e-01
1.64e+02
5.25e+01
4.11e+01
8.21e+00
2.44e-01
,1.55e+00
2.05e+03
3.81e+02
3.51e+02
3.43e+03
1.88e+02
1.04e+03
1.45e+02
7.27e+02
3.33e+03
7.89e+01
                                9-12

-------
        Table 9-4: Slag Pathway Normalized Collective Doses (person-rem per Ci—in
                                        pathway)
Haclide
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-241
Cm-244
- Soil
6.43e+02
6.16e+02
1.81e+04
1.09e+01
6.85e+01
6.62e+01
3.03e-0r
6.61e+01
1.30e+02
1.29e+00
Concrete
1.97e+02
4.05e+01
2.85e+02
4.98e-03
7.57e-02
3.63e-02
1.02e-04
3.34e-02
5.66e+00
8.57e-04
Road Base
8.13e-02
6.97e-01
8.20e-01
7.54e-14
1.28e-06
3.01e-13
4.07e-08
1.25e-12
4.56e-09
2.76e-14
RR Ballast
1.51e+02
3.11e+01
2.19e+02
3.82e-03
5.80e-02
2.78e-02
7.85e-05
2.56e-02
4.34e+00
6.57e-04
River
6.35e+01
6.37e+01
1.80e+03
7.22e+02
8.08e+02
8.07e+02
1.54e+01
7.68e+02
9.42e+02
4.66e+02
Leachate

In addition to the direct exposure pathway, some nuclides could leach from the slag as a result of
infiltration of rainwater through pores or cracks in the pavement or laterally from overflowing
ditches. Some of this leachate would percolate through the soil to an underlying aquifer that is a
potential source of drinking water. Other fractions may drain into streams and rivers, while still
others may inundate fields and contaminate agricultural soils. Although the present analysis
does not attempt to quantify the'amount of radioactivity which is expected to travel these routes,
normalized collective doses from radionuclide release into a river system and soil contaminated
with radionuclides are presented for comparison. The impact of radionuclides leaching into
rivers and streams was assessed using EPA methodologies described in Environmental Pathway
Models for Estimating Population Health Effects from Disposal of High-level Radioactive Waste
in Geologic Repositories, EPA  520/5-85-026, J. M. Smith etal, 1985. Normalized collective
doses from radionuclides which leach from the road base and reach a river are presented in Table
9-4.
                                          9-13

-------
Asphaltic Concrete Aggregate

Thirteen percent of the slag is used as asphaltic concrete aggregate, which is used primarily for
paving roads. The slag constitutes 80% of the pavement.  As stated above, at the end of its
design life, most material used hi road construction is recycled into new roads.  Therefore, as
with slag used in road base construction, all slag used for road paving was assumed to be in place
for the entire 1,000 year evaluation period. The equations used to evaluate the collective dose
from slag used as asphaltic concrete aggregate are the same as those used to evaluate the road
base collective dose, except that the value of the dose conversion factor [Fix(x)] differs. The dose
conversion factor used for asphaltic concrete aggregate are 80% of those given in Federal
Guidance Report No. 12 (EPA 93), to account for the slag component of the pavement. The
normalized collective doses for radionuclides contained in slag used as asphaltic concrete
aggregate are shown in Table 9-4 (Concrete).

9.3.2  Ell

Slag is often used as fill. Typical fill applications include parking lots and road shoulders. Slag
is not recommend for fill applications under buildings because of stability concerns. For this
analysis it was assumed that slag used as fill would be covered with a layer of soil and/or
concrete, for example, if it were used in the construction of a parking lot. The potential
exposures from the soil-covered slag were assessed by modeling the fill as a municipal landfill
(i.e., buried contamination without extraordinary protection features). With these assumptions,
the primary exposure pathway from slag used as fill is ground water contamination.
Alternatively, if it had been assumed that the slag used as fill was uncovered, then it could be
assumed that people live on and grow their food in the slag, similar to the exposure pathways
developed in Section 9.3.4 for slag used as soil conditioner.

The impacts of slag used as fill (and modeled as a municipal landfill) were assessed with  the
Multimedia Environmental Pollutant Assessment System (MEPAS) (Buck 95) which is a
physics-based risk computation code that integrates source-term, transport, and exposure models.
Developed by Pacific Northwest Laboratory for screening and ranking of environmental
problems, MEPAS  is designed for site-specific assessments using readily available information
to estimate potential health impacts. This system has wide applicability to a range  of
environmental problems using air, ground water, surface water, overland, and exposure models.
                                          9-14

-------
Whenever available and appropriate, EPA guidance and models were used to facilitate
compatibility and acceptance. The methodology is illustrated hi Figure 9-3, below1
    Identify
    Sourca
Develop
Database
Transport Pathway Analysis
Exposure
Analysis
  Hazard
Assessment
Assessment
  Results




Atmospheric
Paihway
i
Overland
Paihway
,
S
P

^ Vadose
Ground- ,^"
1* water , r

L saiurs
^ Zon


ited
9
urf
Wa
ath
j
r
ace-
way




"**"




Figure 9-3: Simplified Flow Diagram of the MEPAS Methodology

The exposure scenario assumed for the normalized collective dose model is that ground water
would come into contact with the disposed material and that this ground water would in turn be
consumed by the down-gradient population. The normalized collective dose model does not
include people living directly over the fill who may inadvertently intrude into the fill.  The
municipal landfill characteristics shown below on Table 9-5 come primarily from NUREG/CR-
6147 (Dehmel 94). The information comes from a°I986 EPA survey: Solid Waste Landfill
Survey, OMB No. 2050-0061.  The survey was a stratified sample of 1,076 respondents of which
1,011 observations were used for analysis. The value used for aquifer thickness is the average
                    /'
value hi Newell 90.
       1  From "Overview of the Multimedia Environmental Pollutant Assessment System (MEPAS)," G. Whelan et
al., Hazardous Waste & Hazardous Materials, 9:191,1992
                                          9-15

-------
                      Table 9-5: Typical Landfill Values Assumed
*• •> # t *\ *
Farajmetef
..- , ,,.
Size
Depth
Capacity
Distance to nearest
private well
Depth to Aquifer
Population within
1 mile
M$mcipa1
32.5 acres
11 feet
3.75E+05 yd3
1,850 feet
30ft
3,684 people
' Hazantets
Waste
18 acres
28 feet
8.6E+05 yd3
2,250 feet
49 feet
481 people
The normalized collective doses for slag used as fill are shown in Table 9-6 under the municipal
and hazardous waste landfill scenario.
                             Table 9-6; Landfill Normalized
                                    Collective Doses
                             (person-rem per Ci—disposed)
Nuclide
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
Ag-HOm
Hazardous
NC
NC
8.18e-16
O.OOe+00
NC
O.OOe+00
O.OOe+00
O.OOe+00
1.15e+03
2.41e+02
1.646-30
O.OOe+00
NC
Municipal
5.75e+01
NC
1.03e+00
1.14e-17
NC
6.91e-09
O.OOe+00
5.54e-ll
3.02e+02
1.29e+01
1.05e+02
O.OOe+00
NC
                                         9-16

-------
Table 9-6: Landfill Normalized
       Collective Doses
(person-rem per Ci—disposed)
-Nuclide "
Sb-125
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-24l
Cm-244
i Hazardous
O.OOe+00
NC
NC
O.OOe+00
NC
6.73e-16
NC
O.OOe+00
O.OOe+00
9.14e-06
O.OOe+00
O.OOe+00
O.OOe+00
O.OOe+00
O.OOe+00
O.OOe+00
O.OOe+00
O.OOe+00
O.OOe+00--"
6.25e-09
NC
O.OOe+00
NC
O.OOe+00
NC
O.OOe+00
NC
; J Municipal
2.03e-12
NC
NC
1.14e-19
NC
3.54e+00
NC
NC
5.57e+00
1.06e-32
O.OOe+00
4.31e-34
NC
2.38e-34
6.25e-34
1.27e-04
5.26e+00
9.42e+00
8.86e-02
3.62e+03
NC
9.50e+01
NC
NC
NC
NC
NC
            9-17

-------
9.3.3 Railroad Ballast

Slag is sometimes used as railroad ballast. The collective impacts from slag used as railroad
ballast were calculated in the same manner as the impacts from slag used as a road base, except
that the values of the dose conversion factor [Flx(x)] and the occupancy factor [F0] differ. Since
the slag used as ballast would be uncovered, the railroad dose conversion factors were taken from
Federal Guidance Report No.  12. The railroad occupancy factor of 4.56 hrnr'-y"1 was calculated
from 1993 data obtained from the American Public Transit Association. The normalized
collective doses from slag used as railroad ballast are shown in Table 9-4.

Because it is exposed to the environment, such slag will be susceptible to leaching. The leachate
may either percolate through the underlying soil and potentially reach an aquifer, run off and
infiltrate nearby agricultural soils, or drain into rivers and streams.  As discussed above, this
pathway was not explicitly evaluated in this analysis, however, normalized collective doses from
the radionuclides that contaminate agricultural soils and drain into rivers are provided for
comparison.

9.3.4 Other Purposes

Other purposes for which slag is used includes soil conditioning, ice control, and miscellaneous
uses.

Soil Conditioning

Slag is used to raise the pH of acidic soils—this process is also known as "liming." The resulting
population impacts were assessed by means of the CU-POP model. (See Exhibit 9-A for a brief
description of the CU-POP model and Section 9.2 for how CU-POP was applied in this analysis
to model soil contaminated with radionuclides.) The soil conditioning normalized collective
doses shown in Table 9-4 are identical to the average normalized collective doses from the
deposition of air emissions shown in Table 9-3.

Ice Control

Slag is spread on ice- or snow-covered roadways to provide traction (see attached memo).  After
the precipitation has melted, the slag would be removed by street sweepers or washed into storm

                                          9-18

-------
sewers or roadside ditches.  Therefore, after a relatively short time (~4 months), slag used for ice
control would be removed from the roadways.  The ice control slag use alternative was not
evaluated for the following reasons: 1) the breakdown between ice control and soil conditioning
is unknown, 2) the breakdown between how much ice control slag ends-up in a waterway and
how much end-up on the surrounding land is unknown, 3) the slag spends a short tune on the
road, and 4) some states do not use slag for ice control (e.g., Michigan, southern states. All slag
used for 'Other Purposes' was assumed to be for soil conditioning since much of the slag used
for ice control could be washed onto the surrounding land.

9.3.5  Slag Normalized Collective Doses

Steel slag sales (SOL 93, SOL 95) hi thousand metric tonnes are listed in Table 9-7.


               Table 9-7:  Annual Steel Slag Sales (thousand metric tonnes)
Use
Asphaltic concrete
aggregate
Fill
Railroad ballast
Road bases
Other1
Total
1991'
1085
828
186
3238
1623
6959
1992
903
1073
224
2400
2256
6857
.1993
1090
905
116
2600
1900
6670
1994 i
1140
(14.6%)
1320
(16.3%)
160
(2.1%)
3170
(40.6%)
2000
(25.6%)
7800
1995
1040
(14.5%)
1380
(19.3%)
168
(2.3%)
2820
(39.4%)
1760
(24.6%)
7160
1 -Includes ice control, soil conditioning and misc. uses

The media normalized collective doses for slag were calculated by multiplying the five scenario
specific normalized collective doses calculated in Sections 9.3.1 through 9.3.4 by the percentage
of slag that was sold in 1995 for each re-use scenario and then summing the results:
                                          9-19

-------
                                        = £/„
(5)
      Aj  =   unitized collective dose factor from slag (rem/hr per Ci—slag)
      n   =   slag usage scenario: road base, concrete, fill, RR ballast, soil conditioning
      fn   =   fractional useage of slag in scenairo n, see Table 9-7
      An  =   unitized collective dose factor for scenanio n (rem/hr per Ci)
            i

The resulting slag normalized collective doses are shown in Table 9-15, along with the other
media specific normalized collective doses and the total normalized collective doses. If, as
discussed in Section 9.3.2, it is desired to model slag used as fill as soil conditioner as opposed to
a municipal landfill, then the soil conditioner usage fraction can simply be increased from 0.246
to 0.439; the unitized collective dose factors from slag would increase proportionally.

9.4   BAGHOUSEDUST

The volatile emissions and aerosols that evolve from an electric arc furnace during the melting
process are condensed and captured by the baghouse. The resulting dust is removed from the
mill by tanker trucks—its assumed fate in this analysis is described in this section.

9.4.1  Zinc Recovery

Approximately 86% of the dust is shipped to a processing plant for the extraction of zinc.
Exposure pathways will depend on the uses of the extracted metals and the disposal of the
residue. For this analysis,  it was assumed that the recovered zinc would be used to galvanize
automobile shells.  A description of the reference automobile shell is given in Section 9.5.1.   •,
Normalized collective doses from radionuclides contained within automobile shells are given on
Table 9-9.

9.4.2  Disposal in Landfill

The remainder of the dust is disposed of in a "secure" (i.e., hazardous waste) landfill.  The
collective radiological impacts of this dust will be small, inasmuch as the dust will be largely
isolated from the environment. However, this pathway was nonetheless investigated.  As with

                                          9-20

-------
slag used as fill (Section 9.3.2), the collective impacts from baghouse dust disposed of in a
hazardous waste landfill were calculated with the MEPAS computer program.

The normalized collective dose model exposure scenario assumed for disposal in a landfill is that
ground water would come into contact with the disposed material and that this ground water in
turn would be consumed by the down-gradient population.  The normalized collective dose
model does not include people living directly over the landfill, who may inadvertently intrude
into the landfill.

Information from 51 landfill units at 31 facilities was averaged to provide the typical hazardous
waste landfill characteristics used, as shown in Table 9-5. Geohydrology data from the 10
landfills that had complete and consistent information were averaged to provide the aquifer
characteristics of effective porosity, total porosity, Darcy velocity, and thickness.  Sand was used
as the aquifer soil type as a conservative assumption. The resulting normalized collective doses
are  shown in Table 9-6.

9.5   FINISHED STEEL

Steel is used is used to make a wide variety of finished products.  It is beyond the scope of the
present analysis to attempt to perform an exposure assessment of each product. An estimate of
the  impacts was made by breaking up the use of carbon steel into a number of general categories
and constructing an exposure scenario to represent each category.

Secondary Recycle

Collective impacts from the recycle of scrap from nuclear facilities are being evaluated over a
1,000 year period. It is unrealistic to assume that any commercial product would continue to
function for such a long period of time. Additionally, once a commercial product reaches the end
of its useful life it can either be disposed of in a municipal landfill or recycled into another
commercial product. In order to investigate the sensitivity of the resultant collective dose
impacts to whether the steel in commercial products is recycled, four cases were analyzed:
1) single use, then disposal; 2) 50% recycle, 50% disposal; 3) 90% recycle, 10% disposal; and
4) 100% recycle. In order to simplify the calculation, it was assumed that steel is always
recycled into the same product (e.g., steel initially recycled into an automobile stays hi an
                                          9-21

-------
automobile for the entire 1,000 year evaluation period).  If all commercial products had the same
useful life, then this assumption would not be needed. However, since useful lives vary, it is a
good approximation.

9.5.1 Automotive

Occupants of automobiles manufactured from scrap metal recycled from nuclear facilities could
be exposed to external radiation from the radioactivity in various automotive components. Three
auto components were modeled: the engine, the frame, and the shell. The primary assumptions
used to model the three components of the automobile are shown hi Table 9-8.
              Table 9-8: Primary Assumptions Used in the Automobile Model
Assumption
Weight
Dimensions
Density
Distance
Engine
368 Ib
23" x 23" x 29"
0.66 gm/cc
70cm
P Frame •
834 Ib
192" x 71" x 0.2"
7.87 gm/cc
1ft
Shell
945 Ib
Top: 192" x 71 "x 0.12"
Sides: 192" x 35" x 0.12" each
7.87 gm/cc
1ft
For the engine, an effective density was calculated by dividing the weight by the volume, while
for the frame and shell the effective thicknesses were calculated by dividing the weight by the
product of the density and area. The information in Table 9-8 was used to calculate exposure
rates with the MicroShield™ computer code. An effective exposure rate from the total
automobile was calculated by:

     A,  =  unitized collective dose rate from steel used in an automobile (rem/hr per Ci)
     n   =  component of the automobile: engine, frame, shell
     wn  =  weight of an automobile component (Ib), see Table 9-8
     A,,  =  unitized collective dose rate from an automobile component (rem/hr per Ci)
                                         9-22

-------
Automobile occupancy was estimated at 1,460 person-hr/yr, based on the assumption that two
individuals commute one hour each, way to and from work every day. The automobile's
effective life was assumed to be 7.3 years.

The 1,000 year integrated collective dose from the recycling of steel released from nuclear
facilities into automobiles is shown in Table 9-9. Table 9-9 shows only those radionuclides
which partition into steel, as indicated in Table 6-3.
                 Table 9-9: Automobile Doses (person-rem per Ci—in car)
Nuclide
Secondary Recycle
,0%
-50%
90%
100%
' , , Fin&Md Steel: Whole Car ' ' -
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65 -
Mo-93 .
Tc-99
Ru-106+D
Ag-llOm
Sb-125
O.OOE+00
1.05E+Q3
O.OOE+00
1.12E+04
O.OOE+00
O.OOE+00
5.49E+02
4.27E-01
1.96E-03
3.25E+02
2.72E+03
7.51E+01
O.OOE+00
1.06E+03
O.OOE+00
1.38E+04
O.OOE+00
O.OOE+00
5.49E+02
8.53E-01
3.93E-03
3.26E+02
2.72E+03
7.51E+01
O.OOE+00
1.06E+03
O.OOE+00
1.70E+04
O.OOE+00
O.OOE+00
5.49E+02
4.21E+00
1.96E-02
3.27E+02
2.72E+03
7.51E+01
O.OOE+00
1.06E+03
O.OOE+00
1.81E+04
O.OOE+00
O.OOE+00
5.49E+02
5.29E+01,
2.68E-01
3.27E+02
2.72E+03
7.51E+01
Baghowse Dust; Shell OxSy •, ;
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Mo-93
O.OOE+00
1.59E+03
O.OOE+00
1.68E+04
O.OOE+00
O.OOE+00
8.25E+02
8.70E-01
O.OOE+00
1.59E+03
O.OOE+00
2.07E+04
O.OOE+00
O.OOE+00
8.25E+02
1.74E+00
O.OOE+00
1.59E+03
O.OOE+00
2.56E+04
O.OOE+00
O.OOE+00
8.25E+02
8.58E+00
O.OOE+00
1.59E+03
O.OOE+00
2.72E+04
O.OOE+00
O.OOE+00
8.25E+02
1.08E+02
                                         9-23

-------
                  Table 9-9: Automobile Doses (person-rem per Ci—in car)
Nuclide
Tc-99
Ru-106+D
Ag-1 10m
Sb-125
' ,-- Seeoadaiy Recycle
- - - 0%
3.37E-03
4.92E+02
4.10E+03
1.14E+02
' 5Q& '
6.74E-03
4.94E+02
4.10E+03
1.14E+02
90%
3.37E-02
4.95E+02
4.10E+03
1.14E+02
! '106% "
4.61E-01
4.96E+02
4.10E+03
1.14E+02
9.5.2 Kitchen Appliances
Most major home appliances are located in, or near, the kitchen. Therefore, in order to estimate
exposures from steel recycled into home appliances, a representative kitchen was modeled.
Seven appliances were assumed to be in, or near, the kitchen: refrigerator, stove, dishwasher,
microwave oven, trash compactor, washer and dryer. Furthermore, it was assumed that the
kitchen was divided into two areas—a work area where the food is prepared and a dining area
where the family gathers to eat, talk, do homework, etc. Major assumptions used in the kitchen
model are given in Table 9-10.

                      Table 9-10: Kitchen Model Major Assumptions
Appliance
Refrigerator
Stove
Dishwasher
Microwave
Trash Compactor
Washer
Dryer
Weight
•CR>)
350
166
81
70
140
160
116
Dimensions (in) i
(hxwxd)
69.75 x 35.75 x 26.375
45.5 x 30 x 28
34 x 24 x 23.75
16.5 x 30 x 14
34.25 x 15 x 24.25
36 x 27 x 25.5
36 x 27 x 28.5
Distance (ft)
Work Area
5
1
5
5
5
10
10
0lning Area
15
15
15
15
15
20
20
                                         9-24

-------
Information concerning the weight and dimensions for most of the appliances listed in Table 9-
10 was obtained from the General Electric Answer Center in Louisville, Kentucky. The life
expectancy for each appliance was assumed to be 18 years. In selecting the distances at which to
calculate the dose rates, a representative kitchen approximately 10' x 20' was assumed.  It was
further assumed that a person would be working in front of one appliance and would be an
approximate equal distance from the others, except for the washer and dryer which would be
located in an out-of-the-way location. Likewise, the dining area was assumed to be on the
opposite side of the kitchen from the appliances.

An effective thickness for each appliance was determined from its weight and dimensions. Each
appliance was modeled as two slabs—representing the front and back panels of each
appliance—of this effective thickness and the above assumed height and width. Exposures rates
at the above assumed distances were mpdeled by use of the MicroSbield™ computer code.

To estimate kitchen occupancy times, a family of four was assumed. Furthermore, it was
assumed that all members of the family eat breakfast in the kitchen seven day a weeks and eat
dinner in the kitchen five days a week. Weekday lunch was assumed to be eaten in the kitchen
by only one member of the family. It was also assumed that one member of the family would  ,
spend one hour each week night doing homework in the kitchen. Based on these assumptions,
kitchen occupancy times of 70 person-minutes per day in the work area and 190 person-minutes
per day hi the dining area were calculated.

The normalized collective doses from exposure to home appliances are shown in Table 9-11.
These doses are the weight average of the normalized collective doses for each of the seven
appliances being evaluated (similar to equation 6 for the automobile).
               Table 9-11:  Finished Steel: Kitchen (Seven Appliances) Doses
                            (person-rem per Ci—in appliance)
Nuolide
C-14
Mn-54
Fe-55
Co-60
; Secondary Recycle
, 0%
O.OOE+00 -
3.14E+02
O.OOE+00
4.84E+03
$9%,
O.OOE+00
3.14E+02
O.OOE+00
5.08E+03
I 90%
O.OOE+00
3.14E+02
O.OOE+00
5.29E+03
. 100%
O.OOE+00
3.14E+02
O.OOE+00
5.34E+03
                                         9-25

-------
               Table 9-11:  Finished Steel: Kitchen (Seven Appliances)
                            (person-rem per Ci—in appliance)
Doses
Nuclide "
Ni-59
Ni-63
Zn-65
Mo-93
Tc-99
Ru-106+D
Ag-llOm
Sb-125
'V,.',.,o gecondary Recycle
0%.
O.OOE+00
O.OOE-t-00
L62E+Q2
9.62E-01
2.14E-03
9.78E+01
8.08E+02
2.26E+01
50%
O.OOE+00
O.OOE+00
1.62E+02
1.92E+00
4.28E-03
9.78E+01
8.08E+02
2.26E+01
90% "
O.OOE+00
O.OOE+00
1.62E+02
9.29E+00
2.13E-02
9.78E+01
8.08E+02
2.26E+01
100%
O.OOE+00
O.OOE+00
1.62E+02
4.87E+01
1.20E-01
9.78E+01
8.08E+02
2.26E+OI
9.5.3 Office Buildings

An office building has been selected to represent the exposure from finished steel used in
construction projects. The office building was assumed to have a modular construction, with six
offices being located within each module. The general layout of a single module and some of the
construction details which were assumed are shown on Figure 9-4.
                                         9-26

-------
                  Six Office Module
  Steel  ^
  Column
 DP = Dose Point

DP

DP
-
DP

Corridor

• •4*?* 	
1' DP

LI 	 in* 	 •»-

DP
L« 10' 	 to.
DP
m* >Lr
i

12
\
i
t
1
~]
12
I

Scale: None

1
1;
<

r\

, >v:-:":: ;
xl 1 1
2' ,
r
I;:-;: •-•:;;• v
111
X
(
-~~y




^-
Dffice Construction Details
K / — 2%" Concrete Slab •
/ with Reinforcing Mesh
-x:-:;::-:- . : :> -:.0... ...:.. .:../> /.,: -.•..•::•
1111 11 11^ U 1 1 1
Joists — ' \
on 24" Centers N — steel Deck
16 Gage
1' Dose,
* Point
•T- Steel Column — ^
1 m
::.--:-:,;.-:---V-::x:-: -:.;::-.-if / ' V r.*' ' ' "': -I "I:
11111111' '1111
=** /
^fi'
— - -/




'X—
\,
: -;.:••-•• J
1 1 1 1N

.-,.-, ;;-^A
^xxi?
pw
                           9-27

-------
The MicroShield™ computer code was used to calculate the dose rate to the six office workers
from each component of the module: columns, floor and ceiling joists, steel wall studs, steel
floor and ceiling decking. Each worker was assumed to spend 2,000 hr/yr in the office, and the
office module was assumed to have a 50 year service life. The normalized collective doses from
steel used in an office building are given in Table 9-12.
                     Table 9-12: Finished Steel: Office Building
                              (person-rem per Ci—in office)
Doses
Nuclide
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Mo-93
Tc-99
Ru-106+D
Ag-llOm
Sb-125
••'••• Secondary Recycle - " /
0%
O.OOE+00
3.78E+02
O.OOE+00
6.84E+03
O.OOE+00
O.OOE+00
2.04E+02
3.86E-01
3.32E-03
1.12E+02
9.80E+02
2.50E+01
: 50% ,
O.OOE+00
3.78E+02
O.OOE+00
6.84E+03
O.OOE+00
O.OOE+00
2.04E+02
7.64E-01
6.64E-03
1.12E+02
9.80E+02
2.50E+01
90%
O.OOE+00
3.78E+02
O.OOE+00
6.85E+03
O.OOE+00
O.OOE+00
2.04E+02
3.18E+00
2.91E-02
1.12E+02
9.80E+02
2.50E+01
100%
O.OOE+00
3.78E+02
O.OOE+00
6.85E+03
O.OOE+00
O.OOE+00
2.04E+02
7.01E+00
6.63E-02
1.12E+02
9.80E+02
2.50E+01
9.5.4 Cookware

Radioactivity may leach into food from cookware made from scrap metal originating at nuclear
facilities, and subsequently be ingested by members of the public. The information used to
model this potential pathway is primarily from three sources: SMA 87, KUL 92, and REI 85.
SMA 87 presents a national (United Kingdom) average ingestion rate of 0.12± 10% mg/day of
nickel attributable to stainless steel cookware. KUL 92 gives the results of tests using 5%.acetic
acid boiled for five minutes in stainless steel cookware. Although not strictly applicable to
ordinary cooking, the results can be used to estimate the relative amounts of iron, chromium and
nickel that are likely to be leached. The three utensils that most resemble American stainless
                                         9-28

-------
steel cookware (two were manufactured in the U.S. and one in Brazil) had average concentrations
of 0.267 mg Fe, 0.0963 mg Ni, and 0.0387 mg Cr per kilogram of water. From.this, it is deduced
that the total mass of metal leached is 4.17 times the mass of nickel.

Combining the data from these two references implies that the average total annual ingestion of
metal in food cooked in stainless steel cookware is  183±18.3 mg. Alternatively, the metal
ingested from food cooked in cast iron cookware can be inferred directly from data in REI 85.
From REI 85's concentrations of iron in beef and cabbage cooked in cast iron utensils, an average
total ingestion of metal is 1,950±680 mg.

A family of four was assumed to use a 350 gm frying pan.  The normalized collective doses from
the ingestion of metal in food cooked in a frying pan was integrated over the time it took for all
of the metal initially in the frying pan to be consumed (-480 years for stainless steel and -45
years for cast iron). The resulting normalized collective doses are shown in Table 9-13.
                       Table 9-13: Finished Steel: Frying Pan Doses
                               (person-rem per Ci—in pan)
Nuclide
Secondary Recycle
0%
50%. ...
9m . , .
100%
Cast Iron Frying Pan
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Mo-93
Tc-99
Ru-106+D
Ag-llOm
1.14E+02
9.55E+00
6.53E+00
5.25E+02
9.86E+00
2.95E+01
3.91E+01
7.34E+01
7.98E+01
1.14E+02
2.97E+01
2.15E+02
9.55E+00 ,
6.54E+00
5.43E+02
1.51E+01
5.00E+01
3.91E+01
1.39E+02
1.51E+02
1.14E+02
2.97E+01
7.53E+02
9.55E+00
6.56E+00
5.59E+02
2.62E+01
1.14E+02
3.91E+01 ,
4.81E+02
5.35E+02
1.14E+02
2.97E+01
1.89E+03
9.55E+00
6.56E+00
5.63E+02
3.22E+01
1.66E+02
3.91E+01
1.19E+03
1.37E+03
1.14E+Q2
2.97E+01
                                         9-29

-------
                        Table 9-13: Finished Steel: Frying Pan Doses
                              -  (person-rem per Ci—in pan)
Nuclide
Sb-125
"," ' Secondary Recycle ' ' -
0%
1.31E+00
* 50% ,
1.31E+00
90%-
1.31E+00
100%
1.31E+00
Stainless Frying Pan
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Mo-93
Tc-99
Ru-106+D
Ag-llOm
Sb-125
1.10E+01
8.99E-01
6.18E-01
5.00E+01
9.48E-01
2.83E+00
3.68E+00
7.07E+00
7.68E+00
1.07E+01
2.80E+00
1.23E-01
2.17E+01
8.99E-01
6.20E-01
5.19E+01
1.49E+00
5.00E+00
3.68E+00
1.40E+01
1.53E+01
1.07E+01
2.80E+00
1.23E-01
.1.02E+02
8.99E-01
6.22E-01
5.35E+01
2.76E+00
1.28E+01
3.68E+00
6.50E+01
7.30E+01
1.07E+01
2.80E+00
1.23E-01
4.57E+02
8.99E-01
6.22E-01
5.39E+01
3.50E+00
2.11E+01'
3.68E+00
2.83E+02
3.38E+02
1.07E+01
2.80E+00
1.23E-01
9.5.5 Finished Steel Normalized Collective Doses
The American Iron and Steel Institute (AISI) compiles statistics on carbon steel markets based on
21 major market classifications and 31 types of steel products (e.g., cold rolled sheets,
reinforcing bars, blooms, slabs, billets, etc.). Information included hi the AISI annual report for
1995 (AIS 95) has been collapsed into four categories and is summarized hi Table 9-14.

             Table 9-14: Distribution of Finished Steel into Commercial Products
Commercial Product
Automobile
Building (Office)
AISI Classifications . i
automotive; rail transportation
construction and contractors' products;
containers, packaging, shipping material; all
other AISI classifications
Percentage
36.5%
47.4%
                                          9-30

-------
Commercial Product
Home Appliances
(including Cookware)
Non-Accessible
, ' .AISICIasspcafioBs j
appliances, utensils, cutlery; other domestic and
commercial equipment
oil and gas industry, electrical equipment
Percentage
5.8%
10.3%
"Containers, packaging, shipping material" and "all other AISI classifications" were included in
the Building (Office) classification because they are a small fraction (-12.5%) of the total carbon
steel market and because the normalized collective doses for this classification fall between the
normalized collective doses for the other two classifications which were analyzed. "Oil and gas
industry" and "electrical equipment" were assumed to be composed primarily of oil derricks, off-
shore drilling platforms, and transmission towers, which would not normally be accessible to the
general population.

The media normalized collective doses for finished steel were calculated by multiplying the four
scenario specific normalized collective doses calculated in Sections 9.5.1 through 9.5.4 by the
distribution of steel in each re-use scenario, and then summing the results:
                                                                                    (7)
     Af =   unitized collective dose factor from finished steel (rem/hr per Ci—steel)
     n  =   steel usage scenario: automobile, appliance, office, frying pan, non-accessible
     fn  =   distribution of steel in scenario n, see Table 9-14
     An =   unitized collective dose factor for scenario n (rem/hr per Ci)

The resulting finished steel normalized collective doses are shown in Table 9-15, along with the
other media specific and the total normalized collective doses.

9.6  TOTAL NORMALIZED COLLECTIVE DOSES AND RISKS

For each radionuclide, the total normalized collective dose and risks were calculated, taking into
account the partitioning of the radionuclide to the differing resulting media (i.e., steel, slag,
baghouse dust, and air emissions), and transportation associated with each of the media. The
total normalized collective doses were calculated by:
                                          9-31

-------
                                                                                     (8)
      ATJ  =   total collective dose factor for radionuclide z (person-rem per Ci scrap)
      m   —   transportation media type: scrap, slag, baghouse dust, finished steel
      Pmi =   transportation partition ratio of radionuclide i to media m

      A^j =   radionuclide i specific transportation collective dose factor for media m (person-
              rem per Ci)
      n   =   media type: air emission, slag, baghouse dust, finished steel
      Pni  =   partition ratio of radionuclide i to media n, see Table 6-3
      Ara-  =   radionuclide / specific collective dose factor for media n (person-rem per Ci)

For slag, baghouse dust and finished steel, the transportation partition ratios are the same as the
media partition ratios; for scrap, the transportation partition ratio is one and for air emissions it is
zero.  The total transportation normalized collective doses, the media specific normalized
collective doses, and the total normalized collective doses are shown in Table 9-15. Also, shown
in Table 9-15 are the total normalized collective cancer risks for each radionuclide.
                                          9-32

-------
Table 9-15: Total Normalized Collective Dose and Risks (per Ci—in scrap)
Huclide-
C-14
Mn-54
Fe-55
Co-60
Ni-59
Ni-63
Zn-65
Sr-90+D
Nb-94
Mo-93
Tc-99
Ru-106+D
Ag-llOm
Sb-125
1-129
Cs-134
Cs-137+D
Ce-144+D
Pm-147
' "•' : ]• ' • Do$e$^^0»-rem) - ff ' •
Trps
O.OOE+00
1.42E-02
O.OOE+00
6.35E-02
O.OOE+00
O.OOE+00
5.31E-03
O.OOE+00
5.77E-03
5.60E-07
1.45E-09
4.63E-03
6.58E-02
8.78E-03
8.19E-07
1.64E-02
5.82E-03
1.29E-04
3.27E-09
; Air 1
4.74E+01
2.73E-04
1.61E-06
1.22E-03
8.80E-06
4.86E-05
2.44E-02
6.40E-02
8,02E-02
2.71E-04
1.18E-03
1.47E-04
3.53E-04
3.41E-05
1.54E+03
6.00E-02
L79E-01
5.07E-04
4.41E-05
Slag
O.OOE+00
1.67E-01
3.98E-03
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
5.87E+01
5.35E+02
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
1.33E-01
7.42E-01
5.04E-02
-6.49E-01
Dust
O.OOE+00
O.OOE+00
1.64E-17
O.OOE+00
O.OOE+00
O.OOE+00
1.65E+02
3.09E+00
2.81E+01
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
2.65E-03
3.42E-02
'Steel-
...-'. f ..
2.12E-02
3.80E+02
2.65E-04
1.01E+04
1.33E-03
6.84E-03
6.13E+01
O.OOE+00
O.OOE+00
2.53E+01
1.91E-01
1.76E+02
1.49E+03
4.02E+01
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
Total -
. .< 	
4.75E+01
3.80E+02
4.24E-03
1.01E+04
1.34E-03
6.89E-03
2.26E+02
6.18E+01
5.63E+02
2.53E+01
1.92E-01
1.76E+02
1.49E+03
4.02E+01
1.54E+03
2.09E-01
9.27E-01
5.36E-02
6.83E-01
Cancers
Total
2.34E-02
2.88E-01
1.57E-07
7.66E+00
1.18E-06
6.60E-06
1.72E-01
2.25E-02
3.81E-01
1.92E-02
1.58E-04
1.34E-01
1.13E+00
3.05E-02
1.01E+00
1.29E-04
6.39E-04
2.73E-05
2.05E-06
''Fatal, , .
1.62E-02
1.93E-01
1.56E-07
5.13E+00
1.18E-06
6.58E-06
1.15E-01
1.78E-02
2.54E-01
1.29E-02
1.24E-04
9.01E-02
7.61E-01
2.05E-02
1.04E-01
8.68E-05
4.27E-04
1.74E-05
1.09E-06
                                9-33

-------
Table 9-15: Total Normalized Collective Dose and Risks (per Ci—in scrap)
Nuclide
Eu-152
Pb-210+D
Ra-226+D
Ra-228+D
Ac-227+D
Th-228+D
Th-229+D
Th-230
Th-232
Pa-231
U-234
U-235+D
U-238+D
Np-237+D
Pu-238
Pu-239
Pu-240
Pu-241+D
Pu-242
Am-241
Cm-244
Doses (person-rein)
Trans
4.02E-03
9.19E-07
6.44E-03
3.31E-03
9.45E-04
5.60E-03
7.07E-04
2.57E-07
8.08E-08
7.04E-05
6.35E-08
2.48E-04
6.19E-05
5.09E-04
1.18E-08
4.78E-08
1.16E-08
3.53E-09
1.02E-08
4.66E-06
1.02E-08
Air
4.94E-03
4.08E+00
2.60E+00
1.49E-02
6.86E-01
2.48E-01
8.30E-01
9.42E-01
9.83E-01
1.52E+00
2.98E-01 x
2.83E-01
2.72E-01
5.01E+00
3.25E-01
3.65E-01
3.64E-01
5.43E-01
3.47E-01
5.70E-01
2.85E-01
Slag
1.01E+01
O.OOE+00
2.81E+03
1.12E+01
2.00E+01
2.18E+00
1.99E+02
7.49E+02
1.34E+03
9.68E+02
1.55E+02
1.82E+02
1.50E+02
4.94E+03
2.55E+00
3.34E+01
1.55E+01
7.08E-02
1.55E+01
3.13E+01
3.02E-01
Dust
5.34E-01
O.OOE+00
1.48E+02
5.88E-01
1.05E+00
1.15E-01
1,05E+01
3.94E+01
7.03E+01
5.10E+01
8.18E+00
9.60E+00
7.92E+00
2.60E+02
1.34E-01
1.76E+00
8.15E-01
3.73E-03
8.13E-01
1.65E+00
1.59E-02
Steel
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
O.OOE+00
Total
1.07E+01
4.08E+00
2.96E+03
1.18E+01
2.17E+01
2.55E+00
2.11E+02
7.89E+02
1.41E+03
1.02E+03
1.64E+02
1.92E+02
1.59E+02
5.20E+03
3.01E+00
3.56E+01
1.67E+01
6.17E-01
1.6'6E+01
3.35E+01
6.03E-01
Cancers
Total
7.77E-03
3.67E-04
7.82E-01
8.11E-03
3.67E-03
1.73E-03
8.09E-02
2.11E-01
9.18E-01
2.07E-02
2.59E-02
5.35E-02
4.02E-02
3.41E-01
2.23E-04
1.29E-03
1.24E-03
6.79E-05
1.23E-03
3.57E-03
3.37E-05
Fatal
5.17E-03
2.90E-04
5.49E-01
5.41E-03
2.50E-03
1.16E-03
5.47E-02
1.53E-01
6.15E-01
1.46E-02
1.60E-02
3.46E-02
2.50E-02
2.78E-01
1.97E-04
1.12E-03
1.08E-03
5.88E-05
1.07E-03
2.86E-03
2.92E-05
                                9-34

-------
                                   REFERENCES

AIS 95       American Iron and Steel Institute, "Shipments of Steel Products by Market
             Classification, Carbon," Year 1995.

Buck 95      Buck, J.W. et al., "Multimedia Environmental Pollutant Assessment System
             (MEPAS) Application Guide," PNL-10395, February 1995.

Dehmel 94    Dehmel, J-C., et al., "Characterization of Class A Low-Level Radioactive Waste
             1986-1990 "MJREG/CR-6147, January 1994.  ,
         '"I
EPA 93      U.S. Environmental Protection Agency, "External Exposure to Radionuclides in
             Air, Water, and Soil," EPA 402-R-93-081, September 1993,

EPA 94      U.S. Environmental Protection Agency, "Estimating Radiogenic Cancer Risks,"
             EPA 402-R-93-076, June 1994.

EPA 94a     U.S. Environmental Protection Agency, Health Effects Assessment Summary
             Tables, FY-1995 Supplement," EPA/540/R-95/142, November 1995.

EPA 94b     U.S. Environmental Protection Agency, "Radiation Site Cleanup Regulations:
             Technical Support Document for the Development of Radionuclide Cleanup
             Levels for Soil," Review Draft, EPA 402-R-96-011, September 1994.

KUL 92      Kuligowski, J., and K. Halperin, "Stainless Steel Cookware as a Significant
             Source of Nickel,  Chromium, and Iron," Archives of Environmental
             Contamination and Toxicology, 21, 211-215, (1992).
                                                                               i
NCRP 83     National Council on Radiation Protection and Management, "Iodine-129:
             Evaluation of Releases for Nuclear Power Generation," NCRP Report No. 75,
             December 1, 1983.

Newell 90    Newell, C.J., L.P. Hopkins, and P.B. Bedient, "A Hydrogeologic Database for
             Ground Water Monitoring," Ground Water, 28(5):703-714 (1990).
                                        9-35

-------
NRC 95      National Research Council, "Technical Bases for Yucca Mountain Standards,"
             1995.

O'Neal 90    O'Neal, B.L. and C.E. Lee, "IMPACTS-BRC, Version 2," NUREG/CR-5517,
             April 1990.

RE! 85       "The Dietary Significance of Adventitious Iron, Zinc, Copper, and Lead in
             Domestically Prepared Food," Food Additives and Contaminants, 2, 3,1985.

SMA 87  '    Smart, G. And J. Sherlock, £CNickel in Foods and the Diet," Food Additives and
             Contaminants, 4,1,1987.

SOL 93       Solomon, Cheryl, "Slag-Iron and Steel, Annual Review-1992," U.S. Department
             of Interior, Bureau of Mines, September 1993.

SOL 95       Solomon, Cheryl, "Slag-Iron and Steel, Annual Review-1994," U.S. Department
             of Interior, Bureau of Mines, August 1995.
                                       9-36

-------
                                    EXHIBIT 9-A

  CU-POP: A MODEL FOR ASSESSING THE COLLECTIVE IMPACTS
 OF RADIOACTIVELY CONTAMINATED SOIL
 CU-POP is a simple, generally conservative model developed to estimate the radiological
 impacts of contaminated soil on both on-site and off-site populations. The model accounts for
 radioactive decay and the ingrowth of progeny and for the transit time of contaminants through
 the unsaturated zone.

 The five specific exposure pathways addressed in CU-POP are:  1) external exposure to
 penetrating radiation from contaminated soil;  2) inhalation of suspended dust;  3) exposure to
 indoor radon progeny (when included); 4) ingestion of crops raised on contaminated soil;  and
 5) ingestion of contaminated ground water. A sensitivity analysis has shown that for the
 radionuclides and the environmental conditions reflected in the reference sites, other pathways
 (such as soil ingestion and irrigation) are not important contributors to the radiological impacts.

 The first three pathways affect only the population residing on-site. The radiation exposure of
 any individual at site i is assumed to be proportional to the average concentration of they-th
 radionuclide on the portion of the site where he or she resides. The estimated cumulative,
 collective radiological health impact on the population (e.g., number of radiogenic cancers
 induced) attributable to the &-th pathway is the sum of the impacts on these individuals,
 integrated over space and over the time of site occupation—e.g., from t=0 to t - 1000 years.
. This contribution to the population impact, R,jh, can be expressed as
R.L.  =
                                                           dt/                     (1)
 where C,J(r,t') is the concentration of nuclidej at position r and time /', A, and p are the
 (constant) thickness and density of the contaminated soil layer and S-, its initial area, N, is the
 population density (assumed to be uniform and constant), and Kijk is the appropriate site-,
 nuclide-, and pathway-specific constant of proportionality for the population impact under
 consideration. The quantity inside the square brackets represents the total activity of nuclide/ at

                                          9-37

-------
time  tf, and is independent of its spatial distribution.  C^rJ) depends on Ca(r,0), the concentra-
tion at t = 0 (the time of site occupancy) and, in cases where nuclidej is a member of a radio-
active decay series and j-l,j-2, etc., refer to its progenitors, on Ctij.i(r,0), C,j.2(r,0),..., as well.

      CU-POP assumes that all food raised on the site is consumed, either on- or off-site, and
that a fixed fraction (0.5%) of the activity that reaches the aquifer under the site is also
consumed. The radiological exposures via these two pathways are thus also proportional to the
total radioactivity, but are independent of the site population density. They can be modeled by
Equation 1, but with N, = 1.

      Because the collective impact is a function of the total inventory of all nuclides on the site
at / = 0, but not on the distribution of that activity over the area of the site, it is possible to
simplify Equation 1. Since the time evolution (e.g., leaching, radioactive decay, etc.) of a given
nuclide at a given site is independent of position,  let us set Ct/r,t/) = C,J(r,Q)-c,J(t/) where, by
definition, c^O) = 1. Then Equation 1 may be separated and summed over all nuclides and
pathways as:
   E»,  t/
   nik Kiji
k
                                                      fC(r,0)v p
                                                                                     (2)
where the index k refers to the five exposure pathways modeled by CU-POP; and n,k - JV, if
l£ k £ 3,  otherwise n,k — 1.

      Ku, of Equation 2 relates the cumulative, collective health impact of radionuclide/ at site i
to its total activity there and is calculated with CU-POP. Qg is the total inventory of the nuclide
at that site.
                                          9-38

-------
                                    CHAPTER 10

                 EVALUATION OF UNCERTAINTIES PERTAINING TO
                 -   SCRAP METAL QUANTITIES, ESTIMATES OF
         DOSE AND RISK, AND MINIMUM DETECTABLE CONCENTRATIONS

10.1   INTRODUCTION

The option to recycle large quantities of scrap metal from the decommissioning of DOE facilities
and commercial nuclear power plants may result in small radiation exposures to selected
individuals, such as workers, and to members of the general public. It is anticipated that all
releases of scrap metal for recycling will require a demonstration that residual contamination
levels meet applicable regulatory guidelines.

At present, however, regulatory guidelines aimed specifically at scrap metal recycling from
nuclear facilities do not exist. As a result, it is not possible at this tune to make rigid estimates of
future exposure and risks to individuals and to population groups. In acknowledgment of this
limitation, EPA calculated radionuclide-specific dose estimates for individuals are based on a
unit concentration (i.e., mrem per pCi/g of residual contamination in scrap metal) and for
population groups on a Curie quantity throughput of scrap metal (i.e., person-rem per Ci of
radioactivity hi scrap metal). Exposure estimates of this type are referred to as "normalized."

Chapters 7 and 9 provided detailed estimates of normalized individual and collective doses,
respectively, for a variety of modeled exposure scenarios.  Scenario selection for these modeled
dose estimates are considered reasonable but will HKely yield high end doses.  The relationship
between a starting contamination level in scrap metal and the estimated dose(s) to an individual
or group of individuals is uncertain.  Modeled dose estimates require values to be assigned to a
large number of variables referred to as model parameters. In some instances, values of model
parameters may be unknown. In other instances, the values of model parameters, even if known,
are highly variable.

In the past, models that employed single conservative values for each parameter resulted in
overestimated doses.  Such conservative models were employed as screening tools but otherwise
had limited value in risk and regulatory impact analyses.
                                         10-1

-------
Emphasis is now being placed on removing conservative assumptions and incorporating
"realism" into model predictions. Accordingly, dose calculations and risk assessments are
tailored, whenever possible, to actual locations or conditions of exposed persons. They
incorporate site-specific meteorological data, food production/consumption information, and
demographic information as much as possible. Many critical model parameters cannot, however,
be determined on a site-specific basis. This is especially true for a model that attempts to predict
exposures to future population groups that have yet to be identified.

Attempts at improving the realism of model predictions by removing conservative assumptions,
despite uncertainties, increase the probability of underestimating human dose.  When examining
the results of such evaluations, it is reasonable to question their accuracy and to ask whether
actual doses might exceed regulatory standards. To address this issue, uncertainty-analysis is
conducted. Today, uncertainty analysis comprises an integral part of the Agency's scientific and
regulatory analyses.

EPA has long recognized the usefulness of uncertainty analysis and the need to provide ranges of
estimates rather than point estimates. According to EPA guidance (EPA 95), a complete
uncertainty analysis must address both uncertainty and the variability of individual parameters.
Uncertainty is concerned with gaps in data or parameter information that is incomplete;
parameter variation is concerned with actual variations in values for the parameters.

Deterministic Versus Stochastic Models

When a model employs single values for individual parameters, it is referred to as deterministic;
when individual parameters are defined by ranges or distributions of values, the model is
probabilistic (or stochastic). Figure 10-1 compares the simple operation of a deterministic model
with that of a more complex probabilistic model. When using the deterministic model,  one
simply chooses values of x and y, and the model calculates a single output value of z. In a
probabilistic model, the user specifies the distribution of each variable and the model then
"samples" the distribution of x and y to calculate a value of z. When this calculation is repeated
many times, a distribution of the output values of z is generated; this is known as a Monte Carlo
simulation. In this report, Monte Carlo analyses were not performed due to the unavailability of
information regarding the distribution of the values of the calculational parameters. However,
information was available on the range of values for key parameters.  As such, the sensitivity/
                                          10-2

-------
      DETERMINISTIC MODEL
   PROBABILISTIC MODEL
      SINGLE VALUE OF EACH
      INPUT PARAMETER x, y ...
                 INPUT
              CODE
            OPERATION
                 OUTPUT
         SINGLE VALUE OF
           EACH OUTPUT
            QUANTITY
            z = x + y
    VALUES TO'DEFINE
       PARAMETER
      DISTRIBUTION
             INPUT
CODE OPERATION

            PARAMETER x
           PARAMETER y
        y«3
                                              y(2J
                                                     OUTPUT
                                         DISTRIBUTION 0* VALUES
                                               OF z * y +• x
                                                 2(1} Z(2)
Figure 10-1  Comparison of a Deterministic Model and a Probabilistic Model
                                (from Little 1983)
                                      10-3

-------
uncertainty analysis provided in this report evaluated how the results may change over the
                il              '         *         i
plausible range of the values for the key calculational parameters. As such, the uncertainty/
sensitivity analyses provided in this report are referred to as "semi-quantitative." Examples of
some of these parameters include (1) the partitioning and fate of individual radionuclides during
the smelting process, (2) disposition or commercial uses of products and side-streams associated
with metal smelting, (3) probable ratios of contaminated scrap to clean scrap that would serve to
dilute activity levels in finished products, as well as smelting byproducts, and (4) conditions of
exposure for scrap metal workers and end-users of finished products and byproducts.

This chapter summarizes the uncertainties and sensitivities for each of the four major elements
defined in the TSD:

  (1)   Quantities and characteristics of scrap metals from DOE facilities and nuclear power
       plants that are potentially available for recycling.

  (2)   Radionuclide-specific estimates of potential normalized annual doses and risks to the
       reasonably maximally exposed individual (RMEI) associated with the free release of
       scrap metal from nuclear facilities.

  (3)   Radionuclide-specific estimates of potential normalized collective doses/risks to the
       exposed population due to free release of scrap metal.

  (4)   Minimum detectable concentrations of radionuclide contaminants associated with scrap
       metal.

An expanded discussion of these uncertainties is provided hi Appendix L of the TSD.
                                             a*-

10.2   UNCERTAINTIES IN SCRAP METAL SOURCE QUANTITIES AND LEVELS OF
       CONTAMINATION

Uncertainties pertaining to scrap quantities and contamination levels have widely differing
impacts on dose estimates. Uncertainly of scrap metal quantity is not expected to have a
significant impact on RMEI doses since these are principally dictated by the radionuclide
concentrations (e.g., pCi/g) of scrap at the point of release, and not on the quantity of scrap
recycled. In contrast, uncertainty in scrap metal quantities is expected to significantly impact the
estimated collective doses since these are proportional to the throughput of scrap metal. For
                                          10-4

-------
example, for a given fadionuclide contamination level, doubling the quantity of recycled scrap
metal is expected to double the collective dose.

10.2.1 Scrap Metal from Nuclear Power Plants

The total quantity of scrap metal that may be available for recycling following decontamination,
decommissioning, and dismantlement of 123 commercial power reactors is estimated to be about
650,000 metric tonnes, of which about 77% is carbon steel, 17% is stainless steel, and the
remainder consists of a variety of other metals and metal alloys. This section describes the
uncertainty in these estimates. In summary, it is concluded that, based on currently available
information, the estimated inventory of scrap metal potentially available for recycling from
commercial nuclear power reactors is not likely to be more than a factor of two higher or
lower than the estimated value.

Quantities of scrap metals and their contamination levels used in the analysis were largely based
on a deterministic model that primarily employed data from two reference reactor facilities. For
this approach, uncertainty analysis of modeled estimates is best achieved by first identifying
major differences among the 123 reactor plants.  Differences among U.S. reactors that are
deemed critical to future quantities of scrap metal involve those that define a facility in terms of
its (1) physical design, (2) plant operations, and (3) choice of decommissioning alternatives. A
model that adequately accounts for the variability of these parameters within the universe of
reactors can be expected to yield estimates that can be viewed with reasonable confidence.
                                                -t &i
Physical Design                                  '
                                                                       t
Foremost in defining potential scrap metal quantities are physical parameters that are determined
by the class of reactor, reactor size, and period of construction.  These physical variables are well
documented for the 123 reactor units and were factored into scrap metal estimates by: (1)
employing a Reference plant for each of the two major reactor types (i.e., Reference BWR and
Reference PWR), (2) using an empirical scaling factor where plant power rating served as a
surrogate measure of reactor size, and (3) using Reference facilities constructed about midway
through the 30-40 year construction period that defines the nuclear power industry. (One
Reference facility was pre-TMI-2 era and the other was post-TMI-2 era).
                                          10-5

-------
Operational Factors

Contamination levels and relative radionuclide composition on interior and exterior metal
surfaces of reactors are strongly influenced by numerous operational factors. These include: (1)
years of operation, (2) coolant chemistry and corrosion control, (3) fuel integrity, (4)
performance or failures of critical reactor components/systems and their maintenance, and (5)
health physics practices and routine cleanup efforts.

Operational  factors are not, however, easily incorporated into models and therefore were not
considered.  This is due in part to me sporadic occurrence of some operational factors (e.g.,
system/component failure, fuel leakage) and in part to the subjective nature of others (e.g.,
quality of coolant water chemistry, corrosion control, health physics practices, etc.-).  Although
operational factors undoubtedly contribute to large variability in the metal contamination
characteristics among facilities, the individual plant differences will likely average the
variabilities  so that the uncertainty in the collective characteristics of the industry as a whole is
believed to be relatively small compared to the individual variability among plants.

To reflect the high degree of variability as reported by studies referenced in Appendix A and a
small number of current decommissioning plans, plant systems in this report were grouped into
one of three  levels of contamination, where each level represents a range of values that spans
three or more orders of magnitude.

The projected contamination levels used in this report may represent upper-bound values.  This is
due to the biased data from which contamination levels were necessarily derived. The studies
that were used to derive contamination levels (e.g., NUREG/CR-4289) present data from reactor
facilities with abnormal histories of operation and which are not representative of the industry at
large. Most of the operations at these reactors preceded the 1979 TMI-2 accident and reflect
material composition, plant systems, and operational standards of the pre-TMI era. The 1979
accident triggered major reforms in the commercial nuclear industry in the form of more
stringent Federal regulations and performance standards issued by the NRC.  Post-TMI reforms
also reflect the introduction of new standards, guidance, recommendations, and good practices
from the American National Standard Institute (ANSI), American Nuclear Society (ANS),
American Society for Testing and Materials (ASTM), National Council on Radiation Protection
and Measurements (NCRP), Electric Power Research Institute (EPRI), and others. The
                                          10-6

-------
organization that has most significantly influenced post-TMI plant operations is the Institute of
Nuclear Power Operations (INPO). INPO's efforts to improve and standardize reactor plant
operations can be expected to have two principal effects on reactor contamination levels at the
time of decommissioning. On average, contamination levels can be expected to be below those
identified in this report and the range or variability of contamination levels among individual
plants is likely to diminish.

Decommissioning Alternatives

The final variable affecting scrap metal contamination levels (and consequently scrap metal
quantities) is the choice of available decommissioning alternatives (DECON, SAFSTOR, or
ENTOMB). SAFSTOR, with its extended delay in dismantimg/decommissioning-, will have the
obvious impact of reducing contamination levels by up to several orders of magnitude. The
ENTOMB option is not expected to be used.

Depending on prevailing decontamination technologies and economic factors, a reduction in
residual contamination levels hi scrap could significantly increase scrap metal quantities.  For
example, if prevailing cost-effective decontamination technologies were limited to reducing
contamination to four orders of magnitude, scrap metal at 10-years post-shutdown with activity
levels > 5 x 107 dpm/100 cm2 could not be expected to meet the current release standard of 5,000
dpm/100 cm2 and would, therefore, be excluded from recycling.  Under the SAFSTOR
alternative, if starting contamination levels were reduced by several orders of magnitude through
natural decay, an expanded fraction of the total pool of scrap metal could be expected to meet
potential release criteria.

At this time, however, the vast majority of reactor licensees have not revealed their preference
for a specific decommissioning alternative and speculation regarding decontamination
technologies for nearly a century into the future would be unwise. For these reasons,
uncertainties associated with decommissioning alternatives were not addressed hi this report.
Scrap metal quantities and residual contamination levels were based on a 10-year post-reactor
shutdown period and current decontamination technologies.
                                          10-7

-------
Summary Conclusions Regarding Uncertainty


There is significant diversity among the current U.S. inventory of 123 licensed nuclear reactors
            "i.
that is likely to yield variable quantities and contamination levels of scrap metal among
individual reactor units at the time of decommissioning.  Modeled estimates based largely on two
reference facilities support the following statements:


       (1)     Physical differences inclusive of plant design, power rating, and period of
              construction are thought to be the most important parameters affecting scrap metal
              quantities for individual reactors and were incorporated into the modeled results.

       (2)     Parameters that could not be readily defined (i.e., operational factors) are likely to
              represent a continuum with a symmetrical distribution about a mean value. Thus,
              factors contributing to low quantities of scrap metal containing radioactive
              contamination for some plants will be offset by others yielding higher than
              expected scrap metal quantities. As such, the uncertainty hi the collective
              quantities and radionuclide inventories for all plants combined are likely to be
              considerably smaller than the variability among plants.

       (3)     Variations among reactor plants concerning selection of a decommissioning
              alternative will likely impact residual contamination levels of individual reactor
              systems. Although residual contamination levels undoubtedly affect
              decontamination strategies and costs, the mass quantity of available scrap may
              only be modestly affected.

       (4)     Based on currently available information, it is concluded that the estimated
              collective inventory of scrap metal available for recycling from commercial
              nuclear power reactors is not likely to-vary by more than a factor of two higher or
              two lower than the estimated value of 650,000 metric tonnes.
10.2.2 Scrap Metal from DOE Facilities


In this report, scrap metal estimates for DOE facilities were defined as: (1) existing scrap metal
inventories that are currently stored at 13 DOE sites and (2) future scrap metal inventories that
are anticipated as a result of decommissioning. The estimated existing inventory of about
171,000 metric tonnes of existing scrap is small in comparison to future quantities which are
projected to exceed 925,000 metric tonnes.  Based on available data, the current estimated
value of existing DOE contaminated scrap is considered not likely to differ from the true

                                          10-8

-------
value by more than a factor of two. The quantity of scrap associated with the future
decommissioning of DOE facilities is more uncertain and may be several times higher than
the estimated value.  The actual future quantify could also be perhaps a factor of two lower
if many of the DOE facilities are decontaminated and used for other purposes.

As will become apparent in the discussion that follows, these estimates of uncertainty reflect the
judgment of the authors based on a review of the available literature and are provided as a basis
for further discussion and future investigations.

Table 10-1 identifies site-specific scrap metal quantities and primary source documents from
which data were obtained to estimate scrap quantities. The information is limited to
i
deterministic (i.e., single) values of scrap metal estimates at individual DOE sites, with no
additional data that would further define accuracy or variability of cited values. Moreover,
available data were frequently speculative, incomplete, or insufficiently detailed.  Quantitative
and qualitative deficiencies hi available data, therefore., necessitated the use of surrogate values,
assumptions, and interpolation.

      Table  10-1. Selection of Data Sources for Scrap Metal Quantities at DOE Facilities

DOE Site
1
Fernald
Hanford
INEL
LANL
NTS
OKNL
Y-12
K-25
Paducah
Portsmouth
Rocky Flats
SRS
Weldon Spring
SubTotal
TOTAL
Existing scrap metal ( MTs)
Source Document
MIN96
4,218
377
727
.
264
1,129
9,065
29,357
48,374
8,914

13,183

HAZ95



3,099






24,543

27,839
171,089
Future scrap metal (MTs)
Source Document
MIN96


33,486

--
—
—




3,054
—
EPA/SCA 95
135,623
91,798

2,686
—
—
—



26,303

—
DOE 95




—
—
— /
212,706
230,886
189,072


—
925,614
1,096,703
                                          10-9

-------
In acknowledgement of data limitations and their adverse impact on uncertainty, the DOE stated
the following:
                                                                  **
       "... Because of limited data, this report does not attempt to capture the exact
       amount of each material in inventory. Rather, it attempts to capture the general
       magnitude of the inventory of each material (MIN 96)."

Elsewhere, the DOE (MIN 96) concluded that while the"... Department maintains detailed
inventory systems of weapons components ... there is no reliable system to identify a complete
inventory of scrap metal  and equipment (MIN 96)." (Emphasis added)

A reasonable interpretation of these statements is that cited quantities reflect best estimates (as
opposed to comprehensive measurements) and, therefore, pose a significant but undefined level
of uncertainty. As a rule, deterministic data preclude a rigorous uncertainty analysis. Reliability
of deterministic data, however, may be assessed by means of a subjective evaluation that focuses
on the methods employed by DOE for data collection.

DOE Data-Collection Methods

DOE's 1996 Materials in Inventory Initiative was a year-long Department-wide effort aimed to
improve management and disposition for materials that may no longer be needed.  The objectives
of this effort focused on management approaches for: (1) uncontaminated materials, (2) suspect
materials, (3) contaminated material,  and (4) clearing suspect or contaminated materials to the
property management system for reuse or release.  Data sources for the MIN Scrap Metal and
Equipment Team that developed scrap metal estimates included the following:

       •     DOE Regulations, Policies, and Orders that served as  sources of information on
             requirements and procedures for managing scrap metal and equipment.

       •     Studies conducted within the past five years pertaining to scrap metal inventories
             within the DOE complex.

       •     Information collected in response to surveys, site-visits, and national conferences.

Useable data were defined for 13 sites with significant scrap metal inventories. Data-collection
methods for individual sites varied, however; while some sites have developed and maintained
                                         10-10

-------
databases that are current for scrap-metal inventories, others relied exclusively on historical
knowledge to determine quantities of scrap under their control.

With regard to the reliability of collected data, the MIN Scrap Metal and Equipment Team
offered the following statement:

       "... Data limitations include the following:  (1) no information was received on
       either scrap or equipment for [several DOE facilities]... (2) some information
       was submitted in summary form only, without site-specific breakouts; (3) some
       sites supplied complete information on some topics and partial or no information
       on others ... and (4) some data could not be tabulated because it was descriptive
       rather than quantitative or expressed in units inconsistent with the units used in
       this report and could not be readily converted (MIN 96)."

Noteworthy is the DOE's reference to "partial or no information" that shows that only about one-
fifth of the total scrap-metal inventory had been assessed for the presence of radioactive
contamination.  In other words, four-fifths ( 80%) of existing metal inventories have not been
assessed for radioactive contamination.  About 88% of the assessed fraction was determined to
be contaminated with radioactivity. This relationship was used to estimate the percentage of
contaminated scrap within the unassessed fraction of existing metal scrap.

In summary, the collective uncertainty of existing scrap metal quantities reflects the combination
of uncertainties contributed by the following:

  (1)  the uncertainty hi total scrap metal estimates as reported by individual DOE sites that in
       some instances were based solely on historical records;

  (2)  the large percentage of scrap (about 80%) that was "unspecified" with regard to
       radioactive contamination and the resultant need to apply a scaling factor derived from
       the 20% of scrap that had been assessed for contamination; and

  (3)  the variability of existing scrap metal inventories as a function of tune.

With regard to the third component of uncertainty, most sites reporting data for the MIN
initiative indicated that their inventories of existing scrap may be sold or otherwise dispositioned
on a routine basis. The extent of variation in inventories with time can, therefore, not be assessed
from the snapshot of inventories as currently reported.

On the basis of available data, the current estimated value of 171,000 metric tonnes of
contaminated scrap is considered not likely to differ from the true value by more than a factor of
                                          10-11

-------
two. Thus, lower- and upper-bound values of existing contaminated scrap metal are defined to
be 85,500 and 342,000 metric tonnes, respectively.
                   i
Uncertainties Regarding Future Quantities of Scrap Metal

Of the 13 sites with existing scrap-metal inventories, current data identified only nine sites as
future sources of scrap metal estimated to be about 925,000 metric tonnes (Table 10-1).

For individual DOIi sites, point estimates were largely derived from historical data pertaining to
design specifications of buildings, structures, and process equipment that have been slated for
decommissioning.

The level of uncertainty regarding future quantities of scrap metal is undoubtedly higher than that
of existing scrap-metal quantities. Compounding the shared uncertainty of simply quantifying a
known or suspected aggregate of metal components is the incomplete and uncertain
decommissioning  schedule on which future scrap metal estimates are based.

Assumptions regarding future political, social, and economic factors that may significantly
impact the current decommissionhig schedule cannot readily be factored into a discussion of
uncertainty.  A reduced scope of decommissioning activity is likely to result hi future scrap metal
quantities less than the currently-estimated value of 925,000 metric tonnes. Conversely, an
expanded decommissioning that extends beyond the nine DOE sites defined in this report would
be expected to significantly increase the projected scrap quantities beyond current estimates.

10.3   UNCERTAINTY FOR NORMALIZED RMEI DOSES AND RISKS

Normalized doses and risks for the RMEI are provided hi Chapter 7 for a total of 40
radionuclides. Normalized dose values are reported in units of mrem/yr per pCi/g activity
concentration in released scrap metal and normalized risks are defined hi units of lifetime cancer
risk per pCi/g activity concentration hi released scrap metal. In general, the analyses
demonstrate that the normalized doses for the RMEI could be higher by a factor of 5 to 50,
or lower by up to a factor of 100 to 500, depending on the radionuclide. The uncertainties
                 !>             '             i    ,   ..      , .
in the normalized risks  are similar, except that the possibility exists that the risks could be
zero for extremely low doses and dose rates.
When scrap metal containing nominal levels of residual radioactivity is released for unrestricted
recycling, human exposure may occur at discrete stages of the life cycle of scrap.  At each stage
of the life cycle, exposure may be dominated by  select radionuclides and exposure pathways that
                                         10-12

-------
affect certain individuals within the exposed population more than others. In preceding chapters,
dose estimates were derived by evaluating potential exposures for all life-cycle stages of scrap
and associated groups of exposed individuals. Population groups that were found to have the
highest normalized dose for a given radionuclide provided the basis for modeling the RMEI.
Selected as RMEIs are individuals within each categoiy of exposed individuals who, on the basis
of reasonable assumptions, could be expected to receive doses toward the high end of the
distribution of doses for members of each category. Reasonable assumptions required selecting
values or a range of values for specific model parameters.

The uncertainty in the normalized dose to the RMEI is principally due to the wide range of
potential values that may conceivably characterize each of several critical model parameters
used. Table 10-2 provides a summary of the uncertainties for select groupings of radionuclides
that are considered most limiting to RMEI exposures. For each radionuclide grouping, the
critical stage within the life cycle of scrap is identified along with dominant exposure
pathway(s), and an estimate of the "upper-end multiplier" and "lower-end divisor" is provided.
The upper-end multiplier and lower-end divisor define the potential range (and therefore the
uncertainty) of the normalized RMEI doses.  For example, Table 7-1 in Chapter 7 previously
identified the RMEI dose of 0.899 mrem/yr per pCi/g of Co-60 in released scrap metal. This
dose was estimated for a worker operating a metal lathe fabricated totally from scrap metal
containing the Co-60 contaminant at a concentration of 1 pCi/g. According to Table 10-2, this
normalized dose could be as much as a factor of five greater if the "metal product" was assumed
much larger than the modeled lathe and more time was spent by the individual in close proximity
to the product. Conversely, the normalized dose could be lower by as much as a factor of 100 if
(1) a smaller metal product, (2) shorter exposure times, and (3) a variable percentage (i.e., less
than 100%) of contaminated scrap used to produce the metal product were assumed.
                                              =>r-
The multipliers and divisors are largely based on professional judgment and are designed to
bracket estimated uncertainties and variabilities for normalized RMEI doses.  Detailed
explanations for upper-end multipliers and lower-end divisors for each of the six limiting life-
cycle stages identified in Table 10-2 are provided in Section 5.4.7 of Appendix L.
     »

A summary of multipliers and divisors that were defined  for all 40 radionuclides is displayed in
Figure 10-2 in the form of a bar chart. For most radionuclides, the range of uncertainty, as
defined by lower- and upper-bound values, spans about three to four orders of magnitude.
Normalized values for the RMEI are quantitatively defined by the upper end of the distribution
of values divided by the line that marks the boundary between upper- and lower-bound values.
                                         10-13

-------
                                            Table 10-2. Uncertainty/Variabilit.y in Normalized Individual Doses
RwJltmucildes
Zn-65*
Sb-125
Cs-134*
Cs-137*
Ni-59
Ni-63
Mo-93
Tc-99
Ac-227+D
Fe-55
Mn-54
Co-60
Ru-106
Ag-110m+D
Nb-94
Ce-144+D
Eu-152
Ra-226+D
Ra-228+D
Th-228+D
Pm-147
Th-229/230/232
Pa-231
TJ-234/235/238
Np-237
Pu-all
Am-241
Cm-244
Pb-210
C-14
1-129
Sr-90
LlmWng Stage
Scrap yard
Metal products
Slag pile
Slag pile
Mill
OfFsite exposure to
airborne emissions
Ground water
contaminated by slag
leachate
Primary Pathway
External
exposure
Inhalation
Soot
ingestion
External
exposure
External
exposure
Inhalation
Ingestion
Ingestion of food
Ground water
ingestion
Upper End
Multiplier
10
10
10
5
40
20
20
50
50
Loner End
Divisor
100
500
500
100
100
500
500
NAf
100
NAf
Busts
Upper end due to eliminating dilution factor.
Lower end due to additional dilution (30 fold), reduced occupancy and increased
distance (3).
Upper end due to eliminating dilution factor.
Lower end due to additional dilution (30 fold), reduced occupancy (2), and
reduced dust loading (10)
Upper end due to eliminating dilution factor.
Lower end due to additional dilution (30), reduced occupancy (2), and reduced
soot ingestion (10)
Upper end due to increase in size of component and occupancy time (5).
Lower end due to application of a dilution factor (30) and lower occupancy time
and smaller size component (3).
Upper end due to elimination of dilution factor (9) and increased occupancy time
and slag partition (4).
Lower end due to additional dilution (30) and smaller contaminated area and
occupancy time (3).
Upper end due to elimination of dilution factor (9) and increased occupancy time
and slag partition (2).
Lower end due to additional dilution (30), lower dust loading (10), and lower
occupancy time (2).
Upper end due to elimination of dilution factor (8) and increased occupancy time
and slag partition (2).
Lower end due to additional dilution (30), lower soot ingestion (10), and lower
occupancy time (2).
Upper end due to elimination of dilution factor (8), closer location (3), increased
intake of crops (2).
Lower end due to additional dilution (30), further distance (2), less intake (2).
Upper end due to less dilution in ground water.
Lower end due to elimination of ground water due to increased transit time, and
soot ingestion becomes the limiting pathway.
* These radionuclides partition to baghouse dust. If it is plausible for individuals to be exposed to reconcentrated stages of the metal recovery process for prolonged periods of time, the upper end
multiplier for these radionuclides could be as high as a factor of 100.

t A lower limit for these pathways in not applicable, since the lowest limiting dose will be due to a different pathway (see text).

-------
Ac-227+D
 Th-Series
Th-229+D
  U-Series
Pb-210+D
  Sr-90+D
  Th-232
  Pa-231
Np-237+D
Th-228+D
  Am-241
   Co-60
    1-129
  Pu-239
  Pu-240
  Pu-242
  Pu-238
  Cm-244
  Th-230
 Ag-110m
Ra-226+D
  U-Separ.
   Nb-94
Ra-228+D
  Eu-152
 U-235+D
U-Deplete
   U-234
 U-238+D
  Cs-134
   Mn-54
   Zn-65
Cs-137+D
  Sb-125
Ru-106+D
Ce-144+D
Pu-241+D
    C-14
  Pm-147
   Mo-93
   Tc-99
    Ni-63
   Fe-55
    Ni-59
          i i i mill i i mini i i i nun  t i until  i mini  i mini  I 11 mill  i mill  I I I linn  i limn I Mini
            1E-08       1E-06       1E-04      1E-02       1E+00      1E+02
      1E-09       1E-07       1E-05       1E-03       1E-01       1E+01

                RMEI  Dose (mrem/y per pCi/gm)

     Figure 10-2. Bounding Normalized RMEI Dose Values (mrem/y per pCi/g)

                                   10-15

-------
 10.4   UNCERTAINTY IN NORMALIZED COLLECTIVE DOSE ESTIMATES

Normalized collective doses were derived in the preceding chapter and are reported as
radionuclide-specific values in units of person-rem per Ci in released scrap. Important to note is
that the normalized collective dose does not contain a measure of time. The normalized
population dose represents the sum of all individual exposures for the entire exposed population,
for as long as the radionuclide can reasonably be assumed to result in human exposures. In
general, the results of the uncertainty analyses reveal that the collective doses could be
higher or lower than the estimates by a factor of between 2 and 3. The collective risks
could be higher by the same amount, but the possibility exists that the risks could be zero
for extremely low doses and dose rates.

When contaminated scrap metal is recycled, some radionuclide contaminants are predominantly
(and in some instances exclusively) partitioned in the finished metal from which new products
are made; based on chemical and physical properties, other radionuclides are more likely to be
found in slag, baghouse dust, or entrained in stack gases that are emitted into the ambient
biosphere. Figure 10-3 identifies each of the four dominant media associated with population
exposures and then- most limiting radionuclides.

Radionuclides contained in the four media have the potential for exposing large numbers of
individuals to doses that will vary from near zero up to the dose defined for the RMEI.
Individual doses may, therefore, span four or more orders of magnitude.

A second important characteristic of the collective^ose, and one which has great significance to
the discussion of uncertainty, is the fact that large (and undefined) variations among individual
exposures have little or no impact on the magnitude and variability of the collective dose. In
effect, most model parameters that define individual doses and their variability are
inconsequential to modeling the collective dose of an exposed population.
                                         10-16

-------
                 Partitioning     Media
Usage
Initial Re-Use
Fate     Final
       Disposition
NM « not hdudsd In model
W = value varied (seefexQ
                        Figure 10-3.  Collective Impact Calculational Approach
     This important generic feature of normalized collective dose is readily illustrated by the
     following example:

            From Figure 10-3, carbon-14 is identified to partition to flue gases that may be
            released to air.

            For Case #1, it is assumed that a one-Curie quantity of C-14 is released from a smelter
            stack located in a residential community under very stable atmospheric conditions with
            minimum dilution. In turn, this results in maximum air concentrations to the downwind
            population. However, offsetting the high ah- concentrations and maximally elevated
            individual doses is the fact that the total number of exposed individuals is at a minimum.
                                               10-17

-------
       For Case #2, the same one-Curie quantity when released under highly unstable
       atmospheric conditions results in maximum dilution. Correspondingly, this results in
       greatly reduced individual doses, but to a proportionately larger number of exposed
       individuals.


 For the two contrasting cases, the following conclusions are drawn:
                  i                                 •*'
         (1)   the average individual exposure in each of the two groups will be different
              (perhaps by orders of magnitude);

         (2)   the range of individual doses within each group will differ substantially; but

         (3)   the normalized collective dose will be essentially the same for both.conditions.

Since only a restricted number of parameters significantly impact the estimated normalized
collective dose, its level of uncertainty is proportionally restricted. Thus, the several orders of
magnitude spread in the RMEI dose estimate is likely to be reduced to just one order of
magnitude for collective dose estimates.


Model Parameters that Impact Uncertainty


Critical parameters that are likely to introduce uncertainty to collective dose estimates were
previously identified in Figure 10-3 and include the following:


       (1)    Partitioning Factors. For each radionuclide contained in scrap, fractional values
              were assigned to each of the four media based on physical and chemical
              properties.

       (2)    Usage Factors. With the exception of "air emissions," available data were
              assessed to best determine the various uses for each media. For example, it was
              determined that about 34.9% of slag is currently used for road base, 24.6% for soil
              conditioner, 23% for railroad ballast, etc.

       (3)    Initial Re-Use. This identifies specific exposure scenarios for which additional
              input parameters must be defined, as will be explained below.
                                          10-18

-------
       (4)     pate and Final Disposition. Initial re-use in some instances may not fully define
              the collective dose if the radionuclide persists beyond the useful life of the object
              and the object is recycled again.  Thus, an automobile that is retired after 10 years
              may again be recycled.  Due to natural decay, residual contamination may be
              greatly reduced, but may nevertheless  remain significant.

When combined, the variability of parameters needed to model normalized collective dose is not
likely to yield uncertainties greater than a factor of 10. To illustrate the types of parameters used
for modeling and their variabilities, the following illustration is offered.

Model Parameters Used to Estimate  Normalized Collective Dose for Co-60

Partition Coefficient. Cobalt has chemical and physical properties that closely resemble iron. A
review of the scientific literature confirms the fact that, as a  contaminant in scrap metal, Co-60 is
retained nearly 100% in the metal melt. For Co-60, the assigned partition coefficient of 100% to
finished steel is, therefore,  considered to have an uncertainty approaching zero.

Usage. Steel is used to make a wide variety of finished products; an exhaustive analysis of usage
is not practical. Data provided by the steel industry were used to define principal categories and
markets that employ recycled steel. From current data, information was gleaned that indicates:
(1) nearly half (i.e., 47.4%) of recycled scrap is used to produce automobiles; (2) 36.5% goes to
the production of structural components;  (3) 5.8% is consumed in the manufacturing of consumer
items inclusive of home appliances; and (4) 10.3% is  contained in items that  are "nonaccessible"
and will not contribute to human exposure.  The uncertainty in these values is defined by the
potential that current usage or appropriation of scrap metal may change in the future for reasons
such as a shift in national/global economics; new technologies pertaining to steel production and
manufacturing process; and substitution of steel for other metals/alloys or non-metallic materials.

Since there are major economic commitments to the steel industry and associated manufacturing
processes for consumer items, a significant deviation from current uses of scrap metal is
considered unlikely. The uncertainty in model usage factors is, therefore, considered
insignificant.
                                          10-19

-------
Initial Re-Use Exposure Scenarios

To develop normalized collective dose estimates, scenarios were required that characterize
individuals and their typical use of or exposure to automobile(s), home appliances, and office
building(s) (see Figure 10-3). These scenarios were described in detail in Chapter 9, Section 9.5,
and are briefly summarized here.

Automobile Scenario. The weighted average physical mass, dimensions, and density of primary
steel components that include engine, frame, and shell were derived from current data of
automobiles manufactured hi the United States. These data provided the input parameters for the
computer code MicroShield™ for the calculation of exposure rate to an occupant.
(MicroShield™ is considered a standard hi the industry and can be assumed to yield reliable
results for a given set of input parameters.) Average automobile occupancy was estimated at
1,460 person-hours per year and was based on two occupants for two hours every day.  The
automobile's effective life was assumed to be 7.3 years; however, it was further assumed that it
would be recycled and 100% of its scrap would be returned to the pool of scrap  used to
manufacture automobiles for a period of 1,000 years.

For the free release of scrap containing one-Curie of Co-60, the automobile scenario yielded a
normalized collective dose of 6,606  person-rem.  Uncertainty of this modeled output value is
principally limited to the accuracy of computer-generated (i.e., MicroShield™) dose rates and the
assumption regarding automobile occupancy. Of the two, automobile occupancy is likely to be
the larger contributor to the collective uncertainty. The analysis assumes an occupancy of two
persons per vehicle driving 730 hours per year. TheTlinalysis further assumes 730 hours of
driving per year at an average speed of 35 mph which results hi an average yearly mileage of
about 25,500 miles per vehicle. This is more than twice the national yearly average of 11,834
miles as reported for 1993 by the U.S. Department of Transportation and the Insurance Institute
for Highway Safety. Accordingly, the estimated collective dose of 6,606 person-rem resulting
from automobile use is likely to be conservative with an uncertainty of about a factor of two.

Appliance Scenario. Key home appliances with the potential for human exposure include
refrigerators, stoves, dishwashers, microwave ovens, trash compactors, washers, and dryers.
These items are commonly located hi or near the kitchen, where members of a household spend
considerable amounts of time.  The average exposure dose rate for this scenario was modeled by
                                         10-20

-------
means of the MicroShield™ computer code with input parameters that included average mass of
appliances, shielding factors, and average physical dimensions that define a kitchen/dining area.

To estimate kitchen occupancy times, a family of four was assumed. Furthermore, it was
assumed that all members of the family eat breakfast hi the kitchen seven days a week, and eat
dinner in the kitchen five days a week.  Weekday lunch was assumed to be eaten in the kitchen
by only one member of the family.  It was also assumed that one member of the family would
spend one hour each week night doing homework in the kitchen.  Based on these assumptions,
kitchen occupancy times of 70 person-minutes per day in the work area and 190 person-minutes
per day hi the dining area were calculated.

The normalized collective dose was estimated to be 340 person-rem. (This also includes 30
person-rem from cookware.) The uncertainty of this value is governed primarily by the 70 and
190 person-minutes occupancy times that were assumed typical for U.S. households. As average
values, these occupancy estimates are not likely  to be more than a factor of two (multiplied or
divided by two) different from actual current (and future) values.

Office Building Scenario.  An office building was selected to best represent the exposure
                                                       j
scenario from finished steel used in structural building components.  The model office building
was of modular design with six offices in each module. Dose rates from structural components
consisting of I-beams, steel columns, studs, rebar, etc. were again estimated by means of the
MicroShield™ computer code. Based on an occupancy factor of 2,000 hours per year (per
individual), 2.8 persons per 100 m2 of building floor space, and a 50-year building service life, a
normalized collective dose of 3,250 person-rem  was calculated.

The uncertainty in this estimate is considered to  be small but the dose estimate can be considered
to be conservative because the critical model parameters included the standard 2,000 hours per
year exposure period and the 2.8 persons per 100 m2 worker density values. The latter value was
obtained from the Commercial Building Energy Consumption Survey (CBECS) published by the
Energy Information Agency of the Department of Energy.  The survey, however, also reported
human population densities for the following:

       •       all commercial buildings: 1.04 persons per 100 m2,
       •       buildings used for lodging (hotels/motels):  0.82 person per 100 m2.
                                        10-21

-------
A building exposure scenario that represents a weighted average of all facilities might have
reduced the normalized collective dose by about a factor of two.

Summary Conclusions

Uncertainties in normalized collective doses derived in the TSD are believed to be relatively
small. The basis for this generic assumption is two-fold:

  (1)  The number and complexity of parameters needed to model collective dose are few in
      comparison to those required to model individual doses (e.g., RMEI).

  (2)  Collective model parameters are defined by values that are robust since they reflect
      average values that in most instances are readily defined in the literature.  -

In the above-cited example, the free release of scrap metal containing one-Curie of Co-60 was
estimated to yield a normalized collective dose of 1.02E+04 person-rein. This estimate was
based on the following contributions:
    Source/Scenario                Dose (person-rem)     Estimated Uncertainty Factor
    Automobile                        6.61E+03         2 lower and less than 2 higher
    Appliances (and cookware)           3.40E+02             < 2 higher and lower
    Office Building                     3.25E+03            2 - 3 higher and lower
                         Total        1.02E+04          about 2 higher and 3 lower


Analysis of principal model parameters for each of the three scenarios suggests a modest level of
conservatism that is unlikely to yield a collective dose uncertainty greater than a factor of two
higher or three lower.

10.5  UNCERTAINTIES REGARDING MINIMAL DETECTABLE CONCENTRATIONS
      FOR RADIONUCLIDE CONTAMINANTS

Demonstration of compliance with any future radiation protection standards governing the
release of scrap metal from nuclear facilities will require radiological measurements, most of
which involve use of field survey instruments.
                                        10-22

-------
Whether a particular instrument and measurement procedure is capable of detecting residual
activity at a certain fraction of a future regulatory limits is largely determined by its minimum
detectable concentration (MDC) values.

Chapter 8 of the TSD discusses the standard instrumentation that may be used in surveys for the
free release of scrap metal and identifies the following generic equations that are used to
calculate MDC values:

       *      Direct Measurement (Fixed Position)

             The MDC for direct measurements is calculated using the following equation:
where:
                          MDC -

3 + 4.65 ,
>
f

R 60
* , V^ T, V A
n 	 -4f / % Y *4"P 1 "4* 	 •*
T k / ^ J , % /
60 100
                                                                              (Eq. 10-1)
              MDC
              BR
              t
              60
              Y,
minimum detectable concentration (dpm/lOOcm2 )
detector background count rate (cpm)
count time (s)
conversion factor (s/min)
yield for emission I (ptcle-emitted/d)
detector efficiency for emission I (e/ptcle-emitted)
detector area (cm2 ).
              Surface Scanning

              The MDC for detection of large areas of contamination using surface scanning is
              calculated using the following equation:
                                         10-23

-------
                      MDC =
3
2 * T
60
+ 4.65
*(£r,
D „. ^ * T
* 60
*8) * A
100

* HF
                                                                              (Eq. 10-2)
where:
              MDC   =    minimum detectable concentration (dpm/100cm2)
              BR      =    detector background count rate (cpm)
              t        =    meter time constant (s)
              60      =    conversion factor (s/rnin)
              Yj      =    yield for emission I (ptcle-emitted/d)
              ef       =    detector efficiency for emission I (c/ptcle-emitted)
              A       =    detector area (cm2)
              HF      =    surveyor efficiency (%).

Inspection of Equations 10-1 and 10-2 identifies parameters that can be termed as either
"intrinsic" or "extrinsic." Intrinsic parameters affecting MDCs are defined by the manufacturer's
design of the instrument and include detector dimensions, window thickness., meter time
constant, and other design aspects that affect detection efficiency. Based on reasonable quality
assurance standards by me manufacturer and compliance with stated instrument specifications,
these parameters of MDC are unlikely to significantly impact the uncertainty of the calculated
MDC value.

In general, calculated MDCs, inclusive of those cited in the literature or specified by the
manufacturer, represent optimum instrument capabilities and instrument use under controlled
laboratory conditions. The uncertainty of a calculated MDC value must, therefore, be defined by
extrinsic differences mat define controlled laboratory measurements and field survey
measurements.

Extrinsic factors can be further categorized as operational parameters and conditional parameters.
Critical  operational parameters include the following:
                                         10-24

-------
       •      Instrument Selection and Radionuclide Identification. Survey instruments must
              be selected that properly reflect the radionuclide contaminants that are most likely
              to be encountered.

              In most cases, the identification of critical radionuclides should be straightforward
              since the nature of the operation at the plant is known. Half-lives of isotopes can
              be used to determine what may have decayed away and what is likely still to be
              present.

              For situations in which a fixed ratio between two radionuclides can be established
              throughout a cleanup unit, the measurement of one radionuclide can serve as a
              surrogate for the other. This might also be possible for more than two
              radionuclides if consistent ratios between them can be demonstrated.

       •      Instrument Calibration.  Accurate and reliable measurements of surface
              contamination requires proper selection of calibration standards. Calibration
              sources must be selected that accurately reflect the type/mix of radiation
              emissions and their energies.

       •      Source to Detector Distance.  A critical parameter affecting instrument efficiency
              and, therefore, MDC values is the accuracy and consistency of source to detector
              distance.  This is especially critical for instruments used in a "scanning" mode.

       •      Operator Experience.  Where contamination levels need to be checked over a fine
              spatial scale, all of the surface area should be measured by scanning, i.e., by
              passing a survey meter probe over the surface at a fixed rate and covering the
              entire area. The ability to measure a given level of radioactive contamination is,
              of course, affected by the detector's sensitivity, the particular radionuclide, and the
              subjective ability of the operator to discern a change in the reading either by
              visual or audible means.

       •      Choice of Operating Parameters. For stationary survey measurements, the
              duration of time for which counts are integrated is critical to the MDC. For
              scanning mode operations, the selection of instrument time constant and the
              scanning velocity are critical.

Conditional factors are those that deiGne the survey environment.  Most notable among these is
the ambient radiation background level at the survey location.  Equations 10-1 and 10-2 establish
the relationship between detector background count rate and MDC:  MDC increases in direct
proportion to the square root of the ambient background count rate.
                                         10-25

-------
For nominal background levels ranging from 50 cpm to 100 cpm, an assumed MDC for a
pancake probe is likely to be in close agreement with field-detection limits.  With the realization
that future scrap may have to be surveyed in radiologically controlled areas (RCAs) of a
contaminated facility, ambient background levels may be well above those of a laboratory setting
or assumed for the calculated MDC values. When actual ambient background count rates
significantly exceed assumed background values used to derive MDCs, the potential exists for
the free release of scrap metal that hi fact may not meet a stated release standard.  Figure 10-4
demonstrates the relationship between MDC values and ambient background.

A second conditional factor affecting MDC uncertainty involves surface materials and their
texture. An apriori calculation of the MDC assumes uniform, smooth, and flat surfaces.  For
surfaces that may be pitted, corroded, and three-dimensional, conversion of the surface emission
rate to estimates of residual contamination may yield values that significantly underestimate
actual residual contamination.

The uncertainty introduced by operational and conditional factors cited above on MDCs that are
calculated on an apriori basis are summarized in Table 10-3. The values represent
multiplicative factors to be applied to calculated MDCs. From these data, the following
conclusions should be drawn:
                                          \

  (1)   Before any survey measurements are performed that may result hi release of scrap, the
       survey instrument and measurement procedure to be used must be shown to possess
       sufficient detection capabilities relative to specified surface-contamination release limits.

  (2)   Based on uncertainty values cited hi Table 10-3, the detection limits of the survey
       instrument must be a fraction of the limit that is defined by the reciprocal of the highest
       uncertainty value (e.g., for direct measurement of beta-emitting radionuclide, the
       calculated MDC value of the instrument should be less than one-third (1/3) of the
       regulatory release limit).
                                         10-26

-------
   10000
i
2.
1
o
I
a
1
|
fE
s
    1000
     100
          10                 100                1000
                             AMBIENT BACKGROUND (cpm)
         Figure 10-4. Effects of Ambient Background on MDC Calculation
                                   10000
                    Table 10-3. Relative Range in MDCs*
       Direct measurement
       Scan - small area source
       Scan - large area source
1-3
1-7
1-4
 1-5
1-14
 1-7
 1-7
0.01 - 1
 1-8
    * The values are multipliers to be applied to the MDCs tabulated hi Chapter 8.
                                  10-27

-------
                                   REFERENCES
DOE 95      Gaseous Diffusion Facilities Decontamination and Decommissioning Estimate
             Report, prepared by G.A. Person, et al, Environmental Restoration Division, Oak
             Ridge, TN for U.S. Department of Energy, Office of Environmental Management,
             ES/ER/TM-171, December 1995.
                                                             *
EPA 95      U.S. Environmental Protection Agency, Guidance for Risk Characterization,
             Science Policy Council, February 1995.
                 :

SCA 95      Analysis of the Potential Recycling of Department of Energy Radioactive Scrap
             Metal, prepared by S. Cohen & Associates, Inc. for the U.S. Environmental
             Protection Agency, Office of Radiation and Indoor Air, August 1995.

HAZ 95      U.S. Department of Energy Scrap Metal Inventory Report for the Office of
             Technology Development, Office of Environmental Management, prepared by
             Hazardous Waste Remedial Actions Program for the Department of Energy,
             DOE/HWP-167, March 1995.

Little, C. A.   Development of Computer Codes for Radiological Assessments, In Radiological
             Assessment:  A Textbook on Environmental Dose Analysis. Editors: Till, J.E.
             and"Meyer, H.R., prepared by Oak Ridge National Laboratory, Oak Ridge, TN,
             for the U.S. Nuclear Regulatory Commission, NUREG/CR-3332, 1983.

MIN 96      Taking Stock: A Look at the Opportunities and Challenges Posed by Inventories
             from the Cold War Era, U.S. Department of Energy, Office of Environmental
             Management, DOE/EM-0275, January 1996.

NRC 94      National Research Council, Science and Judgment in Risk Assessment, National
             Academy Press, Washington, DC, 1994.
                                        10-28

-------
Page Intentionally Blank

-------
Page Intentionally Blank

-------
         REVIEW DEAFT
  TECHNICAL SUPPORT DOCUMENT

EVALUATION OF THE POTENTIAL FOR
   RECYCLING OF SCRAP METALS
    FROM NUCLEAR FACILITIES

  VOLUME 2 OF 3: APPENDICES A-F
            Prepared by:

       S. Cohen & Associates, Inc.
          1355 Beverly Road
        McLean, Virginia 22101
               Under

        Contract No. 68D20155
       Work Assignment No. 5-13
            Prepared for:

  U.S. Environmental Protection Agency
    Office of Radiation and Indoor Air
          401 M Street, S.W.
        Washington, D.C. 20460

            Martin Offutt
       Work Assignment Manager

            My 15,1997

-------
Page Intentionally Blank

-------
                                    VOLUME 2



                                 APPENDICES A-F



                                      Contents




Appendix A: Characterization of Scrap Metal Inventories at

U.S. Nuclear Power Plants	 A-l



Appendix B: Recycling of Aluminum Scrap ,	,. B-l



Appendix C: Recycling of Copper Scrap	C-l



Appendix D: Selection of Radionuclides for Radiological Impacts Assessment ..-	 D-l



Appendix E: Distribution of Radionuclides During Melting of Carbon Steel	E-l
                                                                                    »*


Appendix F: Distribution of Contaminants During Melting of Cast Iron	F-l

-------
Page Intentionally Blank

-------
                APPENDIX A

CHARACTERIZATION OF SCRAP METAL INVENTORIES
        AT U.S. NUCLEAR POWER PLANTS

-------
Page Intentionally Blank

-------
                             TABLE OF CONTENTS

1.0   INTRODUCTION	  AM

2.0   CHARACTERISTICS OF REFERENCE REACTOR FACILITIES	  A2-1
      2.1   Reference PWR Design and Building Structures	  A2-1
      2.2   Reference BWR Reactor Design and Building Structures	A2-5

3.0   INVENTORIES OF RESIDUAL RADIOACTIVITY IN REFERENCE   .
      REACTOR FACILITIES  	  A3-1
      3.1   Neutron Activated Reactor Components and Structural Materials	  A3-3
      3.2   Internal Surface Contamination of Equipment and Piping	  A3-6
            3.2.1   Measurements of Internal Surface Contaminants at Six Nuclear
                   Power Plants	  A3-7
            3.2.2   Internal Surface Contamination Levels Reported in
                   Decommissioning Plans	  A3-10
            3.2.3   Levels of Internal Surface Contamination Derived for Reference
                   BWR	  A3-13
            3.2.4   Levels of Internal Surface Contamination for Reference PWR  .  A3-19

4.0 ,  BASELINE METAL INVENTORIES  	  A4-1
      4.1   , Baseline Metal Inventories for Reference PWR	  .  A4-1
      4.2   Baseline Inventories for Reference BWR	  A4-2
      4.3   The Applicability of Reference Facility Data to the Nuclear Industry . . .  A4-5

5.0   METAL INVENTORIES SUITABLE FOR RECYCLING	  A5-1
      5.1   Identification of Steel Components Suitable for Recycling .	  A5-4
            5.1.1   Reference BWR	•.	  A5-5
            5.1.2   Reference PWR 	  A5-26
            5.1.3   Summary Estimates of Steel for Reference BWR/PWR and the
                   Commercial Nuclear Industry	  A5-36
      5.2   Metal Inventories Other Than Steel 	'. . .  A5-38
      5.3   Time-Table for the Availability of Scrap Metal from the
            Decommissioning of Nuclear Power Plants	  A5-39

REFERENCES  	  R-l

ADDENDUM #1	  ADM

ADDENDUM #2	  AD2-1

-------
                                 LIST OF TABLES

Table A3-1   Source of Residual Radioactivity and Curie Quantities at Reference BWR and
             PWR	   A3-2
Table A3-2   Estimated Radioactivity of Neutron-Activated Reactor Components
             in a BWR	   A3-3
Table A3-3   Estimated Radioactivity of Neutron-Activated Reactor Components in a
             PWR	 . !	   A3-5
Table A3-4   Activation Levels at Trojan Nuclear Plant 	  A3-5
Table A3-5   Comparison of Residual Radionuclide Inventories and Operating
             Parameters for the Six Nuclear Generating Stations Examined  	  A3-7
Table A3-6   Radionuclide Composition of Internal Surface Contamination	  A3-8
Table A3-7   Distribution in Percent of the Radionuclide Inventory Estimates for
             Three Pressurized Water Reactors 	  A3-9
Table A3-8   Internal Contamination Levels of Big Point Nuclear Plant at Shutdown   A3-10
Table A3-9   Plant Systems Radioactivity Levels at SONGS 1	   A3-11
Table A3-10  System Average Internal Contamination Levels for Yankee Rowe  ....   A3-13
Table A3-11  Relative Radionuclide Composition of Activated Corrosion Products of
             Reference BWR at Shutdown 	    A3-14
Table A3-12  Distribution of Activated Corrosion Products on Internal Surfaces in
             Reference BWR  	   A3-14
Table A3-13  Contact Dose Rate and Internal Radioactivity Deposition of BWR
             Piping 	   A3-15
Table A3-14  Estimates of Internal Contamination for Reference BWR Piping	   A3-16
Table A3-15  Summary of Internal Radioactivity Levels in BWR Equipment	   A3-17
Table A3-16  Estimated Internal  Radioactivity in BWR Systems	   A3-17
Table A3-17  Estimates of Internal Surface Contaminants in a Reference PWR
             Primary System	   A3-19
Table A3-18  Levels of Contamination and Estimated Quantities of Radioactive
             Corrosion Products Deposited on the Interior of PWR Reactor Systems   A3-20
Table A3-19  Non-RCS Contaminated Piping Data	-.	   A3-22
Table A3-20  Radionuclides in Primary Coolant that Contribute to External Surface
             Contamination in the Reference PWR	   A3-23
Table A3-21  Radionuclide Concentrations in Primary Coolant that Contribute to
             External Surface Contamination in Reference BWR	   A3-24
Table A3-22  Surface Contamination Levels for Reference BWR at Shutdown	   A3-26
Table A3-23  Estimated Structural Surface External Contamination in the Reference
             BWR	   A3-27
Table A3-24  Ranges of Radionuclide Associated with Highly Contaminated External
             Surfaces at Six Nuclear Generating Stations	   A3-29
Table A3-25  Inventory of External  Surface Contamination at Trojan Nuclear Plant .   A3-30
Table A3-26  Areal Surface  Contamination Levels Based on Survey Measurements
             at TNP Preparing for D&D	   A3-31
                                         11

-------
                         TABLE OF CONTENTS (Continued)

Table A4-1   Inventory Estimates of Materials Used to Construct a 1971-Vintage,  1,000
             MWe, Pressurized Water Reactor Facility	  A4-1
Table A4-2   Breakdown of Material Quantities Used in Plant Structures and Reactor
             .Systems	   A4-3
Table A4-3   Inventory Estimates of Materials Used to Construct a 1,000 MWe
             Boiling Water Reactor Facility  	  A4-4
Table A4-4   Summary of Total Metal Inventories Potentially Available for Recycling  A4-6
Table A5-1   Examples of Scrap Metal Grouping Based  on Contamination  	  A5-5
Table A5-2A  Reference BWR Steel Inventories by Location Within the Reactor
             Building	   A5-6
Table A5-2B  Reference BWR Steel Inventories for Locations Within the Radwaste
             Building	   A5-16
Table A5-2C  Reference BWR Steel Inventories by Location Within the Turbine
             Building  ..>.....	   A5-20
Table A5-2D  Reference BWR Piping Inventories by Plant Location 	  A5-24
Table A5-3A  Reference PWR Steel Inventories by Location Within the Reactor
             Building	   A5-26
Table A5-3B  Reference PWR Steel Inventories by Location Within the Auxiliary
             Building and Fuel Storage  	   A5-28
Table A5-3C  Reference PWR Structural Components of Auxiliary Building and Fuel
             Building	,	   A5-35
Table A5-4   Summary Data for Steel Inventories Potentially Suitable for Recycling   A5-37
Table A5-5   Summary of Metal Quantities Other than Steel 	  A5-38
Table A5-6   Time-Table for Available Scrap Metals from Decommissioned Nuclear
             Power Plants	   A5-39

                                                      '                     ;
                                LIST OF FIGURES

Figure A2-1   Pressurized Water Reactor  	  A2-3
Figure A2-2   Boiling Water Reactor	  A2-5
Figure A3-1   PWR Primary System Schematic  and Piping Data	  A3-21
Figure A5-1   Cumulative Availability  of Scrap  Metal from Nuclear Utilities 	  A5-41
                                         111

-------
Page Intentionally Blank

-------
                     CHARACTERIZATION OF SCRAP METAL
                 INVENTORIES AT U.S. NUCLEAR POWER PLANTS

1.0    INTRODUCTION

The U.S. commercial nuclear power industry is represented by 123  reactor plants.  At present,
eight reactors have been shutdown; and in the next two to three decades, most of the reactors
currently in operations will have reached their projected forty-year lifetime.

With the publication of the NRC's Decommissioning Rule in June 1988 (US NRC  1988),
owners and/or operators of licensed nuclear power plants are required to prepare and submit
plans and cost estimates for decommissioning their facilities to the NRC for review.
Decommissioning, as defined in the rule, means to remove nuclear facilities safely from
service and to reduce radioactivity to a level that permits release of the property for
                                               «
unrestricted use and termination of the license.  The decommissioning rule applies  to the site,
buildings, and contents and equipment.  Currently, several utilities have submitted a
decommissioning plan to the NRC for review.

Historically, the NRC has defined three classifications for decommissioning of nuclear
facilities:

   •    DECON is defined by the NRC as "the alternative in which the equipment, structures,
       and portions of a facility and site containing radioactive contaminants  are removed or
       decontaminated to a level that permits the property to be released for unrestricted use
       shortly after cessation of operations."

   •   SAFSTOR is defined as "the alternative in  which the nuclear facility is placed and
       maintained in a condition that allows the nuclear facility to be safely stored and
       subsequently decontaminated (deferred dismantlement) to levels that permit release for
       unrestricted use."

       The SAFSTOR decommissioning alternative provides a condition that ensures public
       health and safety, from residual radioactivity remaining at the site, without the need
       for extensive modification to the facility. Systems not required to be operational for
                                        Al-1

-------
       fuel storage, maintenance, and surveillance purposes during the dormancy period are to
       be drained, de-energized, and secured.

   •   ENTOMB is defined as "the alternative in which radioactive contaminants are encased
       in a structurally long-lived material, such as concrete; the entombed structure is
       appropriately maintained and continued surveillance is carried out until the radioactive
       material decays to a level permitting unrestricted release of the property."
                            i

Over the years, the basic concept of the three alternatives has remained unchanged.  However,
because of the accumulated inventory of spent nuclear fuel (SNF) in the reactor storage pool
and the requirement for about seven years of pool storage for the SNF  before transfer to dry
storage, the timing and steps in the process for each alternative have had to be adjusted to
reflect present conditions. For the DECON alternative, it is .assumed that the owner has a
strong incentive to decontaminate and dismantle the retired reactor facility as promptly as
possible, thus necessitating transfer of the stored SNF from the pool to a dry storage facility
on the reactor site.  While continued  storage of SNF in the pool is acceptable, the Part 50
license could not be terminated until the pool had been emptied, and only limited amounts of
decontamination and dismantlement of the facility would be required.  This option also
assumes that an acceptable dry transfer system will be available to remove the SNF  from the
dry storage facility and to place  it into licensed transport casks when the time comes for DOE
to accept the SNF for disposal at a high level waste repository.

In addition, the amended regulation stipulates that alternatives, which significantly delay
completion of decommissioning, such as use of a storage period, will be acceptable if
sufficient benefit  results. The Commission indicated that a storage period of up to 50  years
and a total  of 60  years between shutdown and decommissioning is a reasonable option for
decommissioning a light water reactor. In selecting 60 years as an acceptable period of time
for decommissioning of a nuclear power reactor,  the Commission considered the amount of
radioactive decay likely to occur during an approximate 50-year storage period and the time
required to dismantle the facility.

In summary, the need to adequately cool the high buraup assemblies from the final fuel core
in the pool for up to seven years and the regulatory requirements that critical support systems
be maintained in  operable conditions, the time between shutdown, decontamination,  and the
earliest date of dismantling efforts that would generate scrap metal is likely to be about 10
                                         Al-2

-------
years.  This interval may extend up to 60 years under the decommissioning alternative
SAFSTOR.  Through natural decay,  a longer time interval has the obvious impact of greatly
reducing radionuclide inventories. However, a simple inverse correlation between reduced
levels bf contamination and increased quantities of scrap metal suitable for recycling cannot
be inferred.  It is likely that for most scrap metal, the reduced contamination induced by
natural decay may merely impact the choice of decontamination method and/or
decontamination effort required to meet a future standard. For example, a storage period that
reduces beta/gamma surface contamination of 1.0 x 107 dpm/100 cm2 at 10 years post-
shutdown to 1.0 x 10s dpm/100 cm2 (i.e., a 100-fold reduction) would still require substantial
decontamination in order to meet current standards defined by NRC Regulatory Guide 1.86.
However, the reduced contamination level that is likely to be dominated by Cs-137  may affect
the method and level of effort required for successful decontamination.

The potential use of scrap metal for recycling is, therefore, dictated by the cost-effectiveness _
with which materials can be decontaminated to levels deemed acceptable for unrestricted use
(or for specified restricted use(s)). Estimates of scrap metal quantities must consider starting
levels of contamination and whether the contamination is surficial or volumetrically
distributed.

Residual radioactivity pertaining to xeactor components/systems and building structures is
generally grouped as (1) activation products that are distributed volumetrically, (2) activation
and fission products in the form of corrosion films deposited on internal surfaces, and (3)
contamination of external surfaces that result from the deposition of liquid and airborne
contaminants associated  with steam,  primary  coolant, and radioactive waste streams.

Most of the metal scrap  available from the complete dismantling of a nuclear power plant is
not expected to be radioactive. The non-radioactive scrap includes the large quantities of
structural metals and support systems that have not  been exposed to radioactivity during
reactor operations.  Conversely, some metal components will undoubtedly be contaminated so
as to render them unsuitable for recycling.
                                         Al-3

-------
Page Intentionally Blank

-------
2.0    CHARACTERISTICS OF REFERENCE REACTOR FACILITIES

A crucial factor affecting the quantity of metal and associated contamination levels is the
basic design of the reactor.  The two types of reactors used in the United States are the
pressurized water reactor (PWR) and the boiling water reactor (BWR). Of the 123 U.S.
reactor units, 40 are BWRs manufactured by General Electric (GE) and 83 are PWRs
manufactured by Westinghouse (W), Combustion Engineering (CE), and Babcock and Wilcox
(B&W).  Addendum #1 provides a complete listing of U.S. reactors along with demographic
data that includes projected year of shutdown.

In the 1976-1980 time frame, two studies were carried out for the U.S. Nuclear Regulatory
Commission by the Pacific Northwest Laboratory (PNL) that examined the technology, safety,
and costs of decommissioning large reference nuclear power reactor plants. Those studies,
NUREG/CR-0130 and NUREG/CR-0672 for a reference PWR and reference BWR,
respectively, reflected the industrial and regulatory  situation of the time.

To support the final Decommissioning Rule issued  in 1988, the earlier PNL studies have been
updated with the recent issuance of NUREG/CR-5884, Revised Analyses of Decommissioning
for the Reference Pressurized Water Reactor Station and NUREG/CR-6174, Revised Analyses
of Decommissioning for the Reference Boiling Water Reactor Power Station.  These four
NUREG reports along with several other NRC reports and select decommissioning plans on
file with the Commission represent the primary source of information used to characterize
Reference PWR and BWR facilities and to derive estimates of scrap metal inventories for the
industry at large.

2.1    Reference PWR Design and Building Structures

The facility described in this section is the 3,500 MWt (1,175 MWe) Trojan Nuclear Plant
(TNP) at Rainier, Oregon -„ operated by "the Portland General  Electric Company (PGE)
Designed by Westinghouse, this reactor is considered a typical pressurized water  reactor that
has been cited as Reference  PWR (NUREG/CR-0130; NUREG/CR-5884).

The NRC granted the operating license for the TNP on November 21, 1975, and  the plant
formally began commercial operation on March 20, 1976.  TNP's operating license was
scheduled to expire on February 8, 2011.  However, pn November 9,  1992, the TNP was
                                        A2-1

-------
shutdown when a leak in the "B" steam generator was detected and the licensee notified the
NRC of its decision to permanently cease operations in January 1993.  Following the transfer
of spent fuel from the reactor vessel to the spent fuel pool in May of 1993, TNP's Operating
license was reduced to a Possession Only license.  TNP's 17-year operating period
encompassed 14 fuel cycles and approximately 3,300 effective full-power days. In the
decommissioning plan submitted by PGE, the licensee has proposed the DECON approach
with a five-year delay period prior to decontamination and dismantlement.

In a PWR, the circulating primary coolant water is heated by the nuclear fuel core but is
prevented from boiling by a pressurizer, which maintains a pressure of about 2,000 psi. The
principal systems and components of the nuclear steam supply system are illustrated in Figure
A2-1. Components of interest are the reactor vessel, which contains the fuel and coolant and
the reactor coolant system (RCS).  The reactor vessel also contains internal support structures
(not depicted) that constrain the fuel assemblies, direct coolant flow, guide in-core
instrumentation, and provide some neutron shielding. The RCS consists of four loops for
transferring heat from the reactor's primary coolant to the secondary coolant system.  Each
loop consists of a steam generator, a reactor coolant pump, and connecting piping.  Steam
generated from secondary feedwater is passed through the turbine, condensed  back to water
by the condenser, and recycled.

Also included in the primary loop is a small side stream of water that is directed to the
chemical volume and control system  (CVCS).  This CVCS provides chemical and radioactive
cleanup  of the primary water through demineralizers  and evaporators. The primary water is
reduced in both pressure and temperature by the CVCS before being processed; therefore, the
CVCS is often referred to as the letdown system. The water processed through the CVCS is
returned to the primary water loops by the charging pumps. Note  that the primary water
processed through the CVCS is brought through the containment boundary or out of the
containment building, but the primary water providing the heat transfer to the steam
generators does not  pass through the containment boundary.

As shown in Figure A2-1, high levels of internally contaminated components  for a pressurized
water reactor are those associated with the primary coolant water system. Low-level
contamination of the secondary loop is a result of steam generator tube leakage in which
limited quantities  of primary coolant are introduced into the recirculating steam/water Other
                                         A2-2

-------
   major contaminated systems of PWRs not .shown in Figure A2-1 include the radioactive waste

   handling system and the spent fuel storage system.
                                          Containment
                                           Boundary i
Steam Jet
Air Ejector
 Steam
Generator
                                                        Chemical
                                                         Volume
                                                       and Control
                                                         System
                      Denotes Reactor Water System
                      or Radioactive Water
                                                                   Feedwater
                                                                     Pump
                                                                                       Cooling
                                                                                       Water
              *.^_.      Secondary
                 I     Makeup Water
                 Primary
               Makeup Water
                              Figure A2-1. Pressurized Water Reactor


    The principal structures requiring decontamination for license termination at the Reference

    PWR power station are the (1) Reactor Building, (2) Fuel Building, and (3) Auxiliary

    Building.  In addition to housing major plant systems, all three buildings contain

    contaminated systems and substantial quantities of contaminated structural metals that are

    potentially available for recycling.


    Reactor Building. The reactor building houses the nuclear steam supply system.  Since its
    primary purpose is to provide a leak tight enclosure for normal as well as accident conditions,

    it is frequently referred to as the containment building. Major interior structures include the

    biological shield, pressurizer cubicles, and a steel-lined refueling cavity.  Supports for
                                               A2-3

-------
equipment, operating decks, access stairways, grates, and platforms are also part of the
containment structure internals.

The Reactor Building is in the shape of a right circular cylinder approximately 64 m tall and
22.5 m in diameter.  It  has a hemispherical dome, a flat base slab with a central cavity and
instrumentation tunnel.
                 t

The Fuel Building - approximately 27 m tall and 19 by 54 m wide - is a steel and reinforced
concrete structure with four floors. This building contains the spent-fuel storage pool and its
cooling system, much of the Chemical and Volume Control System, and the solid radioactive
waste handling equipment. Major steel structural components include fuel storage racks  and
liner, supports structures for fuel handling, and components, ducts, and piping associated with
air conditioning, heating, cooling, and ventilation.
                 w  ,
The Auxiliary Building - approximately 30 m tall and lateral dimensions of 19 by 35 m - is a
steel and reinforced concrete structure with two floors below grade and four floors above
grade. Principal systems contained in the Auxiliary Building include the liquid radioactive
waste treatment systems, the filter and ion exchanger vaults, waste gas treatment system, and
the ventilation equipment for the Containment, Fuel, and Auxiliary Buildings.
                 ,'i
Other major building structures with substantial inventories of metals include the Control
Building and Turbine Building. The principal contents of the Control Building are the reactor
control room, process and personnel facilities.  The principal systems contained in the Turbine
Building are the turbine generator, condensers, associated power production equipment, steam
generator auxiliary pumps, and emergency diesel generator units.

Barring major system failure(s) (e.g., steam generator failure) most scrap metal derived from
these systems can be assumed to be free of contamination and can, therefore, be excluded
from estimates of scrap metal inventories.
                                         A2-4

-------
2.2    Reference BWRJReactor Design andJBuilding Structures


The 3,320 MWt (1,155 MWe) Washington Public Power Supply System (WPPSS) Nuclear

Project No. 2 located near Richland, Washington has historically been cited as the Reference
BWR facility (NUREG/CR-0672; NUREG/CR-6174).
                                      •)                         f              '

The design of a BWR (see Figure A2-2) is simpler than a PWR inasmuch as the reactor

coolant water is maintained near atmospheric pressure and boiled to generate steam.  This
allows the coolant water to directly drive the turbine. Thereafter, the steam is cooled in the

condenser and returned to the reactor vessel to repeat the cycle. In a BWR, the contaminated
reactor'coolant  comes in contact with most major BWR components, including the reactor

vessel and water piping, steam turbine, steam condenser, feedwater system, reactor water
cleanup system, and steam jet air ejector system.  As with the PWR, other  major

contaminated reactor systems include the radioactive waste treatment system and spent fuel

storage system.
                                      Steam Jet
                                      Air Ejector
  Reactor
  Vessel
                                                                    Cooling
                                                                    Water
               Containment
                Boundary
Reactor
 Pump
                                     Denotes Reactor Water Sy$tem
                                     or Radioactive Water
                           Figure A2-2.  Boiling Water Reactor
                                          A2-5

-------
The principal buildings requiring decontamination and dismantlement in order to obtain
license termination, at the reference BWR power station are the Reactor Building, the Turbine
Generator Building, and the Radwaste and Control Building.  These three buildings contain
essentially all of the activated or radioactively contaminated material and equipment within
the plant.

The Reactor Building contains the nuclear steam supply system and its  supporting systems It
is constructed of reinforced concrete capped by metal siding and roofing supported by
structural steel.  The building surrounds the primary containment vessel, which is a free-
standing steel pressure vessel. The  exterior dimensions of the Reactor Building are
approximately 42 m by 53 m in  plan,  70 m above grade, and  10.6 m below grade to the
bottom of the foundation.

The Turbine Building, which contains the power conversion system equipment and supporting
systems, is constructed of reinformed concrete capped by steel-supported metal siding and
roofing.  This structure is approximately 60 m by 90 m in plan and 42.5 m high.

The Radwaste and Control Building houses,  among other systems:  the  condenser off-gas
treatment system, the radioactive liquid and solid waste systems, the condensate demineralizer
system, the reactor water cleanup demineralizer system, and the fuel-pool cooling and cleanup
demineralizer system.  The building is constructed  of reinforced concrete, structural steel, and
metal siding and roofing. This structure is approximately 64 by 49 m in plan, 32 m  in overall
                                                             ^
height, and stands as two full floors and one partial floor above the ground floor.
                                         A2-6

-------
3.0    INVENTORIES OF RESIDUAL RADIOACTIVITY IN REFERENCE REACTOR
       FACILITIES

Significant quantities of radioactivity remain in a nuclear power station following reactor
shutdown even when all spent nuclear fuel has been removed. Neutron-activated structural
materials in and around the reactor pressure vessel contain the majority of residual radio-
activity in relatively immobile condition.  Lesser quantities involve radioactive corrosion
products  and fission products from failed fuel, which are transported throughout the station by
the reactor coolant streams.  The origin and mobility of radioactive nuclides following reactor
shutdown leads to grouping of residual radioactivity into five categories of different binding
matrices.  These categories include:
                                                          >

   1.   Activated Stainless Steel - Reactor internals, composed of Type 304 stainless steel,
       become activated by neutrons from the core.  Radionuclides have very high specific
       activities and are immobilized inside the corrosion-resistant metal.
                                                              >
   2.   Activated Carbon Steel - Reactor pressure vessels are made of SA533 carbon steel that
      , becomes activated by neutrons  bombardment.  The specific activities are considerably
       lower than in the stainless steel internals; and the binding matrix is much less
       corrosion resistant.

   3.   Activated Structural Steel. Steel Reban and Concrete. In the reactor cavity, these
       components become activated from neutrons escaping from the reactor vessel.       /
       Significant activation occurs along approximately 15 feet of the reactor cavity
       vertically centered on the reactor core and to a depth of about 16 inches  in the
       concrete.
                                  i
   4.   Contaminated Internal Surfaces of Piping and Equipment - Activated corrosion and
               »
       fission  products from the fuel travel through the reactor coolant water system
       throughout the radioactive liquid systems in the plant.  A portion forms a hard metallic
       oxide scale on the inside surfaces of pipes and equipment.

   5.   Contaminated External Surfaces - Surfaces may become contaminated over the lifetime
                        »
       of the plant primarily from leaks, spills and airborne migration of radionuclides
                                         A3-1

-------
       contained in the reactor coolant water (RCW).  The specific activity of RCW is low,
       but the contamination is easily mobilized and may be widespread.

All of the neutron activated metals/materials are contained in the reactor pressure vessel,
vessel internals, and structural components inside and within the concrete biological shield.

Total quantities and relative radionuclide composition of residual radioactivity are not only
affected by reactor design (i.e., BWR versus PWR) but are also strongly influenced by
numerous other factors inclusive of (1) fuel integrity,  (2) rated generating capacity  and total
years of operation, (3) composition of metal alloys, of reactor components and reactor coolant
system, (4) coolant chemistry and water control measures, and (5) the performance of or
failures of critical  systems and their maintenance over the 40 years of operation.

Table A3-1 provides summary estimates of typical residual activities for each of the five
major source categories.  Inspection of data reveals the volumetrically activated stainless steel
represents the overwhelming percentage of residual radioactivity.  Much lower quantities are
represented by volumetrically activated carbon steel and internal and external surface
contamination consisting of activation and fission products.  A more detailed discussion of
residual radioactivity by source category is given below.

                  Table A3-1,  Source of Residual Radioactivity and Curie
                         Quantities at Reference BWR and PWR
Source Category
Activated Stainless Steel
Activated Carbon Steel
Activated Structural Components,
Rebar, Metal Plates, I-Beams
Internal Surface Contamination of
Piping and Equipment
External Contamination of Equipment
Floors, Walls, and Other Surfaces
- Q«antitie!s"(Ci}
BWR00
6.6 x 106
2.9 x 103
1.2 x 103
8.5 x 103
1.1 x 102
, ' PWR^
4.8 x 106
2.4 x 103
1.2 x 103
4.8 x 103
(1.1 x 102)  NUREG/CR-0130
  (e)  Implied Value - NUREG-1496
                                         A3-2

-------
3.1    Neutron Activated Reactor Components and Structural Materials

Contamination of reactor components and structural materials by neutron activation is the
result of normal reactor operation.  The interaction of neutrons with constituents of stainless
steel, carbon steel, and concrete in and around the reactor vessel results in large quantities of
in-situ radioactivity.  Major radionuclides include Cr-51, Mn-54, Fe-55, Fe-59, Co-58, M-59,
and Ni-63. The activity concentration or buildup of a particular radionuclide among materials
subject to neutron activation is highly variable and depends upon (1) the concentration of the
parent isotope and its neutron cross-section, (2) the physical half-life of the radioactive
specie(s),  (3) the neutron flux level at that location, and (4) the duration of neutron exposure.

Reference BWR.  The average radioactivity concentrations and estimated quantities of
radioactivity for Reference BWR structural components with significant amounts of neutron
activation are listed in Tables A3-2.

                Table A3-2.  Estimated Radioactivity of Neutron-Activated
                             Reactor Components in a BWR
1 vf f 'f
Component {Quantity)
Core Shroud (1)
Jet Pump Assembly (10)
Reactor Vessel (1)
Cladding
Shell Wall
Steam .Separator Assembly (1)
Shroud Head Plant
Steam Separator Risers
Top Fuel Guide (1)
Orificed Fuel Support (193)
Core Support Plate (1)
Incore Instrument Strings (55)
Control Rod (185)
Control Rod Guide Tube (185)
TOTAL
Volume-Averaged
Radioactive
Concentration.
'{Ci/m3) - _:
1.68 x 106
2.62 x 104
1.07 x 103
1.12x 102
1.03 x 104
2.53 x 103
9.71 x 104
1.01 x 103
2.56 x 102
7.67 x 10s
5.11 x 10s
2.16 x 102

-Estimated.
Total' '
Radioacfivity
. (Ci)
6.30 x 106
2.00 x 103
2.16 x 103
9.60 x 103
3.01 x 104
7.01 x 102
6.50 x 102
1.10 x 104
1.78 x 10s
9.47 x 101
6.55 x 106
             Source: NUREG/CR-0672
                                         A3-3

-------
The Reference BWR reactor vessel is fabricated of SA533 carbon steel about 171-mm thick
and is clad internally with 3  mm of Type 304 stainless steel. .The total mass of the empty
vessel is about 750 metric tons.  The major internal components include the fuel core support
structure; steam  separators and dryers; coolant recirculation jet pumps; control rod guide
tubes; distribution piping for feedwater, core sprays, and liquid control; in-core
instrumentation and miscellaneous other components. Collectively, these internals, made of
stainless steel, represent about 250 metric tons

Reference PWR.  The right circular cylinder of the Reference PWR is constructed of carbon
steel, about 216 mm in thickness and is clad on the inside with stainless steel or Inconel
having a thickness of about 4 mm.  The approximate dimensions of the vessel are 12.6 m
high,  4.6 m outer diameter.   The vessel weighs about 400 metric tons.

The vessel internal structures support and constrain the fuel assemblies, direct coolant flow,
guide in-core instrumentation, and provide some neutron shielding. The principal components
are: the lower core support assembly,  which includes the core barrel  and shroud, with
neutron shield pads and the lower core plate and supporting structure; and the upper core
support and in-core instrumentation support assemblies.   These structures are made of 304
stainless steel and have a total weight of about 190 metric tons.

Based on 40 years of facility operation that assumes 30  effective full-power years (EFPY) of
reactor operation, the total number of curies contained in the activated vessel and internals is
estimated at 4.8 million curies (Table A3-3). Extra-vessel materials subject to significant
neutron activation (—10 Ci) includes the reactor cavity steel liner and  a limited quantity  of
reinforcement steel (rebar).  Additionally, about 1,200 Ci of radioactivity are estimated for the
concrete bioshield.
                                          A3-4

-------
                Table A3-3.  Estimated Radioactivity of Neutron-Activated
                             Reactor Components in a PWR
' .
Component
Shroud
Lower 4.7 m of Core Barrel
Thermal Shield
Vessel Inner Cladding
Lower 5.02 m of Vessel Wall
Upper Grid Plate
Lower Grid Plate
TOTAL
Voluttie-Avera^ed
. , Radioactive
Concentration ,
(Ci/m3)
2.97 x 106
3.07 x 105
1.45 x 10s
7.73 x 103
9.04 x 102
4.20 x 104
1.12 x 106

Radioactivity •
.per
, Component
(Ci)
3.43 x 106
6.52 x 105
1.46 x 105
l.SOx 103
1.76 x 104
2.43 x 104'
5.53 x 10s
4.82 x 106
       Source:  NUREG/CR-0130

The projected estimates of Table A3-4 in behalf of the Reference PWR (i.e., Trojan Nuclear
Plant) made in/1978 can be compared to the more current estimates contained in that plant's
Decommissioning Plan (submitted to the NRC in 1996).  Table A3-4 identifies revised
calculated inventories of activation products for 1993 or one year after shutdown.  The
recalculated value of about 4.2 million curies is about 13% lower than the original estimate of
4.8 million curies and principally reflects the difference between  17 years of actual plant
operation and the initial projection of 40 years.

      Table A3-4.  Activation Levels at Trojan Nuclear Plant (one year after shutdown)
^
;' Location
Reactor Vessel
Reactor Vessel Internals
Vessel Clad and Insulation
Bioshield Wall
Total
Activity
Terabeiequerels
230
154,000
880
31
155,000
- Ccaies
6,200
4,160,000
23,700
830
4,190,000
                                         A3-5

-------
The considerably higher calculated radioactivity values for a Reference BWR primarily reflect
the larger size and weight of the vessel and its internals.

For both PWR and BWR plants, the range of activity concentrations among individual reactor
components at time of shutdown is likely to vary over several orders of magnitude.1
Nevertheless, even those components with the lowest activity concentrations would still
exceed residual activity levels  far in excess of any  conceivable levels that would permit
recycling for unrestricted use.  (Note: At a density  of 8,000 kg/m3, a cubic meter of steel
containing 1 curie represents a specific activity concentration of 275,000 pCi per gram.)
Furthermore, these components also exhibit high levels of interior surface contamination.
While surface contamination is potentially removable, the volumetrically-distributed activation
products are not.
  For this reason, the reactor vessel and all internal components identified in Tables A3-2
  and A3-3 must be excluded from plant material inventories with regard to recycling.
  Excluded for similar reasons are select metal components used for structural support and
  reinforcement (i.e., rebar, I-beams, and floor and reactor cavity liner plates) that exhibit
  significant levels of activation products.
Scrap metal that has the potential for recycling is, therefore, limited to reactor systems and
structural components where contamination is limited to interior and exterior surfaces.

3.2    Internal Surface Contamination of Equipment and Piping

Activated corrosion products from structural materials in contact with the reactor coolant and
fission products from leaking fuel contribute to the presence of radioactivity in reactor coolant
streams during plant operation.  Although most of these radionuclides are removed through
filtration and demineralization by a plant's chemical and volume control systems (letdown
cleanup system), a smaller component escapes removal.  With time, some of the
radionuclides, principally the neutron-activated insoluble corrosion products, tend to deposit
on inner surfaces of equipment and piping systems. The metal oxide layer consists primarily
of iron, chromium, and nickel with smaller, but radiologically significant, quantities of cobalt,
manganese, and zinc.  This section characterizes the mixture of internal surface contaminants
and their relative distribution within major components associated with BWR and PWR power
plants.

                                         A3-6

-------
3.2.1   Measurements of Internal Surface Contaminants at Six Nuclear Power Plants

In a 1986 NRC study, three PWRs and three BWRs were assessed with regard to residual
inventories and distributions of long-lived radionuclides following plant shutdown
(NUREG/CR-4289).  Residual concentrations in the various plant systems decreased in the
following order: (1) primary  coolant loop, (2) radwaste handling system, and (3) secondary
coolant loop in PWRs and condensate system in BWRs.  Table A3-5 identifies total estimated
contamination inventories for the six reactor facilities examined, as well as the electrical
'ratings, and the approximate number of operational years for the units at the time of inventory
assessments.  The operational periods ranged from 8.3 years for Turkey Point Unit 3 to
slightly over  18 years for Dresden Unit One.

       Table A3-5.  Comparison of Residual Radionuclide Inventories and Operating
                   Parameters for the Six Nuclear Generating Stations Examined*
Stations
Humboldt Bay
Dresden- 1
Monti cello
Indian Point- 1
Turkey Point-3
Rancho Seco
Total Inventory
- (Curies)
600
2,350
514
1,050
2,580
4,470
Years of
Operation
13
18.3
10
11
8.3
8.8
MWe
63
210
550
170
660
935
Reactor Type
' r'
BWR
BWR
BWR
PWR
PWR
PWR
   Source:  NUREG/CR-4289
       Inventory includes radionuclides with half-lives greater than 245 days (Zn-65);
       inventory estimates do not include the highly activated metal components of the
       reactor pressure vessel and internals and activated concrete.
The relative radionuclide composition of internally contaminated surfaces observed at the
reactor sites also showed Considerable variation (Table A3-6). Fluctuations in compositions
were influenced by numerous factors including:, (1) the elapsed time since reactor shutdown;
(2) rated generating capacity; (3) materials of construction for the operating systems; (4)
reactor type, e.g., PWR versus BWR; (5) coolant chemistry and corrosion control; (6) fuel
integrity during operations; and (7) episodic equipment failure and leakage of contaminated
liquids.
                                         A3-7

-------
       Table A3-6.  Radionuclide Composition of Internal Surface Contamination*
Radio-
nuclide
Mn-54
Fe-55
Co-57
Co-60
Ni-59
Ni-63
Zn-65
Sr-90
Nb-94
Tc-99
Ag-llOm
1-129
Cs-137
Ce-144
TRLT
Total Plant
Inventory
(Curies)
„„„„,„,. , '"KA £f ,'«*
Composition in Percent of Total At

% •***
BWRs
Humboldt
Bay
3
90
—
6
—
0.2
—
0.004
< 0.004
3x 10-4
—
< 3 x 10-*
0.5
—
0.005

596
Dresden-1
0.9
28
—
46
0.09
5
19
0.007
< 0.003
4 x 10'5
—
< 1 x lO'5
0.04
1
0.1

2,350
Monticello
1
1
—
11
—
0.04
84
0.002
<0.1
8 x 10'5
—
• < 1 x 10^
2
—
0.008

448
stlvity Decay Corrected to Shutdown
X> '>.•' •** f 
-------
Inventories only include the radioactive contamination of corrosion film and crud on surfaces
of the various plant systems and do not include the highly activated components of the
pressure vessel.  The most abundant radionuclides included Mn-54, Fe-55, Co-58, Co-60, and
Ni-63.  Zinc-65 was present in relatively high concentrations in BWR corrosion film samples.
However, Fe-55 and Co-57/Co-60 were the most abundant radionuclides at all stations except
Monticello.  These two radionuclides constituted over 95% of the estimated inventories at
Humboldt Bay and Turkey Point.  At Indian Point Unit One, Dresden Unit One, Turkey Point
Unit Three, and Rancho Seco, they accounted for 82, 74, 98, and 70%, respectively,  of the
total estimated inventory. Although Fe-55 and Co-60 accounted for the majority of the
inventory (greater than 60% at five of the six stations), the relationship between the two
radionuclides was quite variable.

The transuranic radionuclides (Pu-238, Pu-239, Pu-240, Am-241, Cm-242, and Cm-244)
constituted percentages of the total inventory  ranging from 0.001% at Rancho Seco to 0.1% at
Dresden Unit One.
Secondary coolant loops in PWRs and condensate systems in BWRs contained much lower
radionuclide concentrations than observed in primary loop or feedwater samples.  Typically,
radionuclide concentrations were two or more orders of magnitude lower in secondary system
samples.

As expected, the steam generators contained the single largest repository of internally
deposited radionuclides at the PWR stations examined (Table A3-7).  The percentages of the
total residual radionuclide inventories in the steam generators were 77, 89, and 94% for
Indian Point One, Turkey Point Unit 3, and Rancho Seco, respectively.  The other repository
of significance in a PWR is the radwaste system, which typically contained 5 to 10% of the
total residual inventory.

             Table A3-7. Distribution in Percent of the Radionuclide Inventory
                        Estimates for Three Pressurized Water Reactors

Steam Generators
Pressurizer
RCS Piping
Piping (Except RCS)
Secondary Systems
Radwaste
Turkey Point-2
89
0.5
0.9
<0.01
0.1
9.2
Indian Point- 1
77
0.5
2.6
14
0.2
7
Rancho Seco
94
0.33
0.71
<0.01
0.05
5
PWR Average
86.7
0.4
1.4
4.7
0.1
7.1
       Source:  NUREG/CR-4289
                                         A3-9

-------
3.2.2  Internal Surface Contamination Levels Reported in Decommissioning Plans
Due to premature shutdown or projected shutdown within the next few years, only a small
number of commercial reactor facilities have submitted a Decommissioning Plan to the NR.C
for review. Summarized below are system-specific infernal contamination levels reported for
one BWR and two PWRs.

Big Rock Point Nuclear Plant.  The Big Rock Point Nuclear Plant is a small (67 MWe) BWR
designed by the General Electric Company and  constructed by Bechtel Power Corporation.
               , i' ,|i<                    ii      <.    *
Owned and operated by Consumers Power Company, the plant started commercial operation
in March of 1963 and is projected for shutdown in the year 2000. Table A3-8 presents
summary data of systems internally contaminated.

     Table A3-8. Internal Contamination Levels of Big Point Nuclear Plant at Shutdown
System
Liquid Rad Waste Tanks
Nuclear Steam Supply
RDS
Main Steam System
Fuel Pool
Liquid Radwaste System
Condensate System
Contamination
(dpjtrt/100 cm2)
3.0E+10
9.0E+09
3.0E+09
4.0E+08
4.0E+08
4.0E+08
5.0E+07
System
Resin Transfer System
Off-gas System
Control Rod Drive
Rad Waste Storage
Fuel Handling Equip
Heating & Cooling Sys

Contamination
(dpm/100 cm2)
3.0E+07
3.0E+07
6.0E+06
9.0E+05
7.0E+05
3.0E+05

San Qnofre Nuclear Generation Station Unit 1 (SONGS \\  SONGS 1 is a 436-mwe PWR
reactor that started operation in 1968.  As a result of an agreement with the California Public
Utility Commission, operation of SONGS 1 was permanently discontinued on November 30,
1992 at the end of Fuel Cycle No. 11. A preliminary Decommissioning Plan,  submitted to
the NRQ on December 1, 1992, proposed to maintain SONGS 1 in safe storage until the
permanent shutdown of SONGS 2 and 3.  SONGS 2 and 3 are licensed to operate until the
year 2013.
                                       A3-10

-------
In support of the SONGS 1 Decommissioning Plan, scoping surveys and analyses were
performed that supplemented an existing radiological data base.  The containment building,
fuel storage building, and the radwaste/auxiliary building were identified as the principal
structures containing significant levels of radioactivity within plant systems. Systems were
grouped by contamination levels defined .as (1) highly contaminated, (2) medium-level
contaminated, and (3) low-level contaminated.  Based on total radionuclide inventories and
surface areas, an average contamination level for each of the three groupings was derived
(Table A3-9).

               Table A3-9. Plant Systems Radioactivity Levels at SONGS 1
Plant Systems.
/ <• !• f
Hishlv Contaminated Svstems:
LDS Letdown
PAS Post Accident' Sampling System
PZR Pressurizer Relief
RCS Reactor Coolant
RHR Residual Heat Removal
RSS Reactor Sampling
SFP Spent Fuel Pool Cooling
VCC Volume Control
Total for Highly Contaminated Systems
Medium Level Contaminated Svstems:
BAS Boric Acid
CWL Containment Water Level
RCP RCP Seal Water
RLC Radwaste Collection
RMS Radiation Monitoring
RWG Radwaste Gas
RWL Radwaste Liquid
• CRS (Containment Spray) Recirc
SIS Safety Injection
Total for Medium Level Contaminated
Systems
r Estimated Total
Surface Area ,

1.26E+08cm2
•
1.25E+08 cm2
Estimated Level
of Radioactivity
(dpm/100 cm2)

3.6E+09

1.9E+06
Estimated
System Total. ,
Activity

2.08E+03 Ci
s.
1 f
1.08E+01 Ci
                                         A3-11

-------
         Table A3-9.  Plant Systems Radioactivity Levels at SONGS  1 (Continued)
' '
Plant Systems
* "* J A j-
Low Level Contaminated Svstems:
AFW Auxiliary Feedwater
CCW Component Cooling
CND Condensate
SHA Sphere Hydrazine Addition
CSS Condensate Sampling
CVD Condensate Vents & Drains
CVI Cryogenics
CWS Circulating Water
FES Flash Evaporator
FPS Fire Protection
FSS Feed Sampling
FWH Feed Water Heaters
FWS Feedwater
MSS Main Steam
MVS Miscellaneous Ventilation
PSC Turbine Sample Cooling
SDW Service Water
SWC Salt Water Cooling
TCW Turbine Cooling
Total for Low Level Contaminated
Systems
Estimated Total
Surface Area i
•*




















3.18E+08 cm2

Estimated Level
of Badioactivity
(dpm/IOOcm4)
















,



8.3E+03

Estimated
System Total ,
Activity




















1.21E-02 Ci

Yankee Rowe.  Yankee Rowe is a 167-mwe PWR with a startup date of August 19, 1960.  It
started commercial operation in July of 1961  and was" shutdown in October of 1991 following
21 fuel  cycles and 8,052 EFPD. In the 1993  decommissioning plan submitted to the NRC,
systems with significant internal surface contamination were identified as shown in Table
A3-10.

In reviewing the data of facilities that have submitted decommissioning plans, only limited
conclusions can be drawn to derive Reference values due to issues that relate to (1) their
limited  years of operation, (2) abnormal events and operating conditions that prompted
premature shutdown, and/or (3) size and design of the facilities.
                                        A3-12

-------
                                                                 1
       Table A3-10.  System Average Internal Contamination Levels for Yankee Rowe
: - System
Main Coolant
Spent Fuel Cooling
Waste Disposal
Primary Plant Vent &
Drain
Charging & Volume
Control
Shutdown Cooling
Fuel Handling
Letdown/Purifi cation
" CofltamiflMion
(dpm/100 cm*) <
7.1E+09
3!3E+08
1.2E+07
1.2E+07
1.2E+07
1.2E+07
1.7E+06
1.4E+06
System ' "J
Primary Plant Sampling
Safety Injection
Safe Shutdown
Vol. Control Heating &
Cooling
Vol. Control Vent. &
Purge
Post Accident H2 Control
Chemical Shutdown

Contatnitiatiott
JtdptB/lQQ'Cax*)
1.4E+06
1.4E+05
1.4E+05
1.2E+04
1.2E+04
T.2E+04
1.1E+04

3.2.3   Levels of Internal Surface Contamination Derived for Reference BWR

Internal surface contamination levels in BWR systems and piping reflect the radionuclide
concentrations in the reactor water, steam, and condensate.  Summary inventory estimates of
activity in corrosion film deposited on internal surfaces of equipment and piping are cited in
NUREQ/CR-0672 for a Reference BWR.

The radionuclide composition of corrosion films is shown in Table A3-11.  About 86% of the
estimated inventory at shutdown was, due to two nuclides, Co-60 and Mn-54 (Co-60
contributed nearly half or 47% to the total inventory). It should be noted that internal surface
deposited nuclides generally do not include large amounts of fission products.  Although
fission products do exist in  the primary coolant, they are generally soluble and remain in
solution rather than plate out along with neutron activated corrosion products.  The buildup  of
coolant contaminants is controlled by the letdown system, which continuously removes both
insoluble (particulate) and soluble contaminants.
                                        A3-13

-------
          Table A3-11. Relative Radionuclide Composition of Activated Corrosion
                         Products of Reference BWR at Shutdown

Radionuclide
Cr-51

Mn-54
Fe-59
Co-58
Co-60
Zn-65
Zr-95
Nb-95
Ru-103
Ru-106
Cs-134
Cs-137
Ce-141
Ce-144
Totals
/ Fractional Radioactivity at Decay Times of:
Shutdown
2.1 x 10'2

3.9 x 10'1
2.5 x ID'2
9.3 x 10'3
4.7 x 10"1
6.1 x ID"3
4.0 x 10"3
4.0 x 10'3
2.3 x lO'3
2.8 x 10'3
1.9 x 10'2
3.4 x lO'2
3.0 x ID'3
8.1 x 10'3
1.0
10 Years
—-

1.3 x 10-4
—
—
1.3 x 10'1
1.5 x 1C'7
—
—
—
2.7 x 10^
—
2.7 x lO'2
—
8.8 x lO'7
1.6 x lO'1
30 Yeats
—
*
—
—
—
9.1 x 10'3
—
—
—
—
—
—
1.7 x lO"2
—
—
2.6 x 10"2
50 Years
__—

—
—
—
6.6 x 10-4
—
—
. —
—
—
—
1.1 x lO'2
—
—
1.2 x 1C'2
The total radionuclide inventory has been estimated at 8,500 curies with 6,300 curies
associated with internal equipment surfaces and the remaining 2,200 curies associated with
internal piping surfaces (Table A3-12).


               Table A3-12. Distribution of Activated Corrosion Products on
                          Internal  Surfaces in Reference BWR
Location
Piping
Equipment:
Reactor Building
Turbine Building
Radwaste & Control
Total
Surface Area
(m2)
3.2 x 104

8.6 x 103
2.0 x 105
1.4x 103
2.4 x 10s
Activity Level
(Ci/m2)
6.8 x 10-2

2.2 x 10'1
6.0 x 10'3
2.3 x 10°
26x10°
Total Deposited
Activity (Ci)
2.2 x 103

1.9 x 103
1.2 x 103
3.2 x 103
8.5 x 103
                                        A3-14

-------
For the residual equipment inventory of 6,300 curies, an estimated 30% was associated with
equipment in the reactor building; about 19% was associated with the condenser and feed-
water heaters located in the turbine building; and about 51% involved internal deposition of
equipment in the radwaste and control building.
'                                  '
Of the ^2,200 curies present  in piping, approximately 56% was estimated to be associated with
the reactor water piping and 44% with condensate piping.  Presented below is a more
                'I                                 i     . ^                 T
thorough analysis of piping  data.

Contaminated Piping.  Internal surface contamination levels of BWR piping can be most
useful when grouped according to direct or indirect contact with reactor water, steam/air, and
condensate.  Deposition levels  for reactor water and condensate were based on empirical dose
rate measurements that were correlated to contamination levels for a specific pipe size and
schedule. A summary of measured dose rate data and derived deposition levels is shown in
Table A3-13.
    Table A3-13. Contact Dose Rate and Internal Radioactivity Deposition of BWR Piping
'
Piping Contact ;
Media
Reactor Water
Steam/Air
Condensate

Nominal Outside
Diameter (mm)
610
914
610
„
Wall Thickness
(mm)
59.5
2Q.4
26.0
Measured
Contact Dose
Rate (inR/hr)
700
70
50
Estimated
Deposition
Level (Cl/m2)
1.1
0.005
0.05
Table A3-14 provides a detailed accounting of radionuclide inventories derived for various
size piping made of aluminum, carbon steel, and stainless steel in contact with reactor water,
steam/air, or condensate.
                                          A3-15

-------
Table A3-14. Estimates of Internal Contamination for Reference BWR Piping
Pipe Material
(contact
medium)

Aluminum
(Rx water)
(Steam/Air)
(Condensate)
Carbon Steel
(Rx water)
(Steam/Air)
(Condensate)
Stainless Steel
(Rx water)
(Steam/Air)
(Condensate)
Sub-totals
60mmOD


L
(m)

—
4,300
—

380
1,200
7,400

8
280
7,000
NA

A
(my

—
81
—

71
220
1,400

1.5
53
1,300
3,126

Act,
(Ci)

—
0.4
—

78
1.1
7.0

1.6
0.3
66
154
152 mm OD


L
M

—
1,400
14

1,500
1,800
8,300

34
—
1,600
NA

A
(m2)

—
640
6.7

700
880
3,900

16
—
780
6,923

Act.
(Ci)

—
3.2
0.3

770
4.4
200

18
—
39
1,035
356 mm OD


L
(m)

—
130
—

61
5,600
5,100

61
—
220
NA

A
(m*)



140
—

68
6,300
5,700

68


240
12,516

Act.
(Ci)



0.7
—

75
32
280

75


12
475
533 mm OD


L
(m)



—
—

55
1,200
2,800

55


—
NA

A

-------
Contaminated Equipment.  Contamination on internal surfaces of BWR equipment in contact
with reactor water was based on measurements taken for the heat exchanger in the reactor
water cleanup system. In general, equipment in contact with steam or condensate was
assumed to reach the same levels as previously cited for BWR piping. Exceptions were the
lower values assigned to steam surfaces for the turbine and feedwater heaters.  Table A3-15
provides estimates of radioactivity deposition levels assigned to BWR equipment.

Table A3-16 identifies the major system components and radionuclides inventories based on
location and contact with reactor water, steam, condensate, and radwaste.

        Table A3-15. Summary of Internal Radioactivity Levels in BWR Equipment
Equipment Category
Reactor Water Equipment
Steam Equipment
Turbine
Condensate Equipment
Main Condenser
Feedwater Heaters
Concentrated Waste Tanks/Equipment
Radioactivity Deposition Level1" (Ci/m2)
3.6 x 10'1
5.0 x 1CT3
5.0 x ID"4
5.0 x lO'2
5.0 x 10'3
5.0 x 10'3
5.0 x i(r°
       1  Note: 1 Ci/m2 = 2.2 x 1010 dpm/100 cm2
              Table A3-16.  Estimated Internal Radioactivity in BWR Systems
Bijuldihg/System • A • ,,
Reactor Building
Fuel Pool Exchangers
Skimmer Surge Tanks
Fuel Pool, Rx Wall, Dryer & Sep. Pool
RBCC Water Heat Exchangers
RMCU Regenerative Heat Exchangers
RWCU Nonregenerative Heat Exchangers
RHR Heat Exchangers
Reactor Vessel
Total
Total Estimated
internal Surface ,
I Area (rat2)
8.0 x 102
l.Ox 102
1.4 x 103
1.8 x 103
2.5 x 102
1.7x 102
1.5 x 103
2.6 x 103
8.6 x 103
Radioactivity
Deposition Level1
(Cite?) ;
5.0 x 10'2
5.0 x lO'2
5.0 x lO'2
5.0 x lO'2
3.6 x 10'1
3.6 x 10'1
3.6 x lO'1
3.6 x lO'1
_,
Deposited
Radioactivity
(ci) '
4.0 x 101
5.0 x 101
7.0 x 101
9.0 x 101
9.0 x 101
6.0 x 101
5.4 x 102
9.4 x 102
1.9 xlO3
                                        A3-17

-------
     Table A3-16. Estimated Internal Radioactivity in BWR Systems (Continued)
Building/System
Turbine Generator Building
Main Condenser
Steam Jet Air Ejector Condenser
Gland Seal Steam Condenser
Condensate Storage Tanks
Low-Pressure Feedwater Heaters
Evaporator Drain Tanks
Reheater Drain Tanks
Moisture Separator Drain Tank
Main Turbine
Steam Evaporator
Turbine Bypass Valve Assembly
Moisture Separator Reheaters
Seal Water Liquid Tank
Pumped Drain Tank
High-Pressure Feedwater Heaters
Total
Radwaste and Control Building
Condensate Phase Separator Tanks
Condensate Backwash Receiver Tank
Waste Collector Tank
Waste Surge Tank
Waste Sample Tanks
Floor Drain Collector Tank
Waste Sludge Phase Separator Tank
Floor Drain Sample Tank
Chemical Waste Tanks
Distillate Tanks
Detergent Drain Tank
Decontamination Solution Cone. Waste Tk.
Spent Resin Tank
Cleanup Phase Separator Tanks
Decontamination Solution Concentrator
Total
Total Estimated
I Internal Satfece
Area (m2)
7.9 x 104
1.6x 103
3!5 x 102
1.6x 103
7.5 x 104
1.0 x 101
8.4 x 102
3.0 x 101 '
2.6 x 103
2.0 x 103
1.5 x 101
1.8 x 10"
1.2 x 101
2.7 x 101
1.7 x 104
2.0 x 10s
1.8 x 102
8.5 x 101
1.0 x 102
1.9 x 102
1.6 x 102
1.1 x 102
6.1 x 101
7.8 x 101
1.5 x 102
1.5 x 102
3.2 x 101
2.3 x 101
1.3 x 101
6.8 x 101
1.9 x 101
1.4 x 103
Sadioactivily
Deposition Level1
CCiAn2}
5.0 x 10'3
5.0 x lO'2
5.0 x 10-2
5.0 x 10-2
5.0 x lO'3
5.0 x lO'2
5.0 x 10-2
5.0 x 10'3
5.0 x 10-4
5.0 x 10-3
5.0 x 10"3
5.0 x 10-3
5.0 x lO'2
5.0 x lO'2
5.0 x 10'3

5.0 x 10°
5.0 x 10°
5.0 x lO'2
5.0 x 10°
5.0 x lO'2
5.0 x ID'2
5.0 x 10°
5.0 x 1C'2
5.0 x lO'2
5.0 x 10'2
5.0 x 10'2
5.0 x 10°
5.0 x 10°
5.0 x 10°
5.0 x 10°

Deposited
Radioactivity
(Ci)
3.9 x 102
8.0 x 101
1.7 x 101
8.0 x 101
3.7 x 102
5.0 x lO'1
4.2 x 101
1.5 x 10'1
1.3 x 10°
- 1.0 x 101
7.5 x 10-1
9.0 x 101
6.0 x lO'1
1.4 x 10°
8.5 x 101
1.2 x 103
9.0 x 102
4.2 x 102
5.0 x 10°
9.5 x 102
8.0 x 10°
5.5 x 10°
3.0 x 102
3.9 x 10°
7.5 x 10°
7.5 x 10°
1.6 x 101
1.2 x 102
6.5 x 101
3.4 x 102
9.5 x 101
3.2 x 103
Source: NUREG/CR-0672 Vol. 2, Appendix E




1  Note:  1 Ci/m2 = 2.2 x 1010 dpm/100 cm2
                                  A3-18

-------
3.2.4  Levels of Internal Surface Contamination for Reference PWR

Radioactive contamination levels associated with internal surfaces of piping and equipment
for a Reference PWR have been estimated in NUREG/CR-0130.  At time of shutdown, the
fractional contributions of various radionuclides deposited on internal surfaces of the primary
loop of a PWR are shown in Table A3-17.

Estimates of internal surface deposition levels expressed in Ci/m2 for major systems and
components were based on models, which correlated external dose rate measurements with
internal contamination analyses, taking into account source geometry and shielding factors
(Table A3-18).  Empirical dose rate measurements showed that reactor vessel and steam
generator internal surfaces in contact with primary coolant, on average, would yield
contamination levels of about 0.23 Ci/m2 at time of shutdown.
                 Table A3-17.  Estimates of Internal Surface Contaminants
                          in a Reference PWR Primary System
Radionuclide
Cr-51
Mn-54
Fe-59
Co-58
Co-60 ,
Zr-95
Nb-95
Ru-103
Cs-137
Ce-141
TOTAL
Deposited
Radioactivity
(nCi/m2)
5.3 x 103
8.0 x 103
1.8 x 103
1.0 x 105
7.1 x 10*
8.8 x 103
1.2 x 104
5.9 x 103
2.6 x 102
1.5 x 10*
2.3 x 10s
Fractional Radioactivity at Decay Times 
-------
      Table A3-18. Levels of Contamination and Estimated Quantities of Radioactive
           Corrosion Products Deposited on the Interior of PWR Reactor Systems
Systems
Reactor Vessel and Internals
Steam Generators
Pressurizer
Piping (Except RCS)
RCS Piping
TOTALS
Surface
- , Xxn2)
5.7 x 102
1.9 x 104
8.7 x 101
'l.l x 103
1.9 x 102
219 x 102
Activity Level

-------
AUX SPRAY
FROM CVCS
                    SHIELD
                 PRESSURIZES
                                            PRESSURIZER f
                                             RELIEF TANK S
                                HEATER
                              CONTROLLER
          REACTOR         •—0.699raID PIPE
       COOLANT PUMP Q*-~T
        LOOP 2
          .
  0.787 m ID PIPE
  S1EAM
GENERATOR
                           0.736m ID PIPE
                         T SAFETY INJECTION
             STEAM-^            SAFETY INJECTION
            GENERATOR     5HELL      	
                                                                 REACTOR
                                                               COOIANT PUMP
              TUBE
        LOOP1

                         REAQOR
                       COOLANT PUMP
                                      FROM CVCS
                                   NORMAL CHARGING
            Figure A3-1.  PWR Primary System Schematic and Piping Data
                                        A3-21

-------
For non-RCS or auxiliary system piping, an average internal deposition of about 0.06 Ci/m2
was derived based on external dose rate measurements.  The collective inventory of 60 Ci for
all non-RCS piping was estimated based on the piping quantities defined in Table A3-19.

                   Table A3-19. Non-RCS Contaminated Piping Data
Diameter
Size (mm)
12
19
25
25-38
51
71
102
152
203
254
305
356
Total
Schedule
80
160
40
80
160
40
80
160
40
80
160
40
80
160
160
160
160
160
140
140
140

Length
(m}
120
120
240
360
570
60
180
420
120
330
540
300
480
1,050
140
180
300
140
365 '
90
100

Total Wt
(fcg)
198
238
205
400
1,675
152
590
1,800
493
1,811
3,967
1,655
3,642
11,850
2,985
6,128
20,972
15,924
29,750
18,370
25,475
-148,000
Total Inside
Surface, (m2)
5.3
4.1
8.0
10.8
28.3
5.1
14.0
22.7
15.7
40.1
58.6
50.3
75.4
143.6
29.4
50.2
128.7
70.7
134.1
74.0
92.3
-1,110
Total
Activity (Ci)
0.32
0.25
0.48
0.65
. 1.70
0.31
0.84
1.36
0.94
2.41
3.52
3.02
4.52
8.62
1.76
3.01
7.72
4.24
8.05
4.44
5.54
-60
                                       A3-22

-------
3.3    Contamination of External Surfaces of Equipment and Structural Components

External surfaces of system components as well as floors, walls, and structural components
become contaminated over the operating lifetime of a reactor plant from leaks and spills of
radionuclides originating from reactor coolant water. While most liquid contamination
remains highly localized to the vicinity of the leak/spill, some contamination may experience
limited transfer through physical  contact.  More widespread contamination of external
surfaces occurs when contaminants  become airborne and passively settle out. Airborne
contaminants are also the principal  source of contamination of ducts, fans, filters, and other
equipment that are part of the heating and ventilation and air conditioning systems (HVAC).

Mixtures of radionuclides typically  found in primary coolant and their relative abundance in a
PWR and BWR are given in Table  A3-20 and Table A3-21, respectively.

        Table A3-20.  Radionuclides in Primary Coolant that Contribute to External
                      Surface Contamination in the Reference PWR
Radionucltde
Cr-51
Mn-54
Fe-55
Fe-59
Co-58
Co-60
Sr-89
Sr-90
Y-90
Zr-95
Nb-95
Te-129m
1-131
Cs-134
Cs-136
Cs-137
TOTAL
Half-Life
'(days)
2.8E+1
3.1E+2
9.5E+2
4.5E+1
7.2E+1 -
1.9E+3
5.3E+1
l.OE+4
l.OE+4
6.5E+1 '
3.5E+1
3.4E+1
8.0E+0
7.5E+2 -
1.4E+1
1.1E+4

Fractional Radioactivity at Decay Times of:
Shutdown
6.9E-4
1.4E-3
2.2E-2
8.7E-4
7.5E-3
7.5E-2
1.2E-3
6.9E-4
6.9E-4
2.5E-4
2.5E-4
3.1E-4
1.4E-2
1.2E-1
1.1E-3
7.5E-1"
1.0
' , 10 Years
—
4.2E-7
1.7E-3

---
2.0E-2
—
5.4E-4
5.4E-4
—
—
—
—
' 4.1E-3
?
5.9E-1
6.2E-1
30 Years
—
—
9.9E-6
—
— ^
1.4E-3
—
3.4E-4
3.4E-4
—
—
i
—
4.8E-6
—
3.7E-1
3.7E-1
50 Years
—
—
5.7E-8
—
—
l.OE-4
' —
2.1E-4
2.1E-4
—
—
—
—
5.4E-9
— ,
2.4E-1
2.4E-1
                                        A3-23

-------
Table A3-21. Radionuclide Concentrations in Primary Coolant that Contribute
            to External Surface Contamination in Reference BWR
Radionuclide
P-32
Cr-51
Mn-54
Fe-55
Fe-59
Co-59
Co-60
Ni-63
Zn-65
Sr-89
Sr-90
Y-90
Y-91
Zr-95
Ru-103
Ru-106
Ag-llOm
Te-129m
1-131
Cs-134
Cs-136
Cs-137
Ba-140
Ca-140
Ce-141
Ce-144
Pr-143
Nd-147
TOTAL
Half-Life
{days}
1.4E+1
2.8E+1
3.1E+2
9.5E+2
4.5E+1
7.2E+1
1.9E+3
3.6E+4
2.4E+2
5.3E+1
l.OE+4
2.7E+0
5.9E+1
6.5E+1
4.0E+1
3.7E+2
2.5E+2
3.4E+1
8.0E+0
7.5E+2
1.4E+1
1.1E+4
1.3E+1
1.7E+0
3.2E+1
2.8E+2
1.4E+1
1.1E+1

CoiK^tttiatiOH
(liCi/g)
2E-4
5E-3
6E-5
1E-3
3E-5
2E-4
4E-4
1E-6
2E-4
1E-4
6E-6
6E-6
4E-5
7E-6
2E-5
3E-6
1E-6
4E-5
5E-3
3E-5
2E-5
7E-5
4E-4
4E-4
3E-5
3E-6
4E-5
3E-6
1.3E-2
Fractional Radioactivity at Decay Times of:
Shutdown,
1.1E-3
5.3E-2
7.2E-4
3.7E-1
5.3E-4
5.6E-3
2.9E-1
3.4E-3
1.8E-2
2.0E-3
1.5E-2
1.5E-2
8.1E-4
1.6E-4
2.9E-4
3.9E-4
8.8E-6
4.9E-4
1.5E-2
8.8E-3
l.OE-4
1.8E-1
2.0E-3
2.0E-3
3.4E-4
2.9E-4
2.0E-4
1.2E-5
1.0
10 Years
—
—
2.3E-7
2.5E-2
—
—
7.8E-2
3.2E-3
4.5E-7
—
1.2E-2
1.2E-2
—
—
—
—
3.2E-10
—
—
3.1E-4
—
1.4E-1
—
—
—
3.1E-8
—
—
2.7E-1
30 Years :
—
—
—
1.2E-4
—
—
5.6E-3
2.8E-3
—
—
7.0E-3
7.0E-3
—
—
—
—
—
—
—
3.7E-7
—
9.0E-2
—
—
—
—
—
—
1.1E-1
50 Yeats
—
—
—
6.0E-7
—
—
4.0E-4
24E-3
—
—
4.2E-3
4.2E-3
—
—
—
—
—
—
—
4.4E-10
—
5.7E-2
—
—
—
—
—
—
6.8E-2
                                 A3-24

-------
The amount of external surface contamination following 40 years of operation is likely to
vary significantly among plants and is influenced by fuel integrity, primary coolant chemistry,
operational factors, and reactor performance.  In addition, a key operational factor is the effort
at operating plants to clean up spills and to decontaminate accessible areas on a ongoing
basis.
                                                       i
Although all nuclear utilities conduct routine radiological surveys that assess fixed and
removable surface contamination, only limited data have  been published in the open literature
from which average contamination estimates can be derived.  In this section, estimates of
external surface contamination are provided that reflect (1) modeled data, (2) data published
in the open literature, and (3) data submitted by individual utilities that have submitted a
Decommission Plan.

3.3.1   Modeled Data for Reference'Facility

Quantities and locations of structural surface contamination have been modeled for Reference
PWR (NUREG/CR-0672).  The model was based on an assumed release rate of one liter of
primary coolant per day for 40 years. Deposition of contaminants on external surfaces was
also correlated to ambient dose rates by means of the computer code ISOSHLD and placed in
two discrete categories.

The first category is defined as low-level contamination areas with dose rates of 10 mR/hr in
air  at 1 meter from the surface.  The second category was defined as areas of higher
contamination  with dose rates of 100 mR/hr in air at 1 meter from the surface.  The structural
surface contamination levels that correspond to dose rates of 10 and 100 mR/hr for the
Reference BWR coolant water mixture were estimated to correspond to 2.5 x 10"3 Ci/m2 and
2.5  x 10"2 Ci/m2, respectively.

Table A3-22 summarizes the distribution and quantities of external surface contamination at
shutdown.  Surface contamination levels are expressed in units of disintegrations per minute
per one-hundred square centimeters of surface (dpm/100 cm2). The total deposited
radioactivity on structural surfaces in the Reference BWR was estimated at 114 Ci.
                                        A3-25

-------
        Table A3-22.  Surface Contamination Levels for Reference BWR at Shutdown
Building
Reactor Building
(Contamination Level l)(a)
(Contamination Level 2)(b)
Turbine Generator Bldg.
(Contamination Level l)w
(Contamination Level 2)(b)
Radwaste & Control Bids;.
(Contamination Level l)(a)
(Contamination Level 2)(b)
TOTAL
Estimate
Surface Area
- . 
-------
Table A3-23. Estimated Structural Surface External Contamination in the Reference BWR*
; • , - ' "
Building/ f • ' - - -,
Associated Equiproent/Systeni/Structure
/ .. ' ' '.
Reactor Building
Containment Atmosphere Control
Condensate (Nuclear Steam)
Control Rod Drive
Equipment Drain (Radioactive)
Floor Drain (Radioactive)
Fuel Pool Cooling & Cleanup
Fuel Pool Cooling & Cleanup
High-Pressure Core Spray
Low-Pressure Core Spray
Main Steam
Miscellaneous Wastes (Radioactive)
Reactor Building Closed Cooling
Reactor Core Isolation Cooling
Reactor Water Cleanup
Reactor Water Cleanup
Residual Heat Removal
Standby Gas Treatment
Traversing Incore Probe
Primary Containment
Total .
Estimated
Contaminated
Surface Area
~" ' ' («#
1.6 x 101
3.3 x 101
1.8 x 102
1.8 x 101
7.4 x 101
1.2 x 103
2.8 x 102
1.1 x 102
1.4 x 101
3.0 x 102
8.3 x 101
1.2 x 101
1.5 x 101
1.5 x 102
1.7 x 102
1 7 x 102
4.0 x 101
8.0 x 101
2.2 x 103

• -.
Radioactivity
Deposition
'Level
1
1
1
2
2
1
2
1
1
1
1
,1
1
1
2
1
1
1
2

>
Deposited
Radioactivity
(Ci>
4.0 x 10'2
8.2 x 10'2
4.5 x 10'1
4.5 x 10'1
1.8 x 10°
3.0 x 10°
- 7.0 x 10°
2.7 x 10'1
3.5 x lO'2
7.5 x 10'1
2.1 x 10'1
3.0 x 10'2
3.8 x 10'2
3.8 x 10'1
4.2 x 10°
'4.2x 10'1
1.0 x 10"1
2.0 x 10'1
5.5 x 101
7.4 x 101
                                      A3-27

-------
          Table A3-23.  Estimated Structural Surface External Contamination
                11    .      in Reference BWR* (Continued)

Building/
Associated Equipment/System/Structure

Turbine Generator Building
Air Removal
Condensate (Nuclear Steam)
Condenser Off Gas Treatment
Equipment Drain (Radioactive)
Floor Drain (Radioactive)
Heater Drain
Main Steam
Miscellaneous Drain & Vent
Reactor Feedwater
Miscellaneous Wastes (Radioactive)
Total
Radwaste and Control Building
Condensate Filter Demineralizer
Condenser Off Gas Treatment
Equipment Drain (Radioactive)
Equipment Drain (Radioactive)
Floor Drain (Radioactive)
Floor Drain (Radioactive)
Floor Pool Cooling & Cleanup
Miscellaneous Wastes (Radioactive)
Miscellaneous Wastes (Radioactive)
Process Waste (Radioactive)
Process Waste (Radioactive)
Reactor Water Cleanup
Total
Estimated
Contaminated
Surface Area
(m2)

3.9 x 101
6.6 x 102
1.8 x 102
2.5 x 101
2.5 x 101
9.1 x 101
1.7 x 102
1.9 x 101
6.9 x 102
9.0 x 10°


3.6 x 102
3.2 x 102
4.3 x 101
1.8 x 102
1.2 x 101
1.9 x 102
5.4 x 101
2.4 x 101
1.9 x 102
1.8 x 102
2.7 x 102
1.3 x 102

-•
Radioactivity
Deposition
Levei

1
1
1
2
2
1
1
1
1
1


2
1
1
2
1
2
2
1
2
1
2
2

_, t
Deposited
Radioactivity
(GO

9.7 x lO'2
1.6 x 10'1
4.5 x 10'1
6.2 x 10'1
-6.2 x 10'1
2.3 x 10'1
4.2 x 10'1
4.7 x 10'2
1.7 x 10°
2.2 x 10'2
4.4 x 10°

9.0 x 10°
8.0 x 10'1
1.1 x 10"1
4.5 x 10°
3.0 x 10'2
4.8 x 10°
1.4 x 10°
6.0 x 10'2
4.8 x 10°
4.5 x 10'1
6.7 x 10°
3.2 x 10°
3.6 x 101
Estimated total deposited radioactivity on contaminated external surfaces = 1.14 x 102 Ci
*   Source: NUREG/CR-0672
                                     A3-28

-------
Modeled Estimates Versus Empirical Study Data.  External surface contamination
                                ^                5>
corresponding to Level 1 (2.5 x 10"3 Ci/m2 or 5.2 x 107 dprii/100 cm2) and Level 2 (2.5 x 10"2
Ci/m2  or 5.5 x  10s dpm/100 cm2) are not uncommon and have been observed in most reactor
facilities.  Table A3-24 presents study data that focused on the most highly contaminated
surfaces at six nuclear power plants (NUREG/CR-4289).  Contamination levels corresponding
to mbdeled values (i.e., Level 1 and Level 2), however, were restricted to small areas that had
experienced spills, leaks, or  intense maintenance, such as the reactor sump area, RCS coolant
pumps, and radwaste system components. The  study data also showed that when surfaces
were coated with sealant or  epoxy  paint,  nearly all contamination resided on or within the
surficial coating and was readily removable.

       Table A3-24. Ranges of Radionuclide Associated with Highly Contaminated
                   External Surfaces at Six Nuclear Generating  Stations
Radionuclide
Cq-60
Ni-59
Ni-63
Sr-90
Tc-99
Cs-137
Eu-152
Eu-154
Eu-155
Pu-238
Pu-239, 240
Am-241
Cm-244
Half-life ,
, <3& ...
5.27
75,000
100
28.5
2.13E+5
30.2
12.4
8.5
4.96
87.8
24,400
433
18.1
- Concentration Range
(pCi/cm2)
590 - 460,000
30-2,400
3,100 - 6,400
1.6 -.480
0.27 - 2.4
550 - 2.0E+6
9 - 3,100
90 - 1,500
10 - 500
0.025 - 48
0.089 - 21
0.10 - 30
0.013 - 0.026
Average Concentration
(djpm/lOO cm2) :
1 2.4 x 107 (5)*
1.9 x 10s (3)*
1.0 x 106 (2)*
3 7 x 104 (4)*
3.5 x 102 (3)*
8.1 x 107 (6)*
2.2 x 10s (3)*
1.5 x 10s (3)*
' 1.3 x 104 (2)*
3.1 x 103 (4)*
1.7 x 103 (4)*
1.9 x 103 (4)*
3.5 x 10° (3)*
       * Number of reactor units included to calculate the average value.

In summary, the modeled external  surface contamination levels cited in NUREG/CR-0672 for
Reference BWR appear excessive in terms of their projected surface areas and total plant
inventory. The primary model parameter regarding the release of one liter of primary coolant
                                        A3-29

-------
per day that is allowed to buildup over a forty-year period of plant operation is not only
without technical basis but ignores the ongoing decontamination efforts that exist at all
nuclear facilities. For these reasons, modeled data contained in NUREG/CR-0672 are not
considered suitable.
3.3.2  Surface Contamination Levels Reported by Facilities Preparing for Decontamination
       and Decommissioning (D&D)
                                         31
PWR. By coincidence (as was previously acknowledged), the Trojan Nuclear Plant (TNP),
which had served as the Reference PWR facility in the 1978 study (NUREG/CR-0130), has
been permanently shutdown and has submitted a Decommissioning Plan.  External  surface
contamination inventories at this facility are summarized in TNP's Decommissioning Plan and
have been reproduced in Table A3-25.  Estimates were based on historical survey data and
recent structural surveys performed in support of the Decommissioning Plan's required
Radiological Site Characterization.

                Table A3-25. Inventory of External Surface Contamination
                                 at Trojan Nuclear Plant
Location
Structures
Containment Building
Auxiliary Building
Fuel Building
Main Steam Support Structure
Turbine Building
Total
Activity 
-------
removable floor contamination levels obtained by smears. However, such measurements may
reasonably be assumed to also represent metal surfaces of reactor systems and structural metal
components.

A summary of removable external surface contamination levels at TNP are given in Table
A3-26.

    Table A3-26. Area! Surface Contamination Levels Based on Survey Measurements
                              at TNP Preparing for D&D*
Building '
Reactor
Containment
Auxiliary
(6 levels)
Fuel Building
(5 levels)
Turbine
Building
Control
Building
Approximate ,
Floor Surface
Area, m2
1,900
4,000
- 5,000
5,700 per
level
700 per
level
Estimated %
orFloor Area
Contaminated
100
1-5
1 -5
« 1
« 1
, Estimated %'
Needing
Cleanup, m2
1,900
40 - 200
50 - 250
~0
~0
'Removable Measured
Contamination Level,
dpm/100 cnr*
1,100-55,000 -
< 1,100 - 7,900
< 1,100 -5,000
< 1,000
< 1,000
       *  Source:  NUREG-1496

The auxiliary and fuel buildings also exhibited some areas of floor contamination, but not to
the extent of that observed in the reactor containment building.  Based on survey reports,
about 1 to 5% of the floor area (representing about 40 to 200 m2) in the auxiliary building
has radioactive contamination levels in the range of 1,100 to 7,900 dpm/100 cm2. The fuel
handling building also has a small amount of floor contamination, estimated at approximately
50 to 250 m2, with contamination levels ranging from about 1,100 to 5,000 dpm/100 cm2.

Other buildings, including the turbine building and the control building, did not have
measurable contamination on any surfaces.

It is important to note, however, that the quantitative estimates in Table A3-25 reflect
contamination that is removable (i e., by wiping a 100 cm2 area with a dry filter paper).
                                       A3-31

-------
Reasonable estimates of total surficial contamination levels (i.e., fixed and removable) may be
obtained by multiplying values in Column 5 of Table A3-26 by a factor whose value may
range from 5 to 10.

BWR. Values similar to those reported in the TNPs Decommissioning Plan have also been
reported in the decommissioning plan submitted for Humboldt Bay Unit 3.  Excerpts of
survey measurements (as they appear in the D-Plan) are contained in Addendum #2,
Horizontal surfaces (Le., floors) exhibited contamination levels -that on average were about
one order of magnitude higher than vertical surfaces (i.e., walls) with values ranging from
below detection limits up to several million dpm per 100 cm2 for select floor areas (e.g.,
under the reactor vessel). When relatively small areas of high contamination are excluded,
average external surface contamination was generally between 5,000 dpm/100 cm2 to 100,000
dpm/100 cm2.

From the above-cited data, it is concluded that, within the common variabilities of
contamination levels in nuclear plants, the survey data reported in decommissioning plans for
the Trojan and Humboldt Bay facilities provide a reasonable estimate of surface
contamination levels for other PWRs and BWRs.
                                        A3-32

-------
 4.0    BASELINE METAL INVENTORIES

 4.1    Baseline Metal Inventories for Reference PWR

 The total amounts of metals contained in significant quantities in a typical 1,000 MWe
 pressurized water reactor (PWR) power plant have been quantified in a 1974 study of material
.resource use and recovery in nuclear power plants (Bryan and Dudley 1974).  Material
 estimates were made using various methods that included. (1) amounts of raw materials
 purchased for construction (e.g., reinforcing steel and  structural steel required for
 construction), (2) weights of materials contained in equipment and machinery based on
 manufacturers' specifications and technical journals (e.g., determination of carbon steel,
 stainless steel, copper and other metals in electric motors); and (3) the U.S. Atomic Energy
 Commission facility accounting system, which identified individual items.

 Summary estimates of composite materials used to construct a 1971-vintage 1,000 MWe PWR
 power plant are given in Table A4-1.

      Table  A4-1. Inventory Estimates of Materials Used to Construct a 1971-Vintage,
                      1,000 MWe, Pressurized Water Reactor Facility
Material
Metals:
Carbon Steel
(Rebar)
(All Other) •
Stainless Steel
Galvanized Iron
Copper
Inconel
Lead
Bronze
Aluminum
Brass
Nickel
Silver
Total Quantity
(Metric tons)

3.3 x 104
(1.3 x 104)
(2.0 x 104)
2.1 x 103
1.3 x 103
6.9 x 102
1.2 x 102
4.6 x 101
2.5 x 101
1.8 x 101
1.0 x 101
1.0
< 1.0
                  Source:  Bryan and Dudley 1974
                                         A4-1

-------
Carbon stee! is by far the most abundant metal used in the construction of a nuclear power
plant.  It is used in piping and system components when the need for corrosion resistant
stainless steel is not of significant importance. A large percentage is also used in structural
components that include rebar, I-beams, plates, grates, and staircases. A breakdown of
material quantities used in reactor plant structures and plant systems is provided  in Table
A4-2.  Essentially 50% or 16,519 metric tons out of total of 32,731 metric tons of carbon
steel is used for structural components with the other one-half used in plant equipment.  Of
the more than 16,000 metric tons of carbon steel employed in plant equipment/systems, about
two-thirds (i.e., 10,958 metric tons) are contained in turbine plant equipment  Barring
significant leakage in steam generators, equipment in this grouping as well as electric plant
equipment, equipment identified as  "miscellaneous," and "structures" are not likely to be
exposed to radionuclides/radioactivity and are, therefore, not likely to contribute .significant
quantities of scrap metal.

The primary sources of contaminated scrap metal in a PWR are identified as shaded areas in
Table A4-2 and involve all items associated with Reactor Plant Equipment with additional
quantities contributed by "Fuel Storage," select structural components, HVAC systems, and
other items that are identified in detail in Section 5 0 below.

Inspection of Table A4-2 also reveals that the use of corrosion resistant stainless steel is
almost totally confined to reactor plant and turbine plant systems. Of the total 2,080 metric
tons of stainless steel, essentially all of the 1,154.6 metric tons associated with reactor plant
systems and lie 21.1 metric tons that line the fuel pool can be assumed contaminated.

4.2    Baseline Inventories for Reference BWR

Inventories for a 1,000-MWe BWR reference plant have been estimated by adjusting Bryan
and Dudley's 1974 Reference PWR plant data taking into account BWR characteristics
(NUREG/CR-0672).
                                         A4-2

-------
           Table A4-2.  Breakdown of Material Quantities Used in Plant Structures and Reactor Systems
                                                  Metric Tons1-2
System
Structures/Site
Site Improvements
Reactor Building
Turbine Building
Intake/Discharge
Reactor Auxiliaries -
Fuel Storage
Miscellaneous Bld#s.
Reactor Plant Equipment
Reactor Equipment
Main Heat Trans, System
Safeguards Cool, System
Radwaste System
Fuel Handling System
Other Reactor Equipment
Instrumentation & Control '
Turbine Plant Equipment
Turbine-Generator
Heat Rejection Systems
Condensing Systems
Feed-Heating System
Other Equipment
Instrumentation & Control
Electric Plant Equipment
Switchgear
Station Service Equip.
Switchboards
Protective Equipment
Structures & Enclosure
Power & Control Wiring
Miscellaneous Equipment
Transportation & Lifting
Air & Water Service Sys.
Communications Equip.
Furnishings & Fixtures
ENTIRE PLANT
Carbon
Steel
16519.3
1692.9
7264.2
3641.2
333.7
WM
364.6
1864
334,9
43GJ)
168&5
274.2^
' 33.2.
. .:8&G
823.5
113.5
10,958.3
4138.6
2501.1
1359.8
1367.7
1541.3
49.8
965.5
30.4
654.1
87.0
5.9
112.5
75.6
843.2
529.3
232.5
4.7
76.7
32,731.2
Stainless'
Steel '
28.6
0.0
5.7
0.0
0.0
. 0,0
21.1
1.8
1154.6
: '275.1
'202,5
199,1
, 31,9
. 67,0
' 230,3
148,7
883.2
129.9
9.1
392.3
221.2
89.4
41.3
0.0
0.0
0.0
0.0
0.0
0.0
0.0
13.7
0.0
6.0
0.0
7.7
2080.1
Qalvanized
Iron
814.2
17.9
301.2
196.4
3.6
109,$
43.4'
141.9
'5.5
0.0
- : u
1,1
00>0
0.0
124.2
..1244:
o.o,
0,0
- 0.0
0,0
0,0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
124.2
Lead
-33.1
0.7
0.0
0.0
0.0
0<0
, 0,0
32.4
4.5
0,0'
0,0
0,0
0;0
' 0,0
4,5
0,0
0.0 •
0.0
0.0
0.0
0.0
0.0
0.0
6.8
0.0
6.8
0.0
0.0
0.0
0.0
2.0
0.0
0.0
0.0
2.0
46.4
Bronze _
0.2
0.0
0.0
0.1
, 0.0
% 0,0
0,0
0.1
0,5
0,0
* 0.0
0,1
0<0
0,0
0.4
,0.0
21.5
19.7
0.7
0.3
0.3
0.5
0.0
2.5
0.7
0.7
0.1
0.5
0.0
0.5
0.4
0.0
0.0
, 0.0
0.4
25.1
Aluminum
1.2
0.1
0.1
0.8
0.0
0.0'
: OJ
0.1
5,2 '
0,0
o.o -
: * 0*0- .
0.0
0.0
0,0
. 5,2
1.2
0.0
0.0
0.0
0.0
0.0
1.2
4.1
0.0
0.0
4.1
0.0
0.0
0.0
6.5
0.0
0.0
0.4
6.1
18.2
Brass r
2.9
0.0
0.3
1.4
0.0
0,2
, 04
0.9
0.0.
0,0
0.0
0.0
0:0
' 0.0
0.0
' 0.0
6,9
0.0
0,4
1.5
3.9
•1.1
0.0
0,0
0.0
0.0
00
0.0
0.0
0.0
0.3
0.0
0.3
0.0
0.0
10,1
NiQkel
0.1
0.0
0.0
0.0
0.0
0,0
0.0-
0.1
0.0
0,0
0,0
0,0'
"0,0
0.0
0,0
0,0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.6
0.0
0.0
0.0
0.0
0.0
0.6
0.0
0.0
0.0
0.0
0.0
0.7
Silwr
0.1
0.0
0.0
0.0
0.0
0.0
' . 0,0
0.1
- 0,0
'uo
0.0
0.0
.0.0V
0.0
0.0
',
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.4
0.3
0.1
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.0
0.5
Source: Bryan and Dudley 1974
Shaded areas identify PWR equipment/systems with significant amounts of radioactive contamination.
                                                      A4-3

-------
With regard to plant inventories of steel, notable differences between a PWR and BWR are
that the latter has less heat-transfer piping and the absence of steam generators, but more ex-
vessel primary components inclusive of a pressure suppression chamber.  An additional
significant difference between the PWR and BWR is the estimated, quantity of rebar used for
concrete reinforcement.  Of the 32,700 tons total in the Reference 1,000 MWe PWR, Bryan
and Dudley estimated that about 13,300 tons is rebar; for the 1,000 MWe Reference BWR,
the collective weight  of rebar was  estimated at 18,300 tons (NUREG/CR-0672).

Beyond an increase in total steel required to construct a BWR, a second major difference is  .
the enhanced percentage of steel (and other metals)  that is contaminated.  This is due to  the
fact that, under normal operating conditions of a BWR, radionuclides enter the steam flow
and contaminate turbine plant equipment that in a PWR may generally be assumed
uncontaminated.

Table A4-3 identifies material estimates for a 1,000-Mwe BWR plant.  Material estimates for
metals other than carbon and stainless steel for the  1,000-MWe Reference BWR are assumed
to be identical to those of the 1,000-MWe Reference PWR.

             Table A4-3. Inventory Estimates of Materials Used to Construct
                      a 1,000-Mwe Boiling Water Reactor Facility
                  Material
                  Metals:
                    Carbon Steel
                      (Rebar)
                      (All Other)
                    Stainless Steel
                    Galvanized Iron
                    Copper
                    Mconel
                    Lead
                    Bronze
                    Aluminum
                    Brass
                    Nickel
                    Silver
Total 'Quatrtitf
 (Metric tons)
   3.4 x 104
  (1.8 x 104)
  (1.6 x 104)
   2.1 x 103
   1.3 x 103
   6.9 x 102
   1.2 x 102
   4.6 x 101
   2.5 x 101
   1.8 x 101
   1.0 x 101
      1.0
     < 1.0
                    Source: NUREG/CR-0672
                                         A4-4

-------
4.3    The Applicability of Reference Facility Data to the Nuclear Industry

The applicability of material estimates cited by Bryan and Dudley (1974) to all currently
licensed U.S. facilities is not without some difficulty.  The current U.S. nuclear power plant
inventory of 123 units is composed of not only different designs but also highly variable
power ratings.

Reactor power plant designs reflect evolving standards over four decades for plant safety and
environmental concerns. For example, Bryan and Dudley's reference plant used run-of-river
cooling, which is not applicable to more recent nuclear facilities that employ cooling towers
of various designs, holding ponds, sprays, etc.  Significant quantities of materials are involved
in some of these alternative cooling systems.  Additionally, the 1979 accident at the Three
Mile Island facility mandated revised safety standards, which have added to the material
inventory of more recent nuclear plants.

Material inventories that reflect evolving  changes in plant design, however, have not been
adequately  addressed in the open literature.  Adjustments to material inventories for individual
facilities will, therefore, be limited to the reactor's power rating by means of a scaling  factor.

Scaling Factors.  In general,  it is reasonable to assume a positive correlation  between a plant's
power rating and its material inventory.  In a recent draft report prepared by  Argonne
National Laboratory for the U.S. Department of Energy, a scaling method was employed that
was based on PWR and BWR  vessel mass data (Nuclear Engineering International data 1991,
1992, and 1993). In these reports, it is assumed that all metal inventories for both PWRs and
BWRs correlate to the corresponding reference plant in proportion to the design power rating
to the 2/3 power (i.e., MWe2'3). Thus, the scaling factor pf 0.86 and 1.13 would correspond
to plants with an 800 MWe and 1,200 MWe generating capacity, respectively.

This scaling factor was applied multiplicatively to each of the 123 plants (identified in
Addendum #1) for estimating total industry inventories (Table A4-4).  For most metal
groupings, the all-inclusive inventory estimates include a majority component that is not
radiologjcally contaminated.
                                         A4-5

-------
Table A4-4. Summary of Total Metal Inventories Potentially Available for Recycling
                                (Metric Tons)

Carbon Steel
(Rebar)
(All Other)
Stainless Steel
Galvanized Iron
Copper
Inconel
Lead
Bronze
Aluminum
Brass
Nickel
Silver
Reference ,
BWR
3.4 x 104
(1.8 x 104)
(1.6 x 104)
2.1 x 103
1.3 x 103
6.9 x 102
1.2 x 102
4.6 x 101
2.5 x 101
1.8 x 101
1.0 x 101
1.0
< 1.0
PWR
3.3 x 104
(1.3 x 104)
(2.0 x 104)
2.1 x 103
1.3 x 103
6.9 x 102
1.2 x 102
4.6 x 101
2.5 x 101
1.8 x 101
1.0 x 101
1.0
< 1.0
Industry
AllBWRs
1.1 x 106
(6.0 x 10s)
(5.4 x 10s)
7.1 x 104
4.4 x 104
2.3 x 104
4.0 x 103
1.6x 103
8.4 x 102
6.1 x 102
3.4 x 102
3.4 x 101
< 3.4 x 101
AlIPWRs
2.5 x 106
(1.0 x 106)
(1.5 x 106)
1.6 x 10s
1.0 x 10s
5.4 x 104
9.4 x 103
3.6 x 103
2.0 x 103
1.4x 103
7.8 x 102
7.8 x 101
< 7.8 x 101
Total
3.6 x 106
(1.6 x 106)
(2.0 x 106)
2.3 x 10s
1.4 x 105
7.7 x 104
- 1.3 x 104
5.2 x 103
2.8 x 103
2.0 x 103
1.1 x 103
1.1 x 102
< 1.1 x 102
                                    A4-6

-------
5.0    METAL INVENTORIES SUITABLE FOR RECYCLING

From data presented in previous sections, two important conclusions can be stated:  (1) only a
fraction of metal inventories is likely to be significantly contaminated and (2) not all
contaminated metal inventories are potentially suitable for recycling.  "Suitability for
recycling" is largely determined by the practicality and efficacy with which contaminated
scrap can be decontaminated to a level considered acceptable in terms of risks to human
health and the environment.

The choice of available decontamination methods needed to render scrap metal suitable for
recycling (or unrestricted use) is largely dependent on the starting level of contamination
encountered, the type of surface, physical accessibility to the surface, radionuclides involved
and their chemical state(s), and size/configuration of the object requiring decontamination.

Several techniques are currently used in decontamination efforts at nuclear facilities.  Their
applicability, however, is not without restrictions and for nearly all approaches, there are
numerous factors that affect their efficiency.  Examples include the choice of
cleaner/solvent/surfactant for "hand wiping"; the selection of chemical solvent for the
dissolution and removal  of radioactive corrosion films or base metal;  or the innovative use of
dry-ice (CO2) pellets for abrasive blasting. These techniques and their general applicability
and limitations are briefly summarized below.

Hand Wiping.  Rags moistened with water or a solvent such as acetone can be an effective
decontamination process. Wiping can be used extensively and effectively on smaller items
with low-to-medium external contamination levels and easily accessible internal
contamination.  This method may not work well if the item  is rusty or pitted.  It requires
access to all surfaces to be cleaned, is a relatively slow procedure, and its hands-on nature
can lead to high personnel exposure.  On the positive side, wiping can provide a high
decontamination factor (DF), generates easily  handled decontamination wastes (contaminated
rags), requires no special equipment, and can be used selectively on portions of the
component.

Steam Cleaning. This may be-performed either remotely in  a spray booth or directly by
decontamination personnel using some type of hand-held wand arrangement.  In the former
case, only minimal internal decontamination is possible; however, reasonable external
                                         A5-1

-------
cleaning can be accomplished quickly with low exposure expenditures.  Containment of the
generated wastes and protection of personnel from radioactive contamination may be difficult.
            4
Abrasive Blasting.  This is a highly effective procedure even for surfaces that are rusty or
pitted.  As with hand-held steam cleaning, this method suffers from internal accessibility
problems.  It also generates large amounts of solid wastes and, being a dry process, produces
significant quantities of airborne radioactivity.  Abrasive blasting may be used if its high
effectiveness can be justified after taking the exposure, waste, and accessibility limitations
into account.  Some of the aforementioned disadvantages are eliminated when dry ice (CO2)
pellets are used.

Hydrolasing.  The use of high pressure water jets for decontamination falls somewhere
between steam cleaning and abrasive  blasting in effectiveness.  Less effective than abrasive
blasting, it has the advantage of producing liquid wastes (that can be processed) rather than
solid wastes.  As an external cleaning technique, it offers reduced airborne generation
potential although this is offset by the need to control splashing.  The utility of hydrolasing is
generally limited to operations where internal accessibility is not required.

Ultrasonic Cleaning.  Since this is an immersion process that is limited to smaller items, it  is
generally unsuitable for large scale decontamination.  Although ultrasonic cleaning can be
especially effective in removing contamination  from crevices, it is doubtful that releasable
levels can be reached consistently with this technique to make it a viable option.

Electropolishing. This is an electrochemical  process where the object to be decontaminated
serves as the anode in an electrolytic  cell and radioactive contamination on the item is
removed by anodic dissolution of the surface material.  Although it is a relatively new
process and has not yet been used for a full scale decontamination operation, it nevertheless
requires consideration as a technique  on  the basis of its superior effectiveness in cleaning
almost any metallic surface to a completely  contamination-free state.  On the other hand, this
process has several limitations including the size of contaminated objects, the cost of
electrolyte and special equipment, the consumption of considerable power, and the production
of highly radioactive solutions.

Chemical Decontamination.  Chemical flushing is recommended for remote decontamination
of intact piping systems and their components.  This technique  uses concentrated or dilute
                                          A5-2

-------
solvents in contact with the contaminated item to dissolve either the contamination film
covering the base metal or the base metal itself. Dissolution of the film is intended to be
nondestructive to the base metal, and is generally used for operating facilities. Dissolution of
the base metal, however, can be considered in a decommissioning program where reuse of the
item will not occur.                          '

Based on starting levels of contamination and required decontamination efforts, scrap metal
inventories at nuclear power plants can be grouped into the following four categories:

   •    Low-level Surface Contaminated.  This category is likely to consist of components that
       may be removed from previously specified buildings with significant residual  ,
       radionuclide inventories but involve systems that are completely divorced from
       primary coolant, coolant waste streams, and other media with substantial levels  of
       radioactivity. A sizeable fraction of scrap metal within this category will exhibit
       contamination that is limited to external surfaces and not exceed 100,000 dpm/100
       cm2. Decontamination strategies are likely to be routine with essentially 100%  success
       at achieving scrap metal release limits.

   *    Medium-level Surface Contaminated. Metal components in direct contact with
       contaminated media that is below that of primary coolant and liquid radwaste may
       have internal and/or external surface contamination between 100,000 and 10,000,000
       dpm/100 cm2.  Scrap metal in this category  requires substantial decontamination
       efforts with less than 100% success in achieving unrestricted release.

   •    High-level Surface Contaminated.  Scrap metal in this category will be represented by
       systems internally exposed to and contaminated by primary coolant and liquid
       radwastes leading to contamination levels in excess 107 dpm/100 cm2.' Reduced and
       variable fractions of metals .are likely to be decontaminated to a level that permits
       unrestricted release.

   •    Volumetrically Contaminated.   Components proximal to the reactor core may contain
       volumetrically-distributed activation products that range from nominal levels to
       extremely high levels.  (Some of these components may also have high surface
       contamination.) With exception of melt-refining of such activated metals, removal of
    s   contaminants by standard processes is not achievable.
                                         A5-3

-------
Table A5-1 identifies examples of major components that under normal operating conditions
are likely to be grouped in the aforementioned categories. At a minimum, components that
can be reasonably excluded from recycling include the reactor vessel and reactor vessel
internals. For the Reference PWR, the reactor vessel is a right circular cylinder constructed
of carbon steel about 216 mm in thickness and lined with 4 mm of stainless steel.  The
reactor vessel weight exclusive of internals is estimated at about 400 metric tons  The vessel
internal structures support and constrain the fuel assemblies, direct coolant flow, guide in-core
instrumentation, and provide some neutron shielding.  Principal components include the lower
core support assembly (inclusive of core barrel and shroud) and the upper core support and
in-core instrumentation and support assemblies.  These reactor vessel internals are made of
Type 304 stainless steel and are estimated at about 190 metric tons.

The "Reference BWR reactor vessel is also constructed  of carbon steel with a stainless steel
liner. The collective weight of the reactor vessel, top  head, and internals, however,  is
estimated at 1,034 metric tons and is significantly higher than that of a PWR  Major reactor
vessel internals include (1) the core shroud, (2) shroud support plate, (3) core support plate,
(4) top fuel guide, (5) control rod guide tubes, (6) jet pumps, (7) shroud head and steam
separator assembly, (8) steam dryer assembly, (9) feedwater  spargers, and (10)  core spray
lines.

5.1    Identification of Contaminated Steel Components Suitable for Recycling

The elimination of scrap metal with significant levels of volumetrically-distributed activation
products in carbon and stainless steel from recycling consideration yields Reference BWR and
PWR systems/components defined in detail below (Tables A5-2A, B, C, D, and Tables A5-
3A, B, C).  These tables cite system components and their corresponding weights.  The
material composition of individual components have not been adequately defined. While a
considerable number of components could be identified to consist  exclusively of carbon steel
                 , "nl                    '          .      'I       '   J
or stainless steel, large quantities of steel exist as thick-walled carbon steel that is clad with
thin-walled stainless steel (e.g., large piping, valves, vessels, tanks).  When stainless steel
provides corrosion resistant cladding, it is in effect physically inseparable from its large
carbon steel component.  In other instances, a given item will consist of many independent
parts each having different material composition.  For example, a  recirculation pump may
have a carbon steel casing and base with stainless steel shaft, impellers, and other internals.  •
Although potentially separable,  segregation of such individual parts is labor intensive and may
                                         A5-4

-------
be precluded by worker exposure (and ALARA) considerations and/or economic factors.  A
prudent approach may, therefore, assume that all steel scrap containing nickel be categorized
as "stainless steel" (even if the nickel content is well below that of standard stainless steel
alloys) because it is easier to upgrade scrap by adding nickel and other alloying material than
it is to remove nickel for the production of mild steel or carbon steel.

         Table A5-1. Examples of Scrap Metal Grouping Based on Contamination
  Low-level Surface Contamination (on average < 100.000 dpm/100 cm2):  Minimal Effort
  to Decontaminate
    •   Structural metals in the turbine building, auxiliary building, and support buildings
    •   Control and instrumentation cables, cable trays
    •   Mechanical systems/piping not associated with primary coolant and radwastes
  Medium Surface Contamination (on average between 1 x 10s and 1 x 107 dpm/100 cm2):
  Substantial Effort Required to Decontaminate
    •   Containment spray recirculation
    •   Most auxiliary support systems
    •   BWR steam lines
    •   BWR turbines
    •   BWR condenser
    •'   Containment building crane,  refueling equipment, etc.
    •   Reactor building structural steel
    •   Fuel storage pool liner and water cleanup system
  High Internal Surface Contamination (on average > 1 x 107dpm/100 cm2): Aggressive
  Decontamination Required with <100% Success
    •   PWR primary recirculation piping
    •   PWR primary pumps and valves
    •   - Liquid radwaste systems/tanks
    •   PWR steam generators
    •   Primary water cleanup system
    •   PWR pressurizer
    •   Coolant letdown and cleanup
    •   Spent fuel pool cooling
  Significant Volumetric Contamination:  Decontamination Unachievable
    •   Reactor vessel
    •   Reactor vessel top head
    •   Reactor vessel internals
    •   Control rod drive lines
    •   Reactor building components proximal to pressure vessel (< 10%)
    •   Rebar (~ 1% of plant total)
                                        A5-5

-------
5.1.1   Reference BWR
For Reference BWR, a total of 29 contaminated systems is identified that are grouped by
location (i.e., Reactor Building, Radwaste Building, and Turbine Building) (Table A5-2A, B,
and C). Systems are cited in  alphabetical order and identify the system-average level of
contamination as previously defined in Table A5-1. Piping inventories for Reference BWR
have been quantified and segregated by plant locatipn in Table A5-2D.

In total, it is estimated that about 8.4 million kilograms or 8,400 metric tons of contaminated
steel exist in Reference BWR that is potentially available for recycling.  Based on material
composition data cited in NUREG/CR-0672, it is further estimated that of the total 8,400
metric tons of steel, nearly  1,700 metric tons represents stainless steel.  Stainless steel that is
physically associated with carbon steel may, however, not be readily segregated.

               Table A5-2A.  Reference BWR Steel Inventories by Location
                              Within the Reactor Building

                          System:  Containment Instrument Air
                       System Average Contamination Level:  Low
Number
22
1
1
222
TOTAL
Component ;
Instrument air accumulators
6" check valve
6" valve
Valves (3/4 - 2" dia.)

Weight (kg)
Each' :
129
68
82
NA

Total
2,838
68
82
4,008
6,996
                                         A5-6

-------
 Table A5-2A. Reference BWR Steel Inventories by Location
          Within the Reactor Building (Continued)
            System: Control Rod Drive System
System Average Contamination Level: 80% Low; 20% Medium
Number
460
225
185
370
210
2
2
2
2
2,660
TOTAL
-i
^
. .Component
CRD Blade
CRD Mechanism
Direction Control set
t
Scram valve
Scram accumulator
CRD Pump
Scram Discharge Volume
Pump Suction Filter
CRD drive water filter
Valves (3/4 - 4" dia.) & components

Weight (kg)
Each
182
218
36
32
64
1,816
908
182
45
NA

Total
83,720
49,050
6,660
11; 840
13,440
3,632
1,816
364
90
48,830
219,442
           System- Equipment Drain Processing
       System Average Contamination Level: Medium
Number
1 '
1
1
1
1
1
1
1
1
2
2
199
TOTAL
• . /
Component ' -
Waste demineralizer
Waste collector filter
Waste filter hold pump
Waste collector tank & educator
Waste collector pump
Spent resin tank
Spent resin pump
Waste surge tank & educator
Waste surge pump
Waste sample tank & educator
Waste sample pump
Valves (1 - 8" dia.)

Weight (kg)
Each
907
1,812
318
10,229
284
657
102
18,282
284
6,960
230
NA

% Total
907
1,812
318
10,229
284
657
102
18,282
284
13,920
462
5,374
52,631
                          A5-7

-------
Table A5-2A Reference BWR Steel Inventories by Location
         Within the Reactor Building (Continued)
         System: Fuel Pool Cooling and Cleanup
       System Average Contamination Level: High
Number
15
1
2
2
2
2
1
2
1
1
2
1
1
165
TOTAL
.» • rf s< i j .• .- ^
Component
* ' .-
Spent Fuel Racks
Fuel Pool Liner
FPCC pumps
FPCC demin
Skimmer Surge Tank
FPCC Heat Exchanger
Supp. Pool Cleanup Pump
Resin Tank Agitator
Fuel Pool Precoat Pump
(Precoat) Dust Evacuator
FPCC hold pump
FPCC Precoat Tank
FPCC Resin Tank
Valves (1 - 10" dia.) & components

, /Weight 
-------
 Table A5-2A. Reference BWR Steel Inventories by Location
          Within the Reactor Building (Continued)
               System:  HVAC Components
         System Average Contamination Level:  Low
Number
7
5
17
NA
TOTAL
Component
Containment Recirc. Fans
Containment Faa Coil Units
Emergency Fan Foil Units
Ducts (750 linear meters)

Weight (kg)
Each
636
1,500
955
NA

Total ;
4,452
7,500
16,235
29,975
58,162
             System:  Low Pressure Core Spray
       System Average Contamination Level: Medium
Number
2
1
1
1
1
45
TOTAL
Component
24" Suction Strainer
Vent Strainer
14 x 24" pump
Pump pit
1 x 2" pump
Valves (3/4 - 24" dia.)

Weight (kg)
-Each' i
172
43
9,625
182
82
NA

Total
344
43
9,625
182
82
10,523
20,799
                  System: Main Steam
System Average Contamination Level: 60% Medium; 40% Low
Number
1
2
2
2
1
2
Component
HP Turbine
LP Turbine
RFW Turbine •
Steam Chest
Gland Steam Condenser
Ejector Condenser
WeightXfcg)
Each
194,169
371,130
18,160
55,565
1,861
1,816
Total ' \
194,169
742,260
36,320
111,130
1,816
3,632
                         A5-9

-------
Table A5-2A.  Reference BWR Steel Inventories by Location
         Within the Reactor Building (Continued)
                Main Steam (Continued)
Number
1
1
2
2
4
2
2
4
18
36
18
1
6
6
8
1
4
2
1
2
2
2
2
8
951
TOTAL
Component
Moisture Separator
Bypass Valve Assy.
Moisture Separator Reheater
Steam Evaporator
2" Strainer
4" Strainer
12 Stop Check
30" Flow Restrictor
8" AO SRV
10" Vacuum Breakers
24 x 12" Quenchers
72" MOV
Stop Valves
Interceptor Valves
30" MSIV
24" MOV
24" Relief Valve
20" Relief Valve
16" MOV
16" Check Valve
14" Check Valve
14" MOV
12" MOV
28" HOV Governor Valves
Valves (1 - 10" dia.)

Weight (kg) '
Each
908
5,266
208,386
13,472
43
100
894
1,362
921
408
749
51,900
18,160
4,540
636
3,223
4,190
3,496
1,920
1,534
1,008
1,253
1,135
3,632
NA

Total
908
5,266
416,772
26,944
172
-200
1,788
5,448
16,578
14,724
13,482
51,900
108,960
27,240
5,088
3,223
16,760
6,992
1,920
3,068
2,016
2,506
2,270
29,056
69,592
1,922,200
                         A5-10

-------
 Table A5-2A.  Reference BWR Steel Inventories by Location
          Within the Reactor Building (Continued)
           System- Main Steam Leakage Control
         System Average Contamination Level:  Low
.Number
8 -
28
2
14
4
4
20
2
2
4
TOTAL
Component
1/2" Valve
3/4" Valve
1" Flow Element
1" Valve
1" Check Valve
1-1/2" Flow Element
1-1/2" MOV
1-1/2" Check Valve
MSLC Fan (3")
MSLC Heater

Weight (kg)
Each
11
14
17
23
17
•21
23
21
204
57

Total
88
392
34
322
.68
84
460
42
408
221
• 2,125
      System: Miscellaneous Items from Partial System
System Average Contamination Level:  55% Low; 45% Medium
Number
5
2
5
5
5
1 set
2
1
1
Component
TIP Drive Unit
TIP Indexing Unit
TIP Ball Valve
Explosive Shear Valve
TIP Shield Pig
TIP Tubing
Hogger (mechanical vacuum pump)
Refueling Bridge
Reactor Service Platform
Weight (Jcg)
Each .
361
9
23
23
154
295
3,171
24,918
5,210
Total
1,805
72
115
115
770
295
6,342
24,918
5,210
                         A5-11

-------
Table A5-2A. Reference BWR Steel Inventory by Location
        Within the Reactor Building (Continued)
   Miscellaneous Items from Partial System (Continued)
Number
2
1
1
1
1
1
1
2
1
185
1
1
20
9
4
1
1
TOTAL
Component
Refueling Mast
CRD Removal Turntable
CRD Removal Trolley
Incore Instrument Grapple
Fuel Support Piece Grapple
Control Blade Grapple
Spent Fuel Pool Work Table
Fuel Prep Machine
Channel Measurement Machine
Blade Guide
In Core Instrument Strongback
Manipulators, crows feet, etc.
In-vessel Manipulator Poles
Drywell Recirculation Fan
Stud Tensioner
RPV Head Strongback
Dryer/Separator Strongback

Weight (kg)
Each
295
2,492
173
36
41
59
445
381
422
73
100
136
14
254
1,044
2,134
60

Total
590
2,492
173
36
'41
59
445
762
422
13,505
100
136
280
2,286
4,176
2,134
60
67,339
                       A5-12

-------
Table A5-2A. Reference BWR Steel Inventories by Location
         Within the Reactor Building (Continued)
     System: Reactor Building, Closed Cooling Water
       System Average Contamination Level:  Low
Number
3
2
1
5 '
1
3
7
6
4
1
218
TOTAL
Component '
RBCCW Heat Exchanger
RBCCW Pump
RBCCW Surge Tank
Drywell Cooler & Fans
14" MOV
12" valve
10" MOV
10" Valve
10" Check Valve
10" Flow Element
Valves (3/4 - 8" dia.)

Weight (kg)
Each s
7,460
1,597
531
745
449
331
250
250
168
16
NA

Total
22,380
3,194
531
3,725
449
993
1,750
1,500
672
672
6455
42,321
   System:  Reactor Building Equipment and Floor Drains
      System Average Contamination Level:  Medium
Humber
4
3
1
1
97
TOTAL
Component
Drain Sump Pump
Drain Sump Pump
Equipment Drain Heat Exchanger
Drywell Equipment Drain HX
Valves (3/4' -6" dia.)

-Weight (kg) .
Each
523
650
680
680
NA

Total
2,908
1,950
680
680
3,725
9,943
                        A5-13

-------
Table A5-2A. Reference BWR Steel Inventories by Location
         Within the Reactor Building (Continued)
         System: Reactor Core Isolation Cooling
      System Average Contamination Level: Medium
Number
1
1
1
1
1
1
1
2
4
1
1
284
TOTAL
Component
Pelton Wheel Turbine/Pump
Barometric Condenser
Condenser Pump
Water Leg Pump
Vacuum Pump
Vacuum Tank
Steam Condensate Drip Pot
8" Suction Strainers
3/4" Steam Trap
10" Exhaust Drip Chamber
Turbine Exhaust Sparer
Valves (3/4 - 10" dia.)

.Weight (kg) ,,
Each
6,260
553
679
216
453
407
109
66
25
309
241
NA

Total
6,260
553
679
216
453
407
109
112
100
309
241
12,115
21,554
         System: Reactor Water Cleanup System
       System Average Contamination Level: High
Number
2
2
1
1
1
2
3
2
1
2
1
259
TOTAL
Component
RWCU Pump
Clean Up Hold Pump
Clean Up Precoat Pump
Sludge Discharge Pump
Decant Pump
Non-regenerative HX
Regenerative HX
Filter Demineralizer
Batch Tank
Phase Separator Tank
Precoat Agitator
Valves (1/2 - 6" dia.)

Weight (kg)
Each
590
534
454
284
102
4,086
4,131
3,178
227
2,043
27
NA

Total
1,180
1,068
454
284
102
8,172
12,394
6,356
227
4,086
27
13,170
47,520
                        A5-14

-------
Table A5-2A. Reference BWR Steel Inventories by Location
         Within the Reactor Building (Continued)
            System:  Residual Heat Removal
       System Average Contamination Level: Low
Number
3
1
1
1
1
6
2
3
2
1
11
8
5
3
2
4
4
2
3
2
3
3
3
1
2
1
324
TOTAL
Component
RHRPump
Water Leg Pump
Drywell Upper Spray Ring Header
Drywell Lower Spray Ring Header
Wetwell Spray Ring Header
Suppression Pool Suction Strainers
RHR Heat Exchanger
24" MOV
20" MOV
20" Valve
18" MOV
18" Valve
18" Check
18" Flow Element
18" Restricting Orifice
16" MOV
14" MOV
14" Valve
14" Air Operated Check
14" Restricting Orifice
12" MOV
12" Valve
12" Air Operated Check
12" Restricting Orifice
10" Valve
10" Check Valve
Valves (3/4 - 3" dia.)
*
' Weight (kg)
Each '
7,792
397
8,562
13,063
5,347
195
29,190
7,150
4,086
4,086
4,603
4,603
2,762
2,762
2,762
2,724
1,544
1,544
•971
944
1,017
1,017
581
549
731
399
NA

Total
23,376
397
8,562
13,063
5,347
1,171
58,380
21,450
8,172
4,086
50,633
36,828
13,810
8,286
5,524
10,896
6,176
3,088
2,913
1,888
3,051
3?051
1,743
549
1,462
399
12,100
306,401
                        A5-15

-------
 Table A5-2A. Reference BWR Steel Inventories by Location
          Within the Reactor Building (Continued)
              System:  Miscellaneous Drains
       System Average Contamination Level:  Medium
Number
1
1
174
TOTAL
Component
Misc. Drain Tank #1
Misc. Drain Tank #2 w/ pumps
• Valves (1" - 6" dia.)

Weight 
-------
Table A5-2B. Reference BWR Steel Inventories for Locations
         Within the Radwaste Building (Continued)

      Chemical Waste Processing System (Continued)
Number
2
2
2
2
1
2
2
2
293
TOTAL
Component
Decon Sol. Concentrator Tank
Decon Cone. Recycle Pump
Decon Concentrator Condenser
Decon Concentrator Pre Heater
Decon Concentrator Waste Pump
Chemical Waste Stream Mixer
Condensate Receiver Tank
Condensate Receiver Tank Pump
Valves (1" - 8" dia.)
t
Weight (kg)
. Each
711
843
2,305
3,143
254
111
950
102
NA

Total
1,422
1,686
4,610
6,286
-508
222
1,900
204
7,654
59,803
           System:  Condensate Demineralizers
      System Average Contamination Level: Medium
Mmiber
6
6
6
1
1
1
1
2
363
TOTAL
Component
Filter Demineralizers
Resin Trap (w/ basket)
Demin Hold Pump
Condensate Backwash Rcving Tank
Sludge Disc Mixing Pump
Condensate Decant Pump
Condensate Backwash Transfer Pump
Condensate Phase .Separator Tank
Valves & Components (1 - 36")

Weight (kg)
Each
5,300
953
159
6,912
420
420
420
3,178
NA

Total
31,800
5,718
954
6,912
420
420
• 420
6,356
36,783
89,783
                         A5-17

-------
Table A5-2B. Reference BWR Steel Inventories for Locations
         Within the Radwaste Building (Continued)
               System:  HVAC Components
        System Average Contamination Level: Low
Number
11
3
NA
TOTAL
Component .
J •> •! •*''
RadWaste Air Handlers
Filter Units and Fans
Ducts (1,980 linear meters)

Weigtit (kg)
Each
1,327
11,123
NA

Total'
14,597
33,369
54,785
102,751
        System: Radioactive Floor Drain Processing
      System Average Contamination Level: Medium
Number
1
1
1
1
1
1
1
1
1
1
1
171
TOTAL
Component
Floor Drain Demineralizer
Floor Drain Sample Tank
Floor Drain Sample Pump
Floor Drain Filter Aid Pump
Floor Drain Filter Hold Pump
Floor Drain Filter
Floor Drain Collector Pump
Floor Drain Collector Tank
Waste Decant Pump
Waste Sludge Dsch Mixing Pump
Waste Sludge Phase Sep Tank
Valves (1/2 - 8" dia.)

Weight 
-------
Table A5-2B. Reference BWR Steel Inventories for Locations
         Within the Radwaste Building (Continued)
           System: Rad Waste Building Drains
        System Average Contamination Level:  High
Number
1
2
3
38
TOTAL
Component
Chemical Drain Sump Pump
EDR Sump Pump
FDR Sump Pump
Valves & components (3/4 - 3" dia.)

Weight 
-------
Table A5-2C. Reference BWR Steel Inventories by Location
              Within the Turbine Building
             System:  Feed and Condensate
      System Average Contamination Level:  Medium
Number
2
3
3
1
2
1
2
2
3
3
3
3
2
2
2
407
TOTAL
Component
Turbine and Feed Pump
Condensate Booster Pump
Condensate Pump
Gland Exhaust Condenser
Air Ejector Condenser & Ejectors
Off Gas Condenser
#6 Feedwater Heater
#5 Feedwater Heater
#4 Feedwater Heater
#3 Feedwater Heater
#2 Feedwater Heater
#1 Feedwater Heater
Condensate Storage Tanks
Seal Steam Evaporator
Seal Steam Evap. Slowdown Cooler
Valves (1/2 - 24" dia.)
>
Weight (kg)
Each
54,821
12,006
21,883
4,032
6,614
897
73,394
68,863
35,338
50,288
51,194
62,974
50,475
13,451
213
NA

Total
109,642
36,018
65,649
4,032
13,228
897
146,788
137,726
106,014
150,864
153,582
188,922
100,950
26,902
426
350,478
1,592,118
                        A5-20

-------
Table A5-2C. Reference BWR Steel Inventories by Location
         Within the Turbine Building (Continued)
               System:  Extraction Steam
      System Average Contamination Level:  Medium
Number
6
6
10
10
5
5
2
2
6
4
4 ,
10
12
85
TOTAL
Component
24" MOV
24" Stop Check
20" MOV
20" Stop Check
18" MOV
18" Stop Check
16" MOV
16" Stop Check
8" AOV
6" MOV
4" AOV
2" AOV
2" Restricting Orifice
Inst. root (typ,3/4" globe)

Weight (kg)
Each
3,223
2,583
, 2,633
• 2,107
2,225
1,780
1,920
1,536
511
267
122
34
25
15

- Total
19,338
15,498
26,330
21,070
11,125
8,900
3,840
3,072
3,066
1,068
488
340
300
1,275
115,710
           'System: Heater Vents and Drains
      System Average Contamination Level:  Medium
Number
2
2
2
4
4
841
TOTAL
Component '
Steam Evaporator Drain Tank
Heater Drain Tank
Moisture Separator Drain Tank
Reheater Drain Tank
Reheater Drain Tank
Valves & Components (1-1/2 - 20" dia.)

Weight (kg)
Each' '•
898
6,274
1,715
1,134
6,274
NA

Total :
1,796
12,548
3,430,
4,536
- 25,096
151,369
198,775
                        A5-21

-------
Table A5-2C.  Reference BWR Steel Inventories by Location
         Within the Turbine Building (Continued)
              System:  HVAC Components
       System Average Contamination Level:  Low
Number
4
1
10
NA
TOTAL
Component
Exhaust Air Units
Standby Gas Treatment
Air Handlers & Filter Units
Ducts (1,000 linear meters)

Weight (kg) ,
Each
4,900
8,853
829
NA

Total
19,600
8,853
8,290
48,503
85,246
          System:  Offgas (Augmented) System
     System Average Contamination Level:  Medium
Number
2
2
1
1
2
2
2
8
2
2
2
4
2
2
2
9 '
18
175
TOTAL
Component
Catalytic Recombiner Vessel
Preheater Heat Exchanger
Offgas Condenser
Water separator
Lab Vacuum Pump
Lab Vacuum Pump
Water Separator
Charcoal Ads. Vessel
Cooler Condenser
Pre-filter Vessel
After-filter Vessel
Desiccant Dryers
Dryer Heater
Dryer Chiller
Regen. Blower
6" Air Operated Valve
6" Valve
Valves (3/4 - 4" dia.)

Weight (kg)
Each
453
538
897
271
45
45
1,359
4,077
906
1,133
1,133
622
3,625
2,265
636
82
82
NA

Total
906
1,076
897
271
90
90
1,718
32,615
1,812
2,266
2,266
2,488
7,250
4,530
1,272
738
1,476
2,722
64,483
                       A5-22

-------
Table A5-2C. Reference BWR Steel Inventories by Location
         Within the Turbine Building (Continued)
                 System: Recirculation
       System Average Contamination Level:  Low
Number
2
2
4
258
TOTAL
Component
Recirculation Pump w/motor
24" HOV
24" MOV ,
Valves (3/4 - 2" dia.)

Weight (kg)
Each
43,617
4,767
4,767
NA

Total
87,234
• 9,534
19,068
4,700
120,536
            System: Turbine Building Drains
      System Average Contamination Level:  Medium
Number
4
4
25
TOTAL
Component
Equipment Drain Sump Pump
Floor Drain Sump Pump
Small Valves (2 - 3" dia.)

Weight (kg) '
Each .
586
484
NA

Total
2,344
1,936
450
4,730
                        A5-23

-------
Table 5-2D. Reference BWR Piping Inventories by Plant Location
                     Reactor Building
           Average Contamination Level:  Medium
Piping Material
Carbon Steel
length (m)
weight (kg)
Stainless Steel
length (m)
weight (kg)
Total Stock (kg)
Outside Diameter (mm)
<60
2,323
8,479
6,169
18,674

73 - 254
3,922
110,368
500
4,551

305 - 406
505
61,897
54
2,143

457 - 610
952
127,160
—

660 - 762
55
14,850
—

914 - 1,829
—
—

Total
322,754
25,368
348,122
                    Primary Containment
             Average Contamination Level: High
Piping Material T -
Carbon Steel
length (m)
weight (kg)
Stainless Steel
length (m)
weight (kg)
Total Stock (kg)
Outside Diameter (mm)
- <60
263
1,366
3,850
10,603

73 - 254
1,084
63,181
110
3,411

305 - 406-
211
29,760
64
8,789

457 - 610
1,239
554,877
55
21,440

660 - 762
374
145,312
. —

914 - 1,829
559
234,882
—

Total
1,029,378
44,243
1,073,621
                          A5-24

-------
Table 5-2D.  Reference BWR Piping Inventories by Plant Location (Continued)
                           Turbine Building
                   Average Contamination Level:  Low
Piping Material
Carbon Steel
length (m)
weight (kg)
Stainless Steel
length (m)
weight (kg)
Total Stock (kg)
'; Outside Diameter (mm),' - ;
<60
3,336
14,153
H^H

; 73-254.
2,632
115,525
38
1,474

305 - 406
1,647
176,600
103
6,421

457 - 610
1,832
386,321
—

'660-762
465
240,698
—

914 - 1,829
559
234,882
—

total '
1,168,179
7,895
1,176,074
                     RadWaste and Control Buildings
                   Average Contamination Level:  High
Piping Material
Carbon Steel
length (m)
weight (kg)
Stainless Steel
length (m)
weight (kg)
Total Stock (kg)
Outside Diameter (mm) - :,
<60
3,087
10,267
1,150
4,747

73 - 254
3,337
75,778
1,026 .
10,164

305-406
338
29,221
55
1,756

457 - 610 .
12
4,584
**

£60 - 762
—
, —

914 - 1,329'
99
29,410
—

, total.
149,260
16,667
165,927
                                A5-25

-------
5.1.2  Reference PWR

Data in Table A5-3A, B, and C identify major contaminated PWR components by function
and location.  The total inventory of steel (excluding the reactor pressure vessel and its
internals) is estimated at about 4,100 metric tons.  It should be pointed out, however, that
slightly less than half or about 2,000 metric tons are contributed by primary system
components that include steam generators, pressurizer, reactor coolant piping, etc. (Table
A5-3 A). The long-term buildup of activated corrosion products and leaked fission products
on internal surfaces among these components are projected to be high. Even with intense and
aggressive decontamination efforts, their release for unrestricted recycling may not be
technically achievable or may be precluded by radiological and other concerns.
                  i            i       i   «    '
                  rn      "                "   '
The balance of about 2,100 metric tons includes eleven (11) internally contaminated reactor
support systems and piping that are associated with the Auxiliary Building/Fuel Storage
facility and a variety of structural components where contamination is limited to external
surfaces. It is estimates that nearly 20% of all steel is stainless steel.
         Table A5-3A.  Reference PWR Contaminated Steel Inventories by Location
                              Within the Reactor Building

                     System:  External Surface Structures Equipment
              System Average Contamination Level: 70% Low; 30% Medium
Number
NA
NA
NA
NA
NA
NA
NA
TOTAL
Component
Refueling Cavity Liner
Base Liner
Reactor Cavity Liner
Floor and Cavity Liner Plates
CRD Missile Shield
Stairways/Gratings
Miscellaneous Equipment
.
Weight 
-------
 Table A5-3A.  Reference PWR Steel Inventories by Location
         Within the Reactor Building (Continued)
System:  Internally Contaminated Primary System Components
        System Average Contamination Level: High
Number \
4
4
1
NA
1
4
1
2
1
1
NA
TOTAL
t ' ••*
Component ,
Steam Generator
Rx Coolant Pumps
Pressurizer
Containment Spray Piping
Pressurizer Relief Tank
Safety Inject. Syst. Accumulator
Reactor Cavity Drain Pump
Containment Sump Pump
Excess Letdown Heat Exchanger
Regenerative Heat Exchanger
Reactor Coolant Pining Data
Size: 686 - 787 mm ID/Length 81 m
Size: 51-356 mm OD/Length 677 m

Weight (kg)
Each ,
312,000
85,350
'88,530

12,338
34,700
363
635
726
2,994


Total
1,248,000
341,400
88,530
90,800
12,338
138,800
363
1,270
726
2,994
100,698
11,793
2,037,712
                        A5-27

-------
Table A5-3B. Reference PWR Steel Inventories by Location
      Within the Auxiliary Building and Fuel Storage
       System:  Component Cooling Water System
       System Average Contamination Level: Low
Number
2
2
1
1
9
169
TOTAL
j- ^
f S
f
, Component
CCW Heat Exchanger
CCWPump
CCW Surge Tank
Chem. Addition Tank
Sample Heat Exchanger
Valves (3/4 - 24" dia.)

Weight (kg)
Each
31,780
6,810
908
477
3,178


Total
63,560
13,620
1,816
954
28,602
104,700
213,252
           System:  Containment Spray System
     System Average Contamination Level: Medium
Number
2
2
1
6
6
46
TOTAL
i-
Component
t •> - s ?- i ••
Pump
Pump
Tank
Small Electrical Equipment
Large Electrical Equipment
Valves (3/4 - 18" dia.)

Weight (kg)
Each
3,087
45
2,490
34
68
NA

Total'
6,174
90
2,490
204
408
37,875
47,241
                        A5-28

-------
Table A5-3B.  Reference PWR Steel Inventories by Location
Within the Auxiliary Building and Fuel Storage (Continued)
   .System:  Clean Radioactive Waste Treatment System
      System Average Contamination Level:  Medium
Number
1
2
1
1
2
2
2
2
1
2
1
2
1
1
1
1
83
TOTAL
'Component
Rx Coolant Drain Tank
Rx Coolant Drain Pump
Rx Coolant Drain Filter
Spent Resin Storage Tank
Clean Waste Recv. Tank
Clean Waste Recv. Pump
Treated Waste Mon. Tank
Treated Waste Mon. Pump
Aux. Building Drain Tank
Aux. Building Drain Pump
Chem. Waste Drain Tank
Chem. Waste Drain Pump
Waste Cone. Hold Tank
Waste Cone. Hold Pump
Clean Waste Filter
Clean Radwaste Evaporator
Valves (2-3" dia.)

' Weight (kg)
Each'
758
227
159
3,087
4,975
227
5,085
104
949
590
,2,452
91
949
104
30
18,160
NA

Total
758
454
159
3,087
9,950
454
10,170
208
949
1,180
'2,452
182
949
104
30
18,160
3,935
53,181
                        A5-29

-------
Table A5-3B. Reference PWR Steel Inventories by Location
Within the Auxiliary Building and Fuel Storage (Continued)
           System: Control Rod Drive System
       System Average Contamination Level: Low
Number
4
4
1
TOTAL
/ ? <
Component
Small Electric Equipment
Large Electric Equipment
Large Mech. Equipment

Weight (kg)'
Bach_
34
68
68

Total ,
136
272
68
476
     System:  Electrical Components and Annunciators
       System Average Contamination Level:  Low
Number
2
2
1
1
1
7
7
1
12
2
22
TOTAL
Component
125 VDC Power (Small)
125 VDC Power (Medium)
125 VDC Power (Large)
4.16 KV AC & Aux. (Small)
4.16 KV AC & Aux. (Large)
480 KV AC Ld Cntr (Small)
480 KV AC Ld Cntr (Large)
480 KV AC MCC
480 KV AC MCC
Annunciators (elec. port.)
Annunciators (mech. port.)

Weight (kg)
Each
68
227
2,270
227
9,080
227
908
227
9,080
34
34

Total
136
454
2,270
227
9,080
1,589
6,356
227
108,960
68
748
130,115
                        A5-30

-------
Table A5-3B. Reference PWR Steel Inventories by Location
Within the Auxiliary Building and Fuel Storage (Continued)
      System: Chemical and Volume Control System
       System Average Contamination Level: High
Number
3
1
1
1
2 •
1
3
2
2
1
1
1
2
1
2
1
2
1
3
2
2
1
1
2
2
1
1
378
TOTAL
Component
Regenerative Heat Exchanger
Seal Water Heat Exchanger
Letdown Heat Exchanger
Excess Letdown Heat Exchanger
Centrif. Charge Pump
Volume Contrpl Tank
Holdup Tank
Monitor Tank
Boric Acid Tank
Batch Tank
Resin Fill Tank
Reciprocal Charge Pump
Boric Acid Pump
Reactor Coolant Filter
Mixed Bed Demineralizer
Cation Ion Exchange
Seal Injection Filter
Concentrate Hold Tank
Evaporator Feed Ion Exchange
Evaporator Condensate Ion Exchange
Condensate Filter
Concentrates Filter '
Cone. Hold Tank Transfer Pump
Gas Stripper Feed Pump
Boric Acid Evap. Skid Assembly
Ion Exchange Filter
Recirculation Pump
Valves (3/4 - 6" dia.)

Weight (kg)
Eacf*
2,724
772
863
726
7,759
2,202
13,620
9,080
9,080
658
118
8,036
281
91
477
477
749
1,589 "
477
477
18
18
91
227
9,489
68-
288
NA

Total
8,172
111
863
726
15,518
2,202
40,860
18,160
18,160
658
118
8,036
562
91
954
477
1,498
1,589
1,431
954
18
18
182
454
18,978
68
288
17,481
159,288
                        A5-31

-------
Table A5-3B.  Reference PWR Steel Inventories by Location
Within the Auxiliary Building and Fuel Storage (Continued)
    System:  Dirty Radioactive Waste Treatment System
     System Average Contamination Level: Medium
Number
1
2
1
2
1
2
2
46
TOTAL
Component
Rx Cavity Drain Pump
Rx Cont. Sump Pump
Dirty Waste Monitor Tank
Dirty Waste Monitor Tank Pump
Dirty Waste Drain Tank
Dirty Waste Drain Tank Pump
Aux Building Sump Pump
Valves (2 -3" dia.)

Weight (kg)
Each.
363
681
2,633
91
2,969
181
590
NA

Total
363
1,362
2,633
182
2,969
362
1,180
2,280
11,331
       System:  Radioactive Gaseous Waste System
     System Average Contamination Level: Medium
Number
1
4
2
2
2
1
2
4
2
1
83
TOTAL
Component
Surge Tank
Decay Tank
Gas Compressor
Moisture Separator
HEPA/prefilter
Exhaust Fan
Br. Seal Water Heat Exchanger
Large Electrical Equipment
Large Mechanical Equipment
HVAC Equipment
Valves (3/4 - 4" dia.)

Weight (kg)
Each,
404
4,900
3,632
45
91
45
3,496
68
2,270
68
NA

Total
404
19,600
7,264
90
182
45
6,992
272
4,540
68
4,607
44,064
                       A5-32

-------
Table A5-3B. Reference PWR Steel Inventories by Location
Within the Auxiliary Building and Fuel Storage (Continued)
         System:  Residual Heat Removal System
       System Average Contamination Level: High
Number
2
2
12
11
1
42
TOTAL
Component
Pump
Heat Exchanger Unit
Small Electrical Equipment
Large Electrical Equipment
Small Mechanical Equipment
Valves (3/8 - 14' dia.)

, Weight (kg)
Each
3,087
10,487
34
68
34
NA

Total
6,174
20,974
408
748
34
49,032
77,370
            System: Safety Injection System
     System Average Contamination Level:  Medium
Number
4
1
2
1
1
10
10
1
89
TOTAL
u t
Component
Accumul. Tank
B. Inj. Tank
Safety Inj. Pump
Refueling Water Tank
Primary Water Storage Tank
Small Electrical Equipment
Large Electrical Equipment
Small Mechanical Equipment
Valves (3/4 - 10" dia.)
1
Weight 
-------
Table A5-3B. Reference PWR Steel Inventories by Location
Within the Auxiliary Building and Fuel Storage (Continued).
               System:  Spent Fuel System
       System Average Contamination Level:  High
Number
1
2
1
2
1
1
2
53
1



TOTAL
s •,
Component
Pump
Pump
Pump
Filter
Filter
Demineralizer
Heat Exchanger
Valves (3/4 - 10" dia.)
Fuel Pool Liner
Fuel Storage Racks
Fuel Handling System
Overhead Crane

Weighing)
Each
• 454
409
318
163
68
998
2,769
NA
37,000




Total
454
918
318
326
-68
998
5,538
14,117
37,000
49,079
18,470
113,000
240,286
              Structural Steel Components
           Average Contamination Level: Low
Number
NA
NA
NA
NA
NA
NA
TOTAL
Composeat
Wall Support
Roof Support
Stairs/Grates/Tracks/Hand-rails
I-beams
HVAC Ducts
HVAC Components
•
Weight 
-------
          Table A5-3C. Reference PWR Non-RCS Stainless Steel Piping1
(a) (b)
Nominal
Size, In,
1/2

3/4


1


1-1/2


2


3
4
6
8
10
12
14
TOTAL
Schedule
• 80
160
40
80
160
40
80
160
40
80
160
40
80
160
160
160
160
160
140
140
140

Lineal Meters
122
122
122
183
580
61
61
427
122
335
549
305
488
1,067
140
183
311
143
192
88
100

Total inside
• Surface Area (tti2)
5.3117
4.5140
8.0163
10.8278
28.2806
5.1026
13.9652
22.7504
15.6628
40.1299
58.5751
50.2718
75.4540
143.6043
29.3566
50.1697
128.6770
70.6775
134.0704
74.0579
92.2988
1,061.7644
Total Weight (kg)
198
238
205
400
1,671
152 '
590
1,803
493
1,810
3,967
1,655
3,642
11,840
2,985
6,128
20,972
15,923
29,750
18,370
24,474
147,266

-------
5.1.3  Summary Estimates of Steel for Reference BWR/PWR and the Commercial Nuclear
       Industry

Table A5-4 presents summary data for contaminated steel potentially available for recycling.
Estimates for the Reference BWR and PWR were derived by summing component mass
values previously cited in Table A5-2 and Table A5-3, respectively.  Fractional quantities of
stainless steel were developed from information provided by Bryan and Dudley (1974) as well
as NUREG/CR-0672 and NUREG/CR-0130.  For example, Table A4-2 in Section 4.1 of this
report cited a total  stainless steel quantity used in the construction of a Reference PWR.
Based on PWR design parameters, it was concluded that all of the 1,154.6 metric tons of
stainless steel in the Reactor Plant Equipment and 21.1 metric tons for Fuel Storage were
contaminated for a,, total of about 1,175 metric tons.  Included in this total, however, was
about 348 metric tons of stainless steel that is volumetrically  contaminated with activation
products that are considered unsuitable for recycling. Thus, Table A5-4 cites a net mass of
827 metric tons of stainless steel, which, when subtracted from the total mass of 4,138 metric
tons of contaminated steel,  leaves the remainder of 3,311 metric tons of contaminated carbon
steel.                                                                   ' ,

Estimates for the entire commercial nuclear industry were derived by taking Reference BWR
and Reference PWR values and applying previously  cited plant-specific scaling factors for
each of the 40 BWRs and 83  PWRs (see Addendum #1).  Approximately 600,000 metric tons
of contaminated  steel over time may become available for recycling.  About  80% of the
contaminated steel is carbon steel with stainless steel representing the balance.

Because past and current regulatory release criteria (i.e., U.S. NRC Regulatory Guide 1.86)
are defined in activity levels per unit surface area, information cited in this section has been
presented in this fashion. However, for risk analysis pertaining to recycling of contaminated
scrap metals, a more meaningful approach is to express contamination levels in terms of
activity per unit mass.  This conversion required the derivation of the average mass thickness
(g/cm2) of contaminated metal surfaces by the following equation:
                 , .    „_ . ,     ' ,  ,   ,,       Contaminated Surface Areas (cm2)
         Average Mass Thickness (glcnr) = =*- - _ - -
                                                Contaminated Metal Mass (g)
                                        A5-36

-------
For the contaminated systems/components previously identified for Reference BWR and
PWR,  a weighted average mass density of 3.5 g/cm2 for contaminated surfaces was estimated.
At a density of about 8 g/cm3 for steel, this corresponds to an average thickness of about  4.4
mm (0.17 inches).  This average mass thickness can now be readily applied to estimate the   ,
activity level per unit mass of contaminated steel. For example, under the current interim
release criteria of 5,000 dpm/100 cm2 for beta-gamma emitters, the residual contamination on
average would correspond to about 14 dpm/g (or about 6.5 pCi/g.; or 0.23 Bq/g) of steel.
              Table A5-4.  Summary Data for Contaminated Steel Inventories
                             Potentially Suitable for Recycling
- -
Contaminated Material0*
\ ••
Stainless Steel
• Low-level Contamination
(<1 x 10s dpm/100 cm2)
« Medium-level Contamination
(1 x 10s to 1 x 107 dpm/100 cm2)
• High-level Contamination
(>1 x 107 dpm/100 cm2)
Carbon Steel
• Low-level Contamination
(<1 x 10s dpm/100 cm2)
• Medium-level Contamination
(1 x 10s to 1 x 107 dpm/100 cm2)
• High-level Contamination
(>1 x 107 dpm/100 cm2)
TOTALS
' Quantity (metric tons)
Reference
BWRS*
1,688
576

786

326

6,754
2,306

3,146

1,302

8,442
Reference
PWR<*>
827
210

114

503

3,311
841

458

2,012

4,138
All
BWRs8*
56,987
19,446

26,535 ,

11,006

228,015
77,851

106,209

43,955

285,002
All -
PWRsw
64,738
16,439

8,924

39,375

259,185
65,834

35,852

157,499

323,923
Total
Industry
121,725
35,885

35,459

50,381

487,200
143,685

142,061

201,454

608,925
  (b)

  (c)
Although data for stainless steel and carbon steel are presented as independent quantities, it
must be acknowledged that a significant fraction of stainless steel is unlikely to be segregated
as such for recycling purposes.
BWR radionuclide contaminants reflect those of Table A3-11 and Table A3-21.
PWR radionuclide contaminants reflect those of Table A3-17 and Table A3-20.
                                          A5-37

-------
5.2    Metal Inventories Other Than Steel

Although steel is clearly the dominant metal used in the construction and system components
that define a nuclear power plant, there are also significant quantities of other metals that may
be suitable for recycling.  Tables A4-1 and A4-3 in Section 4 of this report had identified •
total plant inventories for Reference BWR and Reference PWR for the following metals and
metal alloys: (1) galvanized iron, (2) copper, (3) inconel, (4) lead, (5) bronze, (6) aluminum,
(7) brass, (8) nickel, and (9) silver.  However, there exist no credible data in the open
literature regarding the estimated fraction(s) of these metal inventories that are likely to be
contaminated or the extent of their contamination.
In the absence of reported data, a reasonable approach may assume that the contaminated
fraction among total plant inventories of these metals parallels the contaminated fraction  of
carbon steel for Reference BWR and Reference PWR. Justification for this modeling
approach is based on the fact that most of these metals exist as sub-components of larger
items consisting primarily of carbon steel.  From data cited in previous sections, the percent
of contaminated carbon steel suitable for recycling to that of total plant inventory corresponds
to 20% and 10% fqr Reference BWR and Reference PWR,  respectively.  The application of
these values yields contaminated metal quantities (suitable for recycling) cited in Table A5-5.
Due to physical differences and chemical properties that affect corrosion and internal
contamination, a parallel approach to quantify metal inventories as low-level, medium-level,
and high-level contaminated seems inappropriate.

                Table A5-5. Summary of Metal Quantities Other than Steel
                               ,      (metric tons)
Metal
Type
Galvanized Iron
Copper
Inconel
Lead
Bronze
Aluminum
Brass
Nickel
Silver
Reference Facility
BWR.
258
137
24
9.1
5.0
3.6
2.0
0.2
<0.2
PWR .
130
69
12
4.6
2.5
1.8
1.0
0.1
<0.1
Industry ,
AllBWRs
8,710
4,625
810
307
169
122
68
7
<7
All PWRs
10,037
5,327
927
355
193
139
77
8
<8
Total
18,747
9,952
1,737
662'
362
261
145
15
<15
                                        A5-38

-------
5.3    Time-Table for the Availability of Scrap Metal from the Decommissioning of Nuclear
       Power Plants
For currently operating nuclear power plants, an operational period of 40 years is assumed.
The projected year of shutdown for each of 123 reactor units is identified in Addendum #1.
Following reactor shutdown, a minimum of 10 years is assumed before significant
dismantling activity could proceed that yields significant quantities of scrap metal. Thus, for
currently operating reactors, the earliest dates for releasing scrap metal are defined by their
startup dates plus 50 years.  Currently, there are 8 reactor units that have been permanently
shutdown (Dresden-1 (1984); Indian Point-1 (1980); LaCrosse  (1981); TMI-2 (1979);
Humboldt Bay (1976); Trojan (1993); Rancho Seco (1989); San Onofre-1 (1992); and Yankee
Rowe (1992)). A conservative assumption for these facilities projects the release of scrap
metal over a ten-year period between 2000 and 2009.

Table A5-6 summarizes the availability of scrap for yearly intervals starting with the year
2010.  The incremental quantity of scrap metal available for recycling is illustrated in Figure
A5-1.  The release of scrap metal based on this time-table must, however, be considered
highly conservative since many, if not  most, facilities are likely to delay D&D activities for
varying portions  of the allowable  50-year SAFSTOR period.

                Table A5-6.  Time-Table for Available Scrap Metals from
                         Decommissioned Nuclear Power Plants
                                Quantities (metric tons)
Year
2000-
2009
2010
2011
2012
2013
2014
2015
2016
2017
2018
2019
CS
15,377
804
—
—
—
3,616
3,616
—
6,464
14,811
1,763
SS
4,107
105
—
—
, —
475
475
—
1,958
2,712
534
Galv.
Iron
609
31

—
^ —
140
140
—
257
580
70
Copper
323
17
— '
—
—
75
75
—
136
308
37
laconel
57
3
—
—
—
13
13
- —
24
54
7
t«ad
21
1
—
—
—
5
5
—
9
20
2
Bronze
12
<1
. —
—
—
3
3
—
5
11
1
Alum,
8
<1
—
—
__
2
2
—
4
8
* " 1
Brass
5
<1
—
—
/
1
1
—
2
4
<1
Kickel
0.5
—
—
—
—
—
—
—
0.2
0.5
0.1
                                        A5-39

-------
Table A5-6.  Time-Table for Available Scrap Metals from
   Decommissioned Nuclear Power Plants (Continued)
              - Quantities (metric tons)
Year
2020
2021
2022
2023
2024
2025
2026
2027
2028
2029
2030
2031
2032
2033
2034
2035
2036
2037
2038
2039
2040
2041
2042
2043
2044
2045
2046+
Total
CS
15,442
2,438
10,328
38,415
41,117
12,927
22,927
5,611
9,574
—
9,078
8,961
10,697
11,191
30,466
26,173
32,396
13,192
9,637
8,366
12,956
—
3,261
—
—
2,703
12,868
397,175
SS
2,740
739
1,674
8,571
8,855
3,175
5,326
1,700
2,185
—
2,750
2,714
1,406
2,468
5,672
6,206
6,248
3,996
2,919
1,639
3,925
—
988
—
—
819
3,902
90,983
Galv.
Iron
604
97
370
1,512
1,616
510
903
223
377
—
362
357
415
440
1,193
1,032
1,269
525
384
328
516
—
130
—
—
107
512
15,609
Copper1
321
51
197
439
859
271
480
118
200
—
192
189
' 221
234
635
548
675
278
203
174
273
—
69
—
—
57
271
7,926
Inconel
56
9
34
141
151
48
84
21
35
—
34
34
38
41
111
96
118
49
36
30
48
—
12
—
—
10
48
1,455
Lead
21
3
13
53
57
18
31
8
13
—
13
12
15
16
42
36
45
18
13
12
18
—
5
—
—
4
18
547
Bronze
12
2
7
29
31
10
17
4
7
—
7
7
8
8
23
20
24
10
7
6
10
—
2
—
—
2
10
299
Alxjm. :
8
1
5
21
22
7
12
3
5
—
5
5
6
6
16
14
17
7
5
4
7
- —
2
—
—
1
7
212
Brass
4
<1
3
12
12
4
7
2
3
—
3
3
3
3
9
8
10
4
3
2
4
—
1
—
—
1
4
121
Kickel
0.5
0.1
, 0.3
1.2
1.3
0.4
0.7
0.2
0.3
—
0.3
0.3
0.3
0.3
1.0
0.8
1.0
0.4
0.3
0.2
0.4
—
0.1
—
—
0.1
0.4
12.2
                       A5-40

-------
100
 20
I     II     II
I     I     I     I     I
     2010  2015  2020  2025   2030   2035  2040  2045  2050+
                                 Year

                               i
  Figure A5-1.  Cumulative Availability of Scrap Metal from Nuclear Utilities
                              A5-41

-------
Page Intentionally Blank

-------
                                   REFERENCES

Bryan, R.H. and I.T. Dudley, 1974, "Estimated Quantities of Materials Contained in a 1000-
       MW(e) PWR Power Plant," ORNL-TM-4515, prepared by Oak Ridge National
       Laboratory for the U.S. Atomic Energy Commission.

Decommissioning Cost Study for the Big Rock Point Nuclear Plant, Consumers Power
       Company, February 1995

Nuclear Engineering International, 1991, World Nuclear Industry Handbook 1991. Surry,
       England.

Nuclear Engineering International, 1992, World Nuclear Industry Handbook 1992, Surry,
       England.

Nuclear Engineering International, 1993, World Nuclear Industry Handbook 1993. Surry,
       England.

NUREG/CR-0130, 1978, "Technology, Safety and Costs of Decommissioning a Reference
       Pressurized Water Reactor Power Station," Vol.  1, prepared by Smith, R.I., et al.,
       Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/CR-0130, 1978, "Technology, Safety and Costs of Decommissioning a Reference
       Pressurized Water Reactor Power Station," Vol.  2, Appendices, \prepared by Smith,
       R.I., et al., Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/CR-0672, 1980, "Technology, Safety and Costs of Decommissioning a Reference
       Boiling Water Reactor Power Station," Vol. 1, Main Report, prepared by Oak, H.D., et
       al., Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/CR-0672, 1980, "Technology, Safety and Costs of Decommissioning a Reference
       Boiling Water Reactor Power Station," Vol. 2, Appendices, prepared by Oak, H.D., et
       al., Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.
                                        R-l

-------
                              REFERENCES (Continued)

NUREG/CR-4289,  1986, "Residual Radionuclide Contamination Within and Around
       Commercial Nuclear Power Plants," prepared by Pacific Northwest Laboratory for the
       U.S. Nuclear Regulatory Commission.

NUREG-1496 (Draft), 1994, "Generic Environmental Impact Statement in Support of
       Rulemaking on Radiological Criteria for Decommissioning of NRC-Licensed Nuclear
       Facilities," U.S. Nuclear Regulatory Commission.

NUREG/CR-6174,  1994, "Revised Analyses of Decommissioning for the Reference Boiling
                •f'\            '          :,!                   -I
       Water Reactor Power Station," Vol  1, Main Report, prepared by Smith, R.I., et al,
       Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/CR-6174,  1994, "Revised Analyses of Decommissioning for the Reference Boiling
       Water Reactor Power Station," Vol 2, Appendices, prepared by Smith, R.L, et al,
       Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/CR-5884,  1995, "Revised Analyses of Decommissioning for the Reference
       Pressurized Water Reactor Power Station," Vol 1, Main Report, prepared by Konzek,
       GJ, et al., Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

NUREG/CR-5884,  1995, "Revised Analyses of Decommissioning for the Reference  .
      'Pressurized Water Reactor Power Station," Vol 2, Appendices, prepared by Konzek,
       G.J, et al., Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

SAFSTOR Decommissioning Plan for the Humboldt Bay Power Plant, Unit 3, Pacific Gas &
       Electric Company, July 1994.

San Onofre Nuclear Generating Station, Unit 1, Decommission Plan, Southern California
       Edison Co., November 1994.

Trojan Nuclear Plant Decommissioning Plan, PGE-1061, Portland General Electric, June
       1996.
                                        R-2

-------
                              REFERENCES (Continued)

U.S. Nuclear Regulatory Commission, 1988, General Requirements for Decommissioning ~
      Nuclear Facilities. Federal Register, Vol. 53, No. 123, June 27, 1988.

U.S. Nuclear Regulatory Commission Regulatory  Guide 1.86,  1974, "Termination of
      Operating Licenses for Nuclear Reactors," U.S. Nuclear Regulatory Commission. v

Yankee Nuclear Power Station Decommissioning Plan, Yankee Atomic Electric Company,
      February 1995.
                                         R-3

-------
Page Intentionally Blank

-------
                ADDENDUM #1
Demographic Data for the U.S. Nuclear Power Industry
Electric Utility
Name
Arizona Power Company
Arizona Power Company
Arizona Public Service
Arizona Public Service
Arizona Public Service
Arkansas Power & Light
Arkansas Power & Light
Baltimore Gas & Electric
Baltimore Gas & Electric
Boston Edison
Carolina Power & Light
Carolina Power & Light
Carolina Power & Light
Carolina Power & Light
Cleveland Electric
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Commonwealth Edison
Reactor
Name
Farley 1
Farley 2
Palo Verde 1
Palo Verde 2
Palo Verde 3
Arkansas Nuc. 1
Arkansas Nuc. 2
Calvert Cliffs 1
Calvert Cliffs 2
Pilgrim 1
Brunswick 1
Brunswick 2
Harris 1
Robinson 2
Perry 1
Braidwood 1
Braidwood 2
Byron 1
Byron 2
Dresden 1
Dresden 2
Dresden 3
LaSalle County 1
LaSalle County 2
Quad Cities 1
Quad Cities 2
Reactor
Type
PWR (West.)
PWR (West)
PWR (CE)
PWR (CE)
PWR (CE)
PWR (B&W)
PWR (CE)
PWR (CE)
PWR (CE)
BWR (GE)
BWR (GE)
BWR (GE)
PWR (West)
PWR (West)
BWR (GE)
PWR (West)
PWR (West.)
PWR (West.)
PWR (West)
BWR (GE)
BWR (GE)
BWR (GE)
BWR (GE)
BWR (GE)
BWR (GE)
BWR (GE)
Power
Rating

812
824
1,270
1,270
1,270
836
858
825
825
663
767
754
860
683
1,167
1,090
1,090
1,120
1,120
200
112
773
1,048
1,048
769
769
Scaling
Factor*
0.87
0.88
1.18
1.18
1.18
0.89
0.90
0.88
0.88
0.76
0.84
0.83
0.91
0.78
1.11
1.06
1.06
1.08
1.08
0.34
0.84
0.84
1.03
1.03
0.84
0.84
Projected
Year of
Shutdown
2017
2021
2024
2025 .
2027
2014
2018
2014
2016
2012
2016
2014
2026
2010
2026
2028
2028
2025
2027
1984
2010
2013
2024
2024
2013
2013
                    AD1-1

-------
Electric Utility
Name
Commonwealth Edison
Commonwealth Edison
Consolidated Edison
Consolidated Edison
Consumers Power
Consumers Power
Dairyland Power Co-op
Detroit Edison
Duke Power
Duke Power
Duke Power
Duke Power
Duke Power
Duke Power
Duke Power
Duquesne Light
Duquesne Light
Florida Power Corp.
Florida Power & Light
Florida Power & Light
Florida Power & Light
Florida Power & Light
Georgia Power
Georgia Power
Georgia Power
Georgia Power
GPU Nuclear
GPU Nuclear
Reactor -
Name
Zionl
Zion2
Indian Point 1
Indian Point 2
Big Rock Point
Palisades
LaCrosse
Enrico Fermi 2
Catawba 1
Catawba 2
McGuire 1
McGuire2
Oconee 1
Oconee 2
Oconee 3
Beaver Valley 1
Beaver Valley 2
Crystal River 3
St. Lucie 1
SL Lucie 2
Turkey Point 3
Turkey Point 4
Hatch 1
Hatch 2
Votlel
Votle2
Oyster Creek
Three Mile Island 1
Reacto?
Type , -
PWR (West.)
PWR (West.)
PWRCB&W)
PWR (West.)
BWR (GE)
PWR (CE)
BWR (AC)
BWR (GE)
PWR (West.)
PWR (West.)
PWR (West.)
PWR(WesL)
PWR(B&W)
PWR(B&W)
PWR(B&W)
PWR (West.)
PWR (West.)
PWR(B&W)
PWR (CE)
PWR (CE)
PWR (West.)
PWR (West.)
BWR (GE)
BWR (GE)
PWR (West.)
PWR (West.)
BWR (GE)
PWR(B&W)
Power
iRa&dg

1,040
1,040
265
931
67
755
50
1,060
1,129
1,129
< 1,129
1,129
846
846
846
810
833
820
839
839
666
666
744
762
1,105
1,103
610
808
Scaling
Factor
1.03
1.03
0.41
0.96
0.16
0.83
0.14
1.Q4
1.09
1.09
1.09
1.09
0.90
0.90
0.90
0.87
0.89
0.88
0.89
0.89
0.76
0.76
0.82
0.83
1.07
1.07
0.72
0.87
Projected
Year of
Shutdown
2013
2014
1980
2013
2000
2011
1981
2025
2025
2026
2021
2023
2013
2013
2014
2016
2026
2016
2016
2