ENVIRONMENTAL RADIATION
PROTECTION FOR
NUCLEAR POWER OPERATIONS
PROPOSED STANDARDS
i40 CFR 190i
SUPPLEMENTARY INFORMATION
JANUARY %, 1976
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ENVIRONMENTAL RADIATION PROTECTION
FOR NUCLEAR POWER OPERATIONS
PROPOSED STANDARDS
[10 CFR 190]
SUPPLEMENTARY INFORMATION
JANUARY 5, 1976
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PREFACE
As a result of the review of comments received on these proposed
environmental radiation protection standards for normal operations of
activities in the uranium fuel cycle, the Agency has identified a number
of areas in which additional information would be desirable in order to
provide a reasonable basis for discussion and comment on this proposed
rulemaking at the public hearing scheduled for February 17, 1976. This
mat'erial has been developed to supplement that contained in the notice
proposing these standards (40 FR 23420), as well as the draft environ-
mental statement and technical reports made available at that time. It
does not constitute a complete response to comments, since the public
record is still open. Modifications of the original proposal made as
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the result of comments received and a complete response to comments will
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be contained in the final environmental statment and notice of final
rulemaking, which will reflect all the information received, including
that developed at public hearings.
Three categories of additional information are contained in this
Supplement. The first includes an extended discussion of the Agency's
intent regarding implementation of these proposed standards, and further
elucidation of the basis used by the Agency for assessing the potential
health impact of exposure to ionizing radiation. The second consists of
technical discussions of several areas not covered or addressed only
briefly by the original material. This includes consideration of
multiple reactors on a single site, the nuclear energy center concept,
transuranic effluents resulting from recycled uranium, and nitrogen-16
skyshine doses and control at BWR's. Finally, in two areas, fuel
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reprocessing and milling, considerable additional technical material has
become available concerning control methods since the original docu-
mentation was prepared. Although the proposed standards 'reflected this •
information, the technical documents accompanying the proposal did not.
Surveys based on this new information complete this collection of
additional materials.
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CONTENTS
A. Implementation of and Verification of Compliance with
the Proposed Standards
B. Dose-Effect Assumptions Used as the Basis of the Proposed
Standards
C. Potential Limitations on Multiple Reactor Sites Imposed
by the Proposed Standards
D. An Analysis of Control Options for Nitrogen-16 Off-site
Skyshine Doses at Boiling Water Reactors
E. The Proposed Standards and the Nuclear Energy Center
Concept
F. Control of Krypton and Iodine Discharges from Nuclear
Fuel Reprocessing Facilities
G. Transuranium Effluents from Re-enriching or Refabricating
Reprocessed Uranium
H. Environmental Analysis of the Uranium Fuel Cycle, Part I
(Fuel Supply): Uranium Milling - Revised
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SUPPLEMENT A
IMPLEMENTATION OF AND VERIFICATION OF
COMPLIANCE WITH THE PROPOSED STANDARDS
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IMPLEMENTATION OF AND VERIFICATION OF COMPLIANCE WITH THE
PROPOSED STANDARDS
Introduction
As pointed out in the notice proposing these standards, the primary
responsibility for implementing and assuring compliance with EPA
standards for environmental radiation from nuclear power rests with the
Nuclear Regulatory Commission (NRC) and, in certain cases, "Agreement
States" operating within NRC regulations. Thus, although EPA must
consider the practicality of implementing its standards, it would clearly
be inappropriate for the Agency to specify the detailed procedures to be
followed. On the other hand, it is important that the Agency clearly ,
spell out what it would consider to be an appropriate implementation, as
well as ones which are overly restrictive or inadequate, so as to provide
guidance to the NRC for its development of the detailed regulations (and
modifications of existing regulations) required. The following comments
are intended in the sense of such guidance, as to the the Agency's
intent, therefore, and should not be interpreted as literal dictates of
the regulations required to implement these standards. That
responsibility rests with the NRC, and will have to be worked out by the
NRC through detailed interaction with the affected components of
industry, with timely consultation by NRC with EPA as to the
appropriateness of any proposed implementing regulations, particularly in
the event that difficulties develop.
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A similar situation obtains with respect to verification of
compliance. Enforcement authorities reside in NRC, not EPA. EPA expects
that the NRC will adequately assure compliance, and EPA's own
"compliance" activities will consists principally of the review of the
performance, as reported by NRC, of fuel cycle facilities and of any
variances permitted by NRC, As required, EPA will in the future provide
NRC with guidance on the adequacy of its compliance and variance posture
with respect to these environmental standards.
Operational vs. Pre-Operational Application of_ the Standards
An important consideration relative to these standards is the NRC's
continuing development of design and operating guidance, codified in
10CFR50, which implements the Federal Radiation Guidance that exposures
of the public be maintained as low "as practicable" (25 FR 4402). The
Commission has already issued such guidance for single light-water-cooled
power reactors and has underway similar guidance for fuel reprocessing,
milling, and fuel fabrication facilities. The Agency has determined that
the guidance issued thus far for light-water-cooled reactors provides
adequate assurance of compliance (unless the NRC finds that extreme
extenuating circumstances exist for a specific site) for sites containing
up to at least five such power reactors. Additional guidance may be
required in the future, as noted by the Commission in its opinion filed
with 10 CFR 50, Appendix I, for sites containing larger numbers of
facilities.
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These standards will supercede, for the nuclear power Industry, the
Federal Radiation Guides codified in 10 CFR 20 as limiting concentrations
in air and water at unrestricted locations. Just as the development of
the guidance expressed by Appendix I to 10 CFR 50 took place within the
limitations specified by those standards, the development of future 10
CFR 50 guidance will now take place within the limits specified by these
standards. However, it is not anticipated that the disparity between
standards and guidance will, in general (but not always), be nearly so
great as formerly. For example, at fuel reprocessing sites, a large
portion of the thyroid individual dose standard could be taken up by new
10 CFR 50 guidance (whereas zero dose may be postulated through liquid
pathways due to the absence of any liquid discharges). It is thus not
the intent of the Agency that the standards for dose be "apportioned" to
various operations of the fuel cycle. They apply equally and in full to
doses from any operation or combination of operations in the cycle, and
it is not anticipated that doses from multiple sites will be either
common or significant. In the few instances where overlap of
significance could occur this should be dealt with on a site-specific
basis — not generically through apportionment.
It is particularly important to recognize that the standards apply
only to doses received by individuals and quantities of radioactive
materials released to the environment from operating facilities. This
situation is in contrast to design guidance set forth, for example, by
Appendix I to 10 CFR 50 for light-water-cooled power reactors, which
applies to pre-operational considerations, such as licensing for
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construction of nuclear facilities. While such guidance is useful for
providing the basis for concluding that such facilities can be expected
to conform to standards which apply to actual operations, it is not a
substitute for such standards.
Consideration of the adequacy of control measures at facilities
during pre-operational stages with respect to these standards should be
limited to a finding, either for specific sites, or on a generic basis,
as appropriate, that the facility has provided or has available to it
adequate means to provide reasonable assurance that these standards can
be satisfied during actual operations. Such means may include the
provision of cleanup controls on discharge streams, the ability to
modify, if necessary, its mode of operation to mitigate environmental,
discharges, or methods which interrupt exposure pathways in the
environment. The important point is that the standards specify maximum
doses to real individuals and maximum quantities of certain materials
actually delivered or discharged to the environment, not the specific
design parameters of individual facilities. Thus, for example, it is the
Agency's view that conformance to Appendix I by a planned reactor on a
site containing up to five such facilities (unless extremely unusual
combinations of liquid and air pathways of exposure are actually present
and are expected to be simultaneously intercepted by real individuals)
should constitute jde_ facto demonstration to the NRC that a reasonable
expectation exists that these standards can be satisfied in actual
operation. The Agency will, in the course of its continuing review of
Environmental Statements, identify any situations for which it believes
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that such an expectation has not been adequately justified. A more
detailed exposition of some areas meriting in-depth discussion of the
Agency's view of an adequate demonstration of reasonable expectation of
compliance, such as for adjacent sites, minor releases of specifically
limited radionuclides from fuel cycle facilities, doses from windblown
material originating from mill sites, and transportation-related doses,
is provided below.
Models for Operational Application of_ thes Standards
a) Limits on doses to individuals.
Conformance to the standards should be measured using the most
reasonable and, as required, realistic means available. Thus, in the
case of dose to the thyroid, measurement of the radioiodine content of
milk at the nearest farm, coupled with a determination of the milk
consumption habits of the residents, would constitute a reasonable basis
for a final determination of noncompliance. Conversely, calculations
based on observed releases and meteorology should generally provide the
basis for a routine finding of compliance. Sites failing this test would
merit progressively more detailed study, leading finally to the above-
described (or a comparable) determination of noncompliance (or
compliance).
In the case of potential doses to the whole-body and other organs a
similar sequence of compliance verification methods is available. The
Agency believes that it may be presumed that existing models for
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calculation of exposure fields due to gaseous and liquid releases, using
measured data on quantities released, local meteorology, and stream-flow
characteristics, are adequately conservative to serve as the basis for
verification of compliance with these standards. If reason exists to
believe, based on use of such source term measurements and models, that
noncompliance may exist at a particular site, than more detailed field
measurements may be employed (or, of course, the facility could reduce
its emissions to achieve model-based compliance).
In a very few special situations when two or more sites are in close
proximity, it may be necessary for the regulatory agency to make
allowance for contributions from several sites in order to assure
compliance with the standards at locations intermediate between such
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sites. For sites as close as a few miles from each other overlapping
contributions of as much as 10 to 20% may be possible. The NRC should
make the necessary adjustments in the individual technical specifications
of facilities at such sites to provide reasonable assurance of
compliance. However, in the vast majority of situations the sum of all
reasonably possible contributions from all sources other than the
immediately adjacent site will be small compared to these standards, and
should be ignored in assessing compliance. It would not be reasonable to
attempt to incorporate into compliance assessment doses which are small
fractions of the uncertainties associated with determination of doses
from the primary source of exposure.
A number of potential difficulties exist regarding implementation of
the standards at mill sites. Gamma surveys in the vicinity of some
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existing mill tailings piles show values ranging up to several hundred
mrem/yr in situations where it is logical to assume that these elevated
gamma radiation levels are the result of windblown tailings. Although
the measurement of 25 mrem/yr increments in such dose rates is possible8
rigorous measurement techniques would be required to identify locations
where new depositions of windblown particulates elevate pre-existing
local levels by 25 mrem/yr. Furthermore, because of the projected 20-
year operational lifetime of a typical mill and the assumed additive
impact of new depositions, 1/20 of 25 mrem/yr, or approximately one
mrem/yr, would have to be measured if the standard were to be implemented
by a regulation based on verification on an annual, incremental baais.
This would be unreasonable, since one mrem/yr is small compared to
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uncertainties in natural gamma-ray background levels.
A recent engineering survey report developed for the Nuclear
Regulatory Commission (ORNL-TM-4903, Volume 1) provides an estimation of
the relative ratio of the respirable particles (<10y) to larger particles
(10-80y) blown off the tailings beach of a well-managed tailings
impoundment system. This ratio averages about one and varies from 0.4 to
1.4 depending on specifics of the milling process and other'variables0
It can be estimated, therefore, that one millicurie/yr of insoluble 0-10y
particles removed from a typical pile by wind could deliver a dose
equivalent of approximately one mrem/yr to the lungs of a person living
one kilometer downwind of the pile. At the same time, one millicurie/yr
of 10-80y particles might be deposited in a ring one-half to one
kilometer from a pile, yielding a surface contamination level of about 3
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nCi/m2. This would result in a ganma-ray exposure level of about 10
yrem/yr. After 20 years of operations, each contributing to surface
contamination at such a rate, this exposure might increase to as much as
approximately 0.2 mrem/yr.
Accordingly, the critical exposure pathway for windblown tailings is
most likely to be to the lungs through the direct inhalation of
radioactive tailings; and if this source of exposure is controlled, direct
whole-body gamma exposure from windblown tailings will also be controlled
to a considerably greater degree.
It does not appear at this time to be practical to measure the
annual release of radionuclides from operational tailings piles to the
air pathway. However, it is practical and reasonable to reduce these
releases to very small values (<1 mCi/yr) by application of control
measures that will insure that maximum doses to individuals in the
vicinity of tailings piles are well within the standards. These measures
include back-filling of exposed tailings, keeping tailings under waters,
and spraying any tailings "beaches" that develop with chemical binders to
prevent blowing. In practical terms9 the standards should be implemented
with regard to operational tailings piles by requiring proper and
reasonable dust control measures and by permanent stabilization following
termination of active milling operations,
It should be noted that the standards apply only to annual doses
delivered as the result of discharges of radioactive materials beginning
two years following the promulgation date. They do not apply to doses
resulting from discharges before this date. Decontamination of areas
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contaminated by'windblown tailings from and management of tailings piles
on previously abandoned mill sites are not covered by end are therefore
not required by this standard.
At a fuel reprocessing or a multi-unit reactor site the number of
shipments of radioactive materials per year in and out of the site could
reach several thousand. However, even for this large of number of
shipments, doses to nearby individuals under present Department of
Transportation regulations would not reach one mrem/yrj if they are
located, on the average, more than a few tens of meters from the shipping
route, and if the vehicles involved remain in motion while in the
vicinity of the site. Implementation of the standard does not require,
therefore, modification of existing packaging and shielding requirements.
It probably will be necessary, however, to reguire guaranteed 'non-stop
shipments (a service which is presently obtainable from the
transportation industry) to avoid buildup of doses to bystanders at
habitual stopping places, or to provide restricted access areas for
layovers. It should be noted that the standards do not apply to
transportation personnel while they are engaged in handling shipments;
such exposure is considered to fall in the category of occupational
exposure.
b) Limits on quantities of specific radionuclides released.
Implementation of the nuclide-specific limits on releases of long-
lived materials will require a determination by the NRC of the operating
decontamination factors that must be achieved at locations that are the
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principle potential sources of environmental releases of these materials.
In order to make such a determination it will be necessary to
characterize before 1983, except in the case of transuranlcs, the maximum
average values of environmental releases of these materials from minor
classes of sources to be permitted essentially unrestricted release
(e.g., krypton-85, iodine-129, and transuranic releases from power
reactors or fuel fabrication facilities). Following this, compliance
should consist of verification that the appropriate decontamination
factors are being realized through frequent inplant measurements at the
principle potential sources reported on a routine basis.
Monitoring of the DF's achieved by inplant control systems for the
three types of radionuclides specifically limited by the standards
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appears to be readily achievable using conventional monitoring techniques
and analytical procedures, and such measurements appear to be provided
for at the one facility approaching operational status. Flow-through
ionization chambers are capable of measurements of krypton-85 at
concentratons of less than 1 pCi/cm3, a concentration 1000 times lower
than that corresponding to the standard for a typical stack effluent
volume. Similarly, x-ray spectrometry is capable of sensitivities of the
order of 1 pCi for iodine-129; at 10% of the proposed limit a charcoal
sample of stack effluent would accumulate, for a 10 minute sample of 0.2%
of the stream, 1000 pCi. Finally, gas-flow proportional counters, using
24-hour filter samples (collected on 0.1% of the gas stream) would
exhibit detection limits at least 1000 times smaller than activities
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corresponding to the standard. Periodic confirmation of the isotopic
distribution of transuranics would also be necessary.
It should not be necessary to routinely monitor minor releases of
these materials from minor classes of sources, once these have been
properly characterized as such, unless normal monitoring of general
releases discloses that an unusual situation exists which indicates that
normal "de minimus" releases of these materials may be being exceeded.
Such an occurrence would, presumably, not constitute a "normal" release
and investigation and correction would be warranted in any case.
c) The variance provision.
It is not anticipated that utilization of the variance provision of
the standards is likely to be either required or appropriate for any '
facility other than a power reactor in the foreseeable future. That is
not to say that it would be inappropriate to use the variance provision
if circumstances warranted, but that such circumstances appear unlikely.
On the other hand, it is quite possible that a power emergency, either
local, regional, or national, could occur, and that continued production
of power by a reactor experiencing higher than normal releases would be
in the public interest.
In proposing these standards the Agency purposely did not specify
detailed procedures to be followed to obtain a variance, since these
should be developed by the NRC with opportunity provided for the views of
the interested public and the industry to be heard. The Agency does,
however, have some general views on the implementation of this provision.
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First, the use of the variance should be predicated upon a
demonstrable public need for power, and not on the needs of a utility,
as, for example, the inconvenience of scheduling a repair to a control or
a fuel reloading. Second, the granting of a variance should be publicly
announced, with prior notification of the Agency, and include a brief
preliminary assessment of the extent of the excess exposure and releases
anticipated, the anticipated duration of the variance, the reason for the
excess release, and the reason for granting the variance. Finally, after
the variance has terminated, a final assessment of each of the above
factors should be issued promptly.
In general it is anticipated, based upon past experiences, that when
a facility is approaching a condition in which excess releases are
possible that normal monitoring and reporting of facility releases wilj.
provide more than adequate forewarning to permit timely consideration by
NRC of the need for a variance. However, in order to provide for quick
response in the case of a sudden power emergency, it may be desirable for
the NRC to establish some basic criteria for semi-automatic invocation of
a temporary variance under such circumstances. Such criteria would hava
to be limited, at a minimum, by considerations such as conforraance with
NRC's safety requirements and FRC occupational exposure limits,
limitations which are not affected by these standards.
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Implementing^ Regulations
A number of regulations or regulatory actions are affected by these
standards, as the above discussion of implementation indicates. These
include:
1) 10CFR20 - Modify, to reflect, by reference, that 40CFR190
supercedes for normal releases from uranium fuel cycle operations.
2) 10CFR50, Appendix I - Modify to indicate that additional
requirements may be required for sites containing more than five light-
water-cooled reactors, or, if the NRC so determines, in other special
cases.
3) Review license conditions for fuel cycle facilities, other
than light-water-cooled reactors conforming to Appendix I, for
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conformance to 40CFR190.
4) Determine whether any sites exist which are close enough to
other sites to receive substantial contributions to dose from such sites,
and-make any necessary modifications of technical specifications in such
cases (the Point Beach and Kewaunee sites appear to be the .only such
potential case presently in existence).
5) Determine the apportionment to be made for unrestricted
release (relative to 40CFR190) of krypton-85, iodine-129, and alpha-
emitting transuranics of half-life greater than one year at fuel cycle
facilities not major sources of emissions of these nuclides, and
determine the decontamination factors required at major sources.
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6) Establish criteria, as required, for granting of variances
under power emergency conditions, and for establishing public need for
orderly delivery of electrical power.
7) Establish, where necessary, requirements on transportation of
nuclear wastes and spent fuel to prevent layovers in areas to which
public access is possible.
Several regulatory activities already required by existing NRC
regulations or underway are also relevant to implementation of these
standards. These include:
8) Continuing development of ALAP guidance for fuel cycle
activities other than light-water-cooled reactors.
9) Definition of regulatory models for doses to individuals near
fuel cycle operations. »
10) Definition of "temporary and unusual operating conditions" for
implementation of limiting conditions for operation under Appendix I to
10 CFR 50.
The most significant efforts required, of these that are not already
required or committed, are items 3), 5), 6), and 7). These concern
directly the implementation of the standards, the balance are either
minor codifications of the standards into existing regulations, or
represent reflection of the existence of these standards into existing
ongoing efforts.
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EPA Verification of Compliance
The Agency will assess compliance with these standards through its
review of NRC implementing regulations, review of operating data supplied
to the NRC by licensees, and review of any variances issued by NRC.
Supporting activities will include the Agency's continuing review of
draft and final environmental statements for all fuel cycle facilities,
field studies at selected fuel cycle facilities, and assistance to the
NRC, when necessary, through field measurements in cases of possible
noncompliance.
Under general NEPA and FRC authorities, the Agency routinely reviews
and comments on all NRC regulations, including 10 CFR 50 guidance and
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regulatory guides, pertaining to environmental releases and exposures of
the public due to nuclear fuel cycle operations. In the future, this*
review will also include consideration of the implementation of these
standards. This review will encompass, among others, the appropriateness
of design basis assumptions, environmental transport models, dose
conversion assumptions, environmental monitoring and reporting
requirements, and, finally, operating compliance requirements. The
Agency will not, however, routinely review technical specifications or
other license requirements pertaining to individual licensees.
The Agency also maintains a continuing review of the state of the
environment with respect to contamination by radionuclldes and doses to
the public, including contributions from fuel cycle sources. Beginning
this year, the results of this review will be published annually. This
report will depend, for fuel cycle sources, primarily upon data collected
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by the NRC. The Agency has requested that the NRC supply this
information in sufficient detail to permit reasonably detailed annual
assessments of the exposures of members of the public and releases to the
environment at fuel cycle facilities. Unfortunately, it will apparently
be some time before data for all fuel cycle facilities can be made
available in a suitable form.
EPA's review of draft and final impact statements for individual
fuel cycle facilities will serve to allow EPA to identify to NRC
situations in which it believes future compliance, when the facility is
completed, may be questionable. However, such findings will remain
advisory, as in the past, since responsibility for compliance with these
standards during actual operations rests with the facility and the NRC.
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EPA has for some years conducted special field studies in order to
characterize the environmental releases, transport, and impact of
radlonuclides from fuel cycle facilities. These have included detailed
general studies at pressurized and boiling water reactors, a fuel
reprocessing facility, and at mill tailings piles. In addition,
specialized studies of iodine pathways and of nitrogen-16 radiation at
reactors have recently been carried out. These studies will continue in
the future. They are of invaluable assistance in providing soundly based
knowledge for assessing the behavior of environmental releases of
radioactive materials, and in judging the adequacy of environmental
models used for assessing compliance. The measurement capabilities
developed for these studies may also prove useful and will be available
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for situations In which the NRC needs assistance in field verification of
compliance.
Timing of Implementation of the Standards
It is proposed that these standards become effective' two years from"
the date of promulgation, with the exception of those for krypton-85 and
iodine-129, which are proposed to become effective in 1983.
All existing reactors are now or will shortly be in compliance. In
any case, it is considered reasonable to expect that any reactor
facilities not now in compliance with Appendix I will be by 1978, three
years after its issuance and the earliest possible implementation date
for these standards. The question of timing of implementation of the
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standards is not significant, therefore, as it applies to reactors. '
Only one fuel reprocessing facility is now likely to become operable
by 1978, and, on the basis of its environmental statement and EPA's
assessment of its projected control capabilities, this facility should be
able to achieve compliance with the standards at that time. Future
compliance with requirements for krypton and iodine releases will depend
on the installation of additional controls by 1983. In this regard, it
should be noted that the effective date of 1983 for this portion of the
standard applies to any release of these nuclides after that date, not to
nuclides produced in fuel irradiated after 1983.
Implementation of these standards at milling facilities will in many
cases require the installation of updated dust collection equipment, and
institution of dust control methods at tailings piles. This equipment is
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commonly available in commerce. The standards do not apply retroactively
to offsite windblown tailings, nor to tailings piles at sites no longer
licensed. In a few instances large instabilized tailings piles may exist
at sites with active licenses. The Agency has these special situations
under study.
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SUPPLEMENT B
DOSE-EFFECT ASSUMPTIONS USED AS THE
BASIS OF THE PROPOSED STANDARDS
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DOSE-EFFECT ASSUMPTIONS USED AS THE BASIS OF THE
PROPOSED ENVIRONMENTAL RADIATION STANDARDS FOR THE URANIUM FUEL CYCLE
Many comments were received concerning the Agency's use of the
linear nonthreshold dose response model for estimating the potential
consequences of doses to populations. While a few commentors believed
this model was insufficiently protective of public health, the majority
of comments questioned the Agency's health effects estimates in the
belief that they were overly conservative. These comments were confined
to estimates of cancer risk; the Agency's use of a linear nonthreshold
model to estimate genetic risks, perhaps the largest class of potential
health effects, was not questioned. The Agency agrees that in certain
cases a linear nonthreshold model could over- or under-estimate somatic
health effects, and has adopted a policy of utilizing other dose-effect
models where clinical data clearly indicate better risk estimates can be
made using other assumptions. For example, the Agency has stated that it
is highly probable that a threshold dose is required for the induction
of skin cancer, and therefore such cancers were excluded from considera-
tion in developing these standards(1).
No specific data was presented by commentors to indicate that any
non-linear dose response model is applicable to exposures from the
uranium fuel cycle. Rather, frequent reference was made to a statement
by the NCRP(2) that extrapolation from the rising portion of dose-
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incidence curves derived from data obtained at high doses and dose rates
cannot be expected to provide realistic estimates of the actual risk of
cancer from low level doses of low LET radiation. The Agency agrees
that dose incidence curves must be interpreted with care, but believes
much human data, such as that discribed in the NAS-BEIR Report (3), is
useful for estimating radiation risks.
Three factors have been identified by the NCRP as influencing the
validity of interpolation between zero dose and effects and existing
data based on the linear nonthreshold hypothesis: dose, dose rate and
the LET (linear energy transfer) of the radiation in question(2). For
high LET radiations, such as alpha particle irradiation due to effluents
from the Uranium Fuel Cycle, NCRP seems to accept the use of linear
nonthreshold hypothesis. In the case of low LET radiation, such as from
effluents emitting beta particles and gamma rays, the Agency accepts the
fact that the epidemological basis for risk estimates is less straight-
forward and indeed discussed the uncertainties in its technical documents
offered in support of these standards(l). The Agency is aware that for
low LET radiations in-vitro cell killing experiments generally show
reduced effects at low dose rates, indicating that repair of cell-
killing damage may be taking place. The case for repair of precar-
cinogenic injury, however, is not nearly as clear-cut. Demonstrations
of decreasing cancer induction at low dose rates have been confined to a
few studies utilizing laboratory animals, most often mice. These
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studies provide conflicting results depending on the species, pattern of
irradiation and even the sex of the animals. The decreases in effects
observed are in any case relatively small; about a factor of 2-5, but
not several orders of magnitude as suggested by some commentors. It is
important to note that the effect of dose rate on radiocarclnogensis in
animals is not likely to provide an adequate predictor for human risk,
since both the life span and the pattern of cancer induction following
irradiation are different in man and animals. Nor is it necessary to
limit consideration of this question to animal data. There is some
cancer incidence data on the effect of dose rate on humans, unfortunately
not cited in NCRP 43, which indicates that low dose rates may be equally
or more carcinogenic, particularly for protracted exposures(3,4). Until
unequivocal contradictory data on radiocarcinogenesis in humans is
available that indicates protracted low dose rate exposures are less
carcinogenic than acute exposure at high dose rates, the Agency
considers allowance for reduced injury due to low dose rates too specu-
lative to be made part of the basis for standards developed to protect
public health. While the Agency does not overrule the possiblity that
such data may become available in the future, it does not believe
sufficient data exists now to warrant a revision in its somatic health
effect estimates based on dose rates.
A separable question from dose rate effects is the question of
interpolation from high doses to low doses. The point is often made
that interpolation from high doses over-estimates risk if made from a
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portion of the dose response curve where the number of cancers is in
proportion to the square of the dose. However, as pointed out in the
Agency's technical documents(1), interpolations from effects observed
following high doses may also under-estimate the number of cancers
induced because cell killing at high doses substantially reduces the
number of cells at risk for radiocarcinogenic transformations.
There is growing evidence, as suggested in NCRP 43(2), that the
Kellerer-Rossi model for initial radiation injury (not radiocarcinogenesis
per se), which predicts a summation of linear and dose squared response,
is useful for interpreting at least some radiation effects data.
However, the available data in support of this model indicate that at
doses less than about 100 rem the linear, not the dose squared, term
dominates the predicted response. Most, but not all, of the health
effect estimates given in the BEIR Report are based on data that include
at least one point for doses less than 100 rems. Therefore, it is not
considered likely that Agency estimates of radiation-induced cancer are
greatly overestimated by the use of BEIR results. In a few cases it is
possible to test for this effect directly by comparing the results of
human experience at high and low doses(4). Such studies show little
difference in effects per rem and may, in fact, indicate an increased
effect at low doses, particularly in cases where the radiation exposure
is protracted over relatively long periods of time. Again the linear
nonthreshold hypothesis cannot be characterized as being overly con-
servative. The Agency recognizes that the interpolation of risk estimates
-------
for humans from high to low doses is uncertain(5)9 but believes this is
a more prudent public health policy than extrapolating laboratory data
on short life span animals to man. None of the comments received
indicated why the latter procedure would be preferable.
A number of comments were received expressing the view that the
Agency had not recognized the NCRP comment cautioning, "...governmental
policy-making agencies of the unreasonableness of interpreting or
assuming 'upper limit' estimates of carcinogenic risks at low radiation
levels, derived by linear extrapolation from data obtained at high doses
and dose rates, as actual risks, and of basing unduly restrictive
policies on such an interpretation or assumption"(2). The Agency agrees
with the NCRP that only reasonable interpolations are warranted, and
believes the proposed Uranium Fuel Cycle Standards are both prudent and
reasonable. If there is any disagreement it is in the Agency's adoption
of the NAS recommendation that linear interpolation be used as a "best"
estimate(3) of risk, and not as an estimate on the "upper limit of
risk," which seems to be the current philosophy of the NCRP. The Agency
has based its health effects estimates on a continuing review of current
scientific information, and it believes these estimatesrepresent the
most reasonable interpretation of the available data. It will, of
course, review new scientific findings as soon as they become available.
Some commentors expressed the view that numerical estimates of
radiation-related risks are of little use if they are not compared with
5
-------
the risk from other environmental pollutants. While the Agency accepts
that such comparisons, including a comparison with "natural background
radiation," may place the radiation risk from man's activities in a
perspective useful to the public, the Agency does not accept such
comparisons as the primary basis for establishing radiation protection
standards, since at best it could only result in equity between
pollutants - not between costs and benefits. Having made an assessment
of potential health risks the Agency believes it is more appropriate to
select appropriate limits by means of a cost-effectiveness of health
risk reduction methodology, rather than via comparative risk assessment.
A number of commentors noted that the reduction of very small risks
even further is either not worthwhile or is not cost-effective. The
Agency agrees that the risk to an individual from certain radioactive
effluents may often be small. However, unless a threshold for radio-
car cinogenesis can be demonstrated, the total risk is not necessarily
small, but depends on the number of persons exposed. In developing the
proposed standards careful consideration was given to the cost-effectiveness
V
of various levels of risk reduction for the entire exposed population,
not just for specified individuals. The standards proposed were chosen
so as to avoid the imposition of any unreasonable costs for control. It
is the Agency's conclusion that all of the costs incurred will be
justified by the concomitant reduction of a potential risk to public
health.
-------
REFERENCES
(1) Environmental Analysis of the Uranium Fuel Cycle - Part III,
Appendix C, U. S. Environmental Protection Agency, EPA-73-003-D,
October 1973.
(2) NCRP Report No. 43, Review of the Current State of Radiation
Protection Philosophy, National Council on Radiation Protection
and Measurements, Washington, D.C., January 1975.
(3) The Effects on Populations of Exposure to Low Levels of Ionizing
Radiation, Division of Medical Sciences, National Academy of
Sciences, National Research Council, November 1972.
(4) Linearity vs. Nonlinearity of Dose Response for Radiation Carcino-
genesis, Brown, J. M., review paper presented at the 20th Annual
Meeting of the Health Physics Society, July 15, 1975.
(5) The Relationship Between Radiation and Effect, Policy Statement,
U. S. Environmental Protection Agency, Office of Radiation
Programs, March 1975.
-------
SUPPLEMENT C
y
POTENTIAL LIMITATIONS ON MULTIPLE REACTOR SITES
IMPOSED BY THE PROPOSED STANDARDS
-------
POTENTIAL LIMITATIONS ON MULTIPLE REACTOR SITES
IMPOSED BY STANDARDS FOR THE URANIUM FUEL CYCLE
Introduction
The number of reactors at a given site could be limited, at least in
principle, by an ambient environmental radiation standard applying to all
activities in the uranium fuel cycle (1,2). In order to examine this
possibility, conclusions developed during the AEC's (now NRC) rulemaking
on as low "as practicable" (ALAP) reactor effluents, AEC and NRC
dosimetric estimates for real sites in environmental statements, the
results of EPA field studies, operating data for reactors, and some
analyses of hypothetical configurations are each examined in turn below.
First, however, we digress for a brief assessment of the number and sizes
of multiple reactor sites to be expected, based on actual commitments by
utilities during the next decade.
Multiple Reactor Site Projections
Originally, nuclear power reactors were constructed as individual
units, each on its own site. As nuclear power became more attractive
economically and technologically, multiple reactors were ordered for
-------
single sites. A recent listing of all reactors in operation, under
construction, or on order (3) reveals that there are only six sites for
which as many as four reactor units are presently committed. These four-
unit sites are:
Site
Alan R. Barton
Hartsville
North Anna
Shearon Harris
Surry
WPPSS
Commercial Operation •
Expected for Last
Location Unit
Verbena, Ala.
Hartsville, Tenn.
Mineral, Va.
Newhill, N.C.
Gravel Neck, Va.
Richland, Wash.
1987
1982
1981
1990
1984
1982
TVA also has plans for four more reactor units at as yet unspecified
locations, which may or may not be built on the same site. Thus, it is
likely to be at least five years before a four unit site could be in
operation. No sites containing more than four reactor units are
presently committed. Considering the lead time of eight years necessary
(from contract award to commercial operation) for a single reactor unit,
it will apparently be at least a decade before a five or six unit site
could become operational.
Considerations from the ALAP Rulemaking
One of the basic questions considered by the NRC in the rulemaking
for as low "as practicable" discharges from light-water-cooled nuclear
power reactor effluents was whether the design objectives of Appendix I
to 10 CFR 50 should apply to each reactor or each site. The original
-------
proposal would have applied the basic dose Halts to entire sites.
However, in the words of the Commission (4),
"We have chosen to express the design objectives on
a per light-water-cooled nuclear power reactor basis
rather than on a site basis, as was originally
proposed. While no site limits are being adopted,
it is expected that the dose commitment from multi
light-water-cooled reactor sites should be less than
the product of the number of reactors proposed for a
site and the per-reactor design-objective guides
because there are economies of scale due to the use
of common radwaste systems for multi-reactor sites
which are capable of reducing exposures."
Later, in a more detailed discussion of this question (A), the Commission
expressed the view:
"We are also of the opinion that it will be at least
several years before sites containing as many as
five light-water-cooled nuclear power plants are
developed. Consequently, we see no way that design-
objective guides set on a per-reactor basis can, in
the near future, result in individual exposures that
are more than 5% of present-day (10 CFR 20)
radiation standards. Indeed, we believe that, with
the required inclusion of all radwaste augments
justified on a cost-benefit basis and with the
realization that several reactors cannot physically
be placed so as to all be a minimum distance from
the maximally exposed individual, the actual doses
received by individuals will be appreciably less
than this small percentage."
Thus, it was the opinion of the Commission that the radiation doses from
multi-reactor sites, containing up to five light-water-cooled nuclear
power reactors, will remain at small percentages of present-day (10 CFR
20) radiation standards, specifically, at less than 25 mrem/yr to the
whole body and 75 mrem/yr to the thyroid.
-------
Results of NEPA Reviews
For the last few years, the AEC and NRC have filed environmental
statements under the provisions of the National Environmental Policy Act;
these environmental statements assess the expected performance
characteristics for projected nuclear facilities, including nuclear power
reactors. Table 1 summarizes the results of these analyses for
radioactive releases from all sites projected to contain three or more
reactors. The table shows that:
(1) For the eleven such sites analyzed, in only one case is a whole
body dose by any pathway greater than 2 mrem/yr projected. The
exception, 12 mrem/yr to a hypothetical individual consuming 18 kilograms
per year of shellfish collected from the reactor discharge canal, is
based upon the assumption that public access to that canal is permitted.
(2) For no site is a maximum dose of more than about 15 mrem/yr to
the thyroid of an infant at the nearest farm necessary if reasonable and
readily available control measures are instituted.
It must be emphasized that the estimated doses in Table 1 have been
calculated using conservative models. Even though the most recent
environmental statements employ models specified by regulatory guides
which are more realistic than those used in the past, these models are
still conservative. Again, in the opinion of the Nuclear Regulatory
Commission on Appendix I to 10 CFR 50 (A):
"It must be understood in discussing the matters of
calculational conservatism and realism that Appendix
I means, implicitly, that any facility that conforms
to the numerical and other conditions thereof is
acceptable without further question with respect to
-------
Table 1. Environmental Impacts of Three and Four-Unit Sites
Site
Four Unit Sites
Hartsville
Alan R, Barton
WPPSS(Hanford)
Iodine
16
2.2
9
ScS
4
4.9
1.8
2.9
10
7
<1
<1
<1
5
FOOTNOTES
(a) Dose equivalent to infant thyroid via oow^nilk-pathway at nearest farm
(b) All are final environmental statements except Barton, Davis-Besse, and
Pilgrim, which are draft statements.
(c) 500 hours unshielded occupancy of boundary per year
(d) The applicant's facility design, as proposed, would result in a dose of
74 mrem/yr, which was not deemed "as low as practicable" by the NRC staff.
Addition of a turbine building ventilation treatment system could reduce
the total dose to about 16 mrem/yr, as indicated in the statement.
(e) Does not include dose equivalents from the Hanford N-Reactorj, which is a
light-water-cooled, graphite moderated reactor.
(f) 98% of the release is from the condenser air ejector and steam generator
blowdown tank vent pathways of Units 1 and 2 and can be eliminated or dras-
tically reduced through simple modifications of existing control equipment.
(g) Assumes public access to cooling water discharge canal and annual consumption
of 18 kg of fish and mollusks raised in discharge.
(h) Monitoring and appropriate operational practices will be required by NRC
to assure that dose levels do not exceed 15 mrem/yr? NRC considers the
calculated dose without such measures (28 mrem/yr) very conservative (i.e.,
the actucal dose will be lower).
(i) Based on augmented system committed by applicant (p.11-40 of EIS)
NR- Not Reported
-------
section 50.34a...The numerical guidelines are, in
this sense, a conservative set of requirements and
are indeed based upon conservative evaluations."
In any event, the results presented in Table 1 indicate that for all
multi-reactor sites for which environmental assessments are available,
the maximum projected dose is less than 5 mrem to the whole body, even.
under the highly unlikely presumption that the maximum whole body doses
for gaseous and liquid effluents add arithmetically. Thyroid doses would
limit the number of such reactors at a given site to no greater extent
than do whole body doses. This conclusion is, of course, in harmony with
that reached by the NRC that sites containing as many as five light-
water-cooled nuclear power reactors would result in individual exposures
that are appreciably less than 25 mrem/yr to the whole body and 75
mrem/yr to the thyroid.
Results from Field Studies
In addition to the estimates of dosimetric impact made using
''realistically conservative" calculational models, the EPA and its
predecessor organizations have conducted detailed surveillance programs
at selected facilities (5-9). These studies have confirmed the accuracy
of reported effluents of noble gases and liquids, but appear to reveal
significantly lower iodine concentrations in milk than projected by
models for the milk pathway currently used for environmental analysis.
Field studies conducted by the EPA at Dresden (Unit 1), Yankee Rowe,
and Haddam Neck (formerly Connecticut Yankee) have shown the following
maximum individual doses to the various organs listed (5-8):
-------
Maximum Individual Dose (mrem/yr)
Organ
Whole body
Thyroid
Bone
GI (LLI)
Dresden
8.0
0.74
0.026
0.008
Yankee
3.0
0.006
0.20
0.26
Haddam Neck
3.8
6.0
3.0
0.4
It should be noted that these values are absolute maximum doses for each
organ; all pathways possibly contributing dose to a particular organ were
summed to arrive at the above totals. These doses thus presume that an
individual could be simultaneously exposed to all pathways of exposure
and that he would receive the maximum possible dose from each pathway.
Thus, these doses are extremely unlikely to have been received by any
real individuals, as was pointed out by the authors of the Dresden and
Yankee studies (8):
"...a farmer near Dresden may eat beef, green
vegetables, and drink milk, but he would not also
eat 100 gms of fish per day that had been caught at
Starved Rock Dam, neither would he consume Peoria
drinking water, nor does he reside in the areas for
which inhalation and external whole-body exposures
were calculated. Consequently, actual radiation
exposures to existing populations in the vicinity of
both nuclear power plants are less than the total
dose rates listed..."
Furthermore, most of the whole body dose listed for the pressurized water
reactors (PWRs), Yankee Rowe and Haddam Neck, result from direct
radiation originating from stored radioactive waste (gaseous and liquid
storage tanks). This exposure may be minimized by simple shielding or
-------
careful placement of these tanks relative to the site boundary.
Virtually all of the thyroid dose and bone dose at Haddam Neck results
from the hypothetical consumption of fish (18 kilograms per year) caught
in the discharge canal. Almost all of the whole body dose listed for
Dresden results from exposure to the gaseous effluent (principally noble
gases) discharged from the stack; boiling water reactors (BWRs) such as
Dresden are presently augmenting (or have already augmented) their noble
gas treatment systems to provide additional dose reduction factors of 8
to 180 beyond those in force at the time the above studies were carried
out (2). The three reactors studied are also of early design. Reactors
going into operation today or in design and construction stages
incorporate considerably more sophisticated radwaste treatment systems
having larger processing capacities, greater cleanup efficiency, and
increased flexibility.
Doses due to gamma radiation (directed and scattered, or "shine")
originating onsite can be significant at BWR sites because of the
circulation of activation-produced nitrogen-16 through the turbines and
associated equipment, particularly the moisture separators„ The EPA
field studies discussed above considered the whole body dose from direct
gamma radiation only for the PWR field studies (Yankee Rowe and Haddam
Neck). Field measurements made by the EPA, ERDA, NRC, and others have
shown that dose rates on the order of 10 mrem/yr (whole body) at 500
meters are possible without supplementary shielding of turbine building
components; these dose rates, however, decrease very rapidly with
distance so as to produce very small population doses (10-13). In
-------
addition, dose rates are very dependent upon the design and layout of the
turbine and its associated equipment. Appropriate design of shielding
and location of turbine components relative to the site boundary can
assure that offsite doses from "turbine shine" are minimized. The siting
of many reactor units at a single site should also result in
significantly smaller offsite doses from turbine "shine," as the
exclusion distance increases with the number of reactor units on a site.
According to a recent study (14), the exclusion distance averaged 460
meters for single unit BWRs and 860 meters for twin-unit BWR sites; for
PWRs, single units sites averaged 750 meters, while twin-unit sites
averaged 900 meters. Since the dose from turbine "shine" falls off very
rapidly with distance, such doses should be significantly reduced for
multi-reactor sites. For example, using the data from the most recent
study (13), the dose rate falls off by/a factor of five as the distance
increases from 460 meters to 860 meters. Therefore, it is to be expected
that dose rates from turbine "shine" at multi-reactor sites will not be
significant compared to those from the single unit sites at which field
studies have taken place.
Studies of iodine pathways and potential thyroid doses have been
conducted jointly by EPA and AEC over the past two years at the Dresden,
Monticello, Oyster Creek, and Quad Cities sites (9). The available
results present a consistent picture of iodine concentrations in milk
less than these projected by models for the milk pathway currently used
for environmental analyses.
-------
Results from Reactor Operation
In addition to conservative environmental dose pathway models,
radionuclide source term models have generally been conservative. For
example, fuel experience for PWRs has been much better than the 0.25%
fuel leakage rate now used as a design basis for calculating
environmental releases. Westinghouse, which has manufactured the great
majority of operating PWRs, reports that fuel integrity has generally
been in the neighborhood of 99.98% (i.e., a fuel leakage rate of 0.02%)
for zircaloy-clad fuel. Exceptions to this high level of fuel integrity
occurred in 1969-1970, when hydriding lowered fuel integrity to the 99.8-
99.9% range, and in 1972, when fuel densification lowered fuel integrity
to the 99.9% range (15). On the other hand BWRs, which have typically
been designed for fuel leakage corresponding to the release of 100,000
uCi/sec of noble gases from the air ejector, after a nominal' 30 minute
delay, exhibit a more variable performance. Figure 1 shows that this
design value had yet to be reached by BWRs operating through 1973;
indeed, most were very much below the design value (16). Recent data,
however, indicate a rising trend of releases from BWRs, and EPA is
maintaining a continuing surveillance of this trend, which may indicate
that the present design basis is too low to provide adequate assurance
that Appendix I design objectives will be satisfied in actual operation.
In general, however, fuel integrity at PWRs and for pre-1974 BWR
performance has been considerably better than predicted by conventional
source term models used in environmental analyses.
10
-------
5.0r-
>-
i
ID
I 1.0
LU
Q
§
Q.
CO
0.5 -
0.2 -
01
I
J I
I
I
30-MINUTE
HOLDUP
BASELINE
0.010.05010.2 05
2 5 10 20 30 40 50 60 70 80 90 95
PERCENT OF SAMPLE NOT EXCEEDING PRODUCTION RATE
98 99
99.8 999
Figure 1 Distribution of noble gas releases in 1971-73 for boiling water reactors that commenced operation after
1968. The solid line is a fitted log normal distribution. (Ref. 15)
-------
A second important consideration with respect to conservatism in
source term models is the fact that, especially for PWRs, effluents are
postulated for inplant pathways which require simultaneous levels of
degradation of several parameters in order to lead to a postulated
release to the environment. For example, effluents from the PWR
secondary system (e.g., steam generator blowdown vent or condenser air-
ejector exhaust) require the simultaneous existence of a "design basis"
assumed fuel leakage and a "design basis" assumed steam generator leakage
rate of primary coolant into the secondary coolant. Since the
probability of each "standard" assumption is generally significantly less
than one, the probability of both occurring at the same time must be
smaller than either of the individual probabilities. Thus, if the annual
probability of having the "design basis" number of fuel failures is five
percent and the probability of having a- "design basis" primary to
secondary leak is twenty percent, the probability of operating a PWR with
"design basis" fuel leakage and primary to secondary leakage is of the
order of one percent. ,In .spite of this, light-water-cooled reactors have
been evaluated as if these "design basis'* conditions occur
simultaneously, for periods of time comparable to a year (17).
Analysis of the Additivity of Doses from Multiple Facilities
Similar considerations apply to the assessment of doses from
multiple facilities on a single site. A variety of site specific factors
exist, including the site size, the relative location of individual
facilities on the site, and economies available through utilization of
12
-------
design incorporating shared control measures, each of which mitigate
against arithmetic additivity of doses to a maximum exposed individual
outside the site boundary. In general, these effects are quite
significant, as is reflected by the low doses projected for those'*sites
which have been subjected to analysis, as, for example,' in the
environmental statements quoted above. Indeed, these sites project lower
doses than many single unit sites. In addition, however, there is
significant operational flexibility available at a multi-unit site not
available to sites containing single or double units. For example, if a
reactor at a four-unit site is experiencing a severe rate of1 fuel
failure, the output of the site could be maintained at- a respectable 75%
of capacity while that reactor is serviced, by operating the remaining
units, at full fuel capacity, a degree of flexibility not available to a
one- or two-unit site without calling upon another portion of .the power
grid to take up the loss of capacity.
In addition to the above considerations, which in actual situations
should generally be overriding, it is, however, also instructive to
consider the following hypothetical example. Assume that all units on a
site are located at exactly the same point, and that each is designed to
no more than conform exactly, using "design basis" assumptions, to the
design objective doses specified by Appendix I (say, 5 mrem whole body
dose via the air pathway) to some common hypothetical worst case
receptorj Assume further, since under Appendix I this dose is to be
\
exceeded only in "temporary" and "unusual" situations (4), that one may
assign some reasonable probability that, on an annual basis, the design
13
-------
objective dose for any single unit will not, in fact, be exceeded. For
example, the 0.25% fuel failure assumption currently used as a design
basis for PWRs is not exceeded, on the basis of current operating ,
history, at least 95% of the time. What then, is the dose that can be
expected to be not exceeded at the same confidence level (95%) for 4, 5',
or 6 such units? That the answer is not 4, 5, or 6 times 5 mrem/yr is
obvious. The exact result is dependent upon the variance of the
operating data, and, to a lesser degree, the shape of the distribution of
the data. A statistical analysis utilizing actual operating data for
PWRs and BWRs yields the following projections (18):
Dose Levels (mrem/yr) that will be Satisfied 95% of the Timet
' 4 Units 5 Units 6 Units
PWR 15 18 21
BWR 16 19 22
tFor single units which each satisfy Appendix I at the 95% confidence
level; each value has an uncertainty of approximately 1 mrem/yr.
Each of these values is significantly lower than that predicted by
an assumption of additivity, even for the extreme case of colocation of
all units, no exercise of operational flexibility, and design for the
maximum release permitted by Appendix I considered here.
Conclusion
On the basis of a) results projected by the AEC and NRG for all
0
multi-unit sites presently committed, b) the flexibility available
through proper selection and utilization of future sites, c) the
14
-------
conservative nature of design dose calculations, as opposed to the
applicability of these standards to exposures actually received, d) the
nonadditivity of design basis dose contributions from single units, and
e) the operational flexibility available to sites with multiple units, it
is concluded that the proposed standards can be readily achieved at all '
presently planned and all properly designed future multi-unit sites of up
to at least five units. It is further noted that in "unusual"
circumstances during which the design objectives specified for light-
water-cooled reactors by Appendix I may be "temporarily" exceeded (4),
that the variance provision of the proposed standards would permit
continued operation in times of necessity. Questions associated with
even larger configurations of units, such as nuclear energy centers, are
addressed separately.
15
-------
REFERENCES
Environmental Radiation Protection for Nuclear Power Operations:
Proposed Standards, Federal Register, Vol. 40, No. 104, pp. 23420-
23425, May 29, 1975. . • -
Draft Environmental Statement: Environmental Radiation Protection
Requirements for Normal Operations of Activities in the Uranium Fuel
Cycle, U.S. Environmental Protection Agency, Office of Radiation
Programs, May 1975.
3. Nuclear News. 18. p. 63, August, 1975.
4. Opinion of the Commission in the Matter of Rulemaking Hearing,
Numerical Guides for Design Objectives and Limiting Conditions for
Operation to Meet the Criterion "As Low as Practicable" for
Radioactive Material in Light-Water-Cooled Nuclear Power Reactor
Effluents, Docket No. RM-50-2, U. S. Nuclear Regulatory Commission,
May 5, 1975. - .. .
5. Kahn, B., R. L. Blanchard, H. E. Kolde, et al., "Radiological
Surveillance Studies at a Pressurized Water Nuclear Power Reactor,"
U. S. Environmental Protection Agency, RD 71-1, August 1971.
6. Kahn, B., R.L. Blanchard, H.L. Krieger, et. al., "Radiological
Surveillance Studies at a Boiling Water Nuclear Power Reactor," U.S.
Environmental Protection Agency, BRH-DER 70-1, March 1970.
7. Kahn, B., R.L. Blanchard, W.L. Brinck, et ale, "Radiological
Surveillance Study at the Haddam Neck PWR Nuclear Power Station,"
U.S. Environmental Protection Agency, EPA-520/3-74-007, December
1974.
8. Blanchard, R.L., and B. Kahn, "Pathways for the Transfer of
Radionuclides from Nuclear Power Reactors through the Environment to
Man," Proceedings of the International Symposium on Radioecology
Applied to the Protection of Man and His Environment,'EUR 4800,
Rome, 7-10 September 1971.
9. Detailed Measurement of Iodine-131 in Air, Vegetation, and Milk
Around Three Operating Reactor Sites, Weirs, B.H., Voilleque, P.E.,
Keller, J.H., Kahn, B., Krieger, H.L., Martin, A., and Phillips,
C.R., (IAEA/SM-180/44), presented at the Symposium on Environmental
Surveillance Around Nuclear Installations, International Atomic
Energy Agency, November 1973; and unpublished data, U.S.
Environmental Protection Agency and U.S. Atomic Energy Commission.
16
-------
10. Lowder, W.M., Raft, P.O., and Goglak, C.V., "Environmental Gamma
Radiation Through Nitrogen-16 Decay in the Turbines of a Large
Boiling Water Reactor," HASL-271, January 1973.
11. Brinck, W., Gross, K., Gels, G., Patridge, J., "Special Field Study
at the Vermont Yankee Nuclear Power Station," Internal Report, U.S.
Environmental Protection Agency, Office of Radiation Programs, 1974.
12. Hairr, L.M., Leclare, Philbin, T.W., Tuday, J.R., "The Evaluation of
Direct Radiation in the Vicinity of Nuclear Power Stations," 18th
Annual Health Physics Meeting, June 17-21, 1973.
13. Phillips, C., Lowder, W., Nelson, C., Windham, S., and Partridge,
J., "Nitrogen-16 Skyshine Survey at a 2400 MW(t) Power Plant," U.S.
Environmental Protection Agency, EPA 520/5-75-018, December 1975.
14. ''Land Use and Nuclear Power Plants, Case Studies of Siting
Problems," U.S. Atomic Energy Commission, WASH-1319, 1974.
15. Kramer, F.W., "PWR Fuel Performance — The Westinghouse View,"
Nuclear Energy Digest, No. 2, 1975, Westinghouse Nuclear Energy
Systems, P.O. Box 355, Pittsburgh, Pennsylvania 15230.
16. Martin, J.A., Jr., Nelson, C.B., and Peterson, H.T., "Trends in
Population Radiation Exposure from Operating Boiling Water Reactor
Gaseous Effluents," CONF-741018, Proceedings of the Eighth Midyear
Topical Symposium of the Health Physics Society, October 1974.
17. Calculation of Releases of Radioactive Materials in Liquid and
Gaseous Effluents from Pressurized Water Reactors, Draft Regulatory
Guide l.BB, Nuclear Regulatory Commission, September 9, 1975.
18. ("A Statistical Analysis of the Projected Performance of Multi-unit
Sites, Based upon Operating Data for Existing FacilitiesJ Office of
Radiation Programs, U.S. Environmental Protection Agency, Technical
Note (in preparation).
17
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SUPPLEMENT D
AN ANALYSIS OF CONTROL OPTIONS FOR NITROGEN-16
OFF-SITE SKYSHINE DOSES AT BOILING WATER REACTORS
-------
AN ANALYSIS OF CONTROL OPTIONS FOR N-16 OFFSITE SKYSHINE DOSES
AT BOILING WATER REACTORS
Introduction
The turbine system at a boiling water reactor (BWR) is a potentially
, !
significant source of radiation due to the presence of nitrogen-16, a
relatively short-lived (^=7.14 sec), high energy (2.75 Mev (1%), 6.13
MeV (69%), and 7.11 MeV (4.9%)) gamma emitter in the steam leaving the
reactor. Nitrogen-16 is produced in the reactor core by neutron
activation of oxygen in water, and, although short-lived, can be present
in the turbine system in significant quantities due to the rapid transit
of steam from the reactor vessel through the turbine system and to the
condenser. The result is a flux of direct and scattered gammas which can
result in high occupational exposure rates in and close to' the turbine
building, as well as potentially significant exposure rates to members of
the public beyond site boundaries near the turbine building.
Sources
Detailed expositions of nitrogen-16 sources are presented in the
safety analysis report for the General Electric standard boiling water
reactor, the BWR/6 (1); and for operating BWR's in a comprehensive report
recently released by General Electric (2). In these reports a nitrogen-
16 activity concentration of 50 yCi/gm of steam at the reactor nozzles is
assumed, based on experimental measurements of contact dose rates on
cross-around pipe sections of operating BWRs. Other analyses (3,4) have
assumed nitrogen-16 activities of up to 100 VCi/gm of steam at the
-------
nozzles; however, this is probably due to the desire for conservatism in
the design of shielding.
In a typical modern boiling water reactor, steam flows directly from
the reactor nozzles through the main steam header to the high pressure
turbine (HPT). Steam extraction is also made from this flow path for
steam to the steam jet air ejector (SJAE), feed water heaters (FWH),
gland seal system, and the moisture separater/reheater units (MSRH).
Steam leaving the HPT is routed through the shell side of the MSRH's,
where it is dewatered and reheated for injection into the low pressure
turbines (LPT). Steam extractions are also made at the HPT, MSHR's, and
in several places along the LPT for the various feedwater heater stages
(usually 6).
Typical delay times to and transit times through these components
are shown in Table 1. At a concentration of 50 yCi/gm of steam, the
nitrogen-16 source term at the nozzles is 100 Ci/sec. Thus, it is
obvious that the potential exists for considerable equilibrium activity
to be present in these turbine system components.
Table 2 lists the calculated inventories for the various turbine
building components. The dose significance of these sources depends on
the shielding (both exterior and self-shielding of components) as well as
the geometry of the component layout. The typical order of the dose
significance by component is a) moisture separater/reheaters, b)
intermediate piping, c) high pressure turbine, and d) all other
components.
-------
Turbine Building Configurations
The configuration in which components are placed in a turbine
building has undergone several changes in recent years. Several
different turbine manufacturers have supplied turbines for BWR reactor
plants and component layout has varied as a function of both turbine
manufacturer and of architect-engineer. Turbines have been supplied by
General Electric, Westinghouse, and Kraftwerk-Union, for example, and
facilities using BWR's have been engineered by a variety of architect-
engineering firms. The major significant system design changes have been
with respect to the placement of moisture separators and reheaters.
Earlier BWR designs had vertically-oriented moisture separaters and
separate reheaters located on the mezzanine level of the turbine building
(below the operating floor) as shown in Figure 1. Considerable
shielding was afforded by the concrete structure of the turbine building
around these components, and, particularly above, by the operating floor.
For a variety of engineering reasons, including Increased efficiency
of turbine operation, reduction in building size, and reduction in time
of construction, recent designs have incorporated horizontally-oriented
combined moisture separaters and reheaters located above the turbine
building operating floor level, as shown in Figure 2. The high
equilibrium nitrogen-16 activity levels in tube and shell side of these
systems, combined with the relative lack of self-shielding, compared to
that of the thick steel shells and massive internals of turbines, result
in these "exposed" MSRH's and their supply and return piping producing a
potentially high gamma flux in comparison with all other components.
-------
A system which can perhaps be considered an example of a "worst
case" is the combination of a General Electric BWR with a Westinghouse
turbine system. In this case the steam piping runs overhead from the top
of the HPT to the top or side of the MSRH. Since there is considerable
nitrogen-16 activity in these pipes, they can provide a significant
additional source of gamma exposure beyond the MSRH's themselves.
Dose Assessment
The gamma flux existing at a point outside a turbine building due to
sources of nitrogen-16 inside is difficult to 'calculate. Gammas may-
arrive at a given point by direct paths, by scattering in shielding and
other components, or from air scattering, as shown in Figure 3. The
shielding geometry is complicated due to the variety of component 'shapes
and locations, and each component also has different self-shielding
factors for the gammas involved.
A variety of types of computer codes have been developed to
calculate the air-scattered contrubution to the gamma exposure field
(see, for example, refs. 2,6,7). The potentially most-accurate of these
are Monte Carlo transport codes. However, these models have not been
verified by EPA, and they are sufficiently complex and expensive to
prohibit performing such analyses on a case-by-case basis. No discussion
of analytical techniques for quantitatively analyzing these exposure
rates based on transport codes was undertaken, although the results of
some calculations performed by industry (5) provide the basis for the
present comparison of several options.
-------
Insight into the relation between various shielding options and
anticipated dose rates can be obtained, however, through an examination
of existing shielding studies in conjunction with field measurement
studies. This examination indicates the principal contributors to and
magnitudes of potential doses and permits an informed, if not detailed,
understanding of what might be required to reduce such doses.
Shielding of Components
Because of the high radiation field resulting from nitrogen-16
activity, existing turbine systems are already well-shielded. This is
not primarily because of consideration of doses beyond site boundaries,
but due to the need to comply with existing occupational exposure limits.
In order to restrict the extent of high radiation areas adjacent to
turbines and to allow more frequent or even uncontrolled access to other
areas in the turbine building, the turbines and MSRH's are heavily
shielded. Usually this shielding consists of a thick concrete "shadow
shield" surrounding the turbine (as much as 4 ft thick), and upward
extension of the turbine building lower side walls (up to 3 ft thick) to
shadow-shield the MSRH's. While such shielding substantially reduces the
direct components of the gamma flux, air-scattered contributions from
i
gammas leaving the unshielded top of the turbines and MSRH's can still
produce considerable exposure rates. Therefore, often as a design
f
option, many recent designs have included concrete shields (up to 20"
thick) over the MSRH's and vertical steel plating running between the
turbines and MSRH's to reduce this air-scattered flux (see Figs. 4,5).
In order to assess the effectiveness of such additional shielding as a
-------
means to reduce site boundary doses we have chosen to analyze a variety
of such shielding options for the turbine building component
configuration shown in Figure 4. The assumption is made that concrete
walls are already in place around the MSRH/turbine area as shown to allow
required access in the remainder of the turbine building area within
applicable limits for occupational exposure. These walls are assumed to
consist of three feet of reinforced concrete; this thickness will provide
an attenuation of approximately 99.7% of the incident gamma flux
(neglecting buildup), leaving only the scattered flux as a potentially
significant contributor to, the off-site dose.
Such a characterization of skyshine as the principle source of
exposure from nitrogen-16 at distances greater than a few hundred meters
from the turbine building is supported by a recent field study performed
at the Cooper Nuclear Station by>EPA and ERDA (8). Cooper station is a
BWR with a Westinghouse turbine and horizontally-oriented moisture
separaters located on the turbine building operating floor. Field
measurements were made by EPA in February, 1975, and by ERDA's Health and
Safety Laboratory in April, 1975. Cooper is a reasonable example of the
"base" case turbine building discussed above, since shielding consists of
side walls only, although in this case these consist of 3 ft of high
density concrete. A.significant finding of the study was that nearly
\
100% of the dose measured was due to air-scattered (skyshine) gammas.
The contribution to dose of the direct flux was negligible.
Referring to Table 3, it can be seen that for the base case the
total net equivalent activity above the turbine operating floor is 34 Ci.
-------
Out of this total, 21 Ci are associated with the moisture
separator/reheater and 10.3 Ci are associated with the' intermediate
piping.
The shielding options considered, calculated doses, and anticipated
costs are presented in Table 4. These have been derived in part from
information provided the Agency by General Electric (5). With these
options and their associated dose rates as a basis, and using Means 1975
Building Construction Cost Data (9), we have made independent cost
estimates for installing the additional shielding required by each of the
options considered. The costs presented do not include any additional
basic building structure which might be required within the turbine
building to support the additional weight of the shielding, because for
most of the cases considered the additional weight involved does not
appear to require any additional support beyond that already available in
the basic structure supporting the turbine and other components. The
costs presented here are appropriate to plants in the design stage, and
would not necessarily apply to retrofit situations.
All cases above the base case include the cost of poured-in-place
reinforced concrete, which is supported by an assembly of steel girders
bridging the MSRH's between the exterior turbine shielding wall and
inside panel wall. The inside panel includes steel columns to provide
additional support for the overhead assembly. The dimensions required
for each of two overhead shields are conservatively estimated to be 140'
long by 35' wide. The inside panel walls are assumed to be 140' long by
25' high. . The concrete for exterior side walls and end walls is assumed
-------
to be already present as the "base case." Costs of materials,
installation, engineering, financing, overhead, and profit, were based on
standard estimating assumptions (10). Details of the estimation
procedure used are available upon request. Table 4 provide a summary of
costs for the various shield options, and Figure 6 displays annual dose
at 500 meters vs. cost of shielding.
Doses are presented for the various shielding options both as
calculated by the industry and as projected from values measured in the
field. The field study resulted in data which indicates that the
calculated doses are high by approximately a factor of two. In addition,
the assumption of 100% occupancy, no additional shielding by offsite
building structures, and annual operation at 100% power are considered to
be unreasonably conservative assumptions for estimating real doses to
individuals at real sites. It is concluded, therefore, that it should be
readily possible to restrict the dose from nitrogen-16 skyshine to a real
individual located at reasonable distances from the center of the turbine
building for realistic occupancy times to less than 2 mrem/yr. These
dose levels should be attainable for no more than approximately $250,000
and even these costs should be incurred only in those few instances where
actual site boundaries are so close to turbine buildings as to create the
possibility of significant offsite exposures from nitrogen-16 sources.
-------
Notes
1. BWR/6 Standard Safety Analysis Report, General Electric Company,
NEDO 10741, Vol. 8.
2. Rogers, D. R., "BWR Turbine Equipment Nitrogen-16 Radiation
Shielding Studies," General Electric Report NEDO-20206
(December 1973).
3. "Radiation Shielding Design and Analysis Report - Nine Mile
Point Nuclear Station Unit 2," Stone & Webster Engineering
Co., RP-6, (January 1974).
4. Preliminary Safety Analysis Report - Newbold Island Nuclear
Station, Public Service Electric and Gas Company of N.J.,
February 1970.
5. Information provided EPA by General Electric and Bechtel
Engineering Staff, (January 1975).
6. Woolsen, W.A., A.E. Profio, D.L. Huffman, "Calculation.of the
Dose at Site Boundaries from Nitrogen-16 Radiation in Plant
Components," JRB 72-507 LJ, JRB Associates, (Dec. 1972).
7. Ward, J.T., Jr., "A Dose Rate Kernel for Air-Scattered
Nitrogen-16 Decay Gamma Rays," Ph.D. Thesis, University of
California, Berkley
8. Phillips, C., Lowder, W., Nelson, C., Windham, and
Partridge, J., Nitrogen-16 Skyshine Survey at a 2400 MW(t)
Power Plant, EPA -520/5-75-018.
9. Godfrey, R.G., Editor, "Building Construction Cost Date 1975,"
33rd Ed., 1974, Robert Snow Means Company, Inc.
10. The following markups were applied to materials and installa-
tions: 25% overhead and profit, 2.5% engineering, 10%
contingency. A short term financing factor of 1.375 was then
applied to the total, representing a 10%/per annum financing
cost over a period of three years.
-------
TABLE 1.
Nlr> CHARACTERISTICS OF A STANDARD BT,JR TUHBINF SYSTUi
Component
Main Stean Line and Header System
a. Reactor Nozzle to Main Steam Header
b. Main Stream Header to HPT
High Pressure Turbine
Low Pressure Turbines
Moisture Separator Shell-Side (Steam)
a. Inlet to Vanes
b. Vanes
c. Vanes to Outlet
Moisture Separator Shell-Side (Liquid)
(Vanes, Drain Trough)
Decay Time
at Inlet
(seconds)
0.00
2.09
3.18
5.86
4.29
4.64
4.73
4.64
Estimated
Mass Inventory
(Ibs)
8. 933x10 3
4. 464x10 3
13. 397x10 3
3. 784x10 2
7. 611x10 2
1. 256x1* 3
3. OOxin 2
2. 119x10 3
3. 675x10 3
4. 059x10 3
"lass Flowrate
(Ib/hr x 10~6)
15.396
14.764
14.743
10.678
13.171
11.460
10.904
1.712
Component
Transit Time
(seconds)
.2.09
1.09
0.0924
0.257
0.343
0.0942
0.700
8.54
-------
TABLE 1 (Continued)
Oomponent
Moisture Separator Drain System
a. Steam
b. Liquid
First Stage Reheat System
a. Supply Pipe - HPT to Tube Inlet
b. Tubes
Second Stage Reheat System
a. Supply Pipe-Main Header to Tube
Inlet
b. Tubes
First Stage Reheat Drain System
Second Stage Reheat Drain System
Decay Time
at Inlet
(seconds)
4.73
13.13
3.27
4.33
2.09
3.73
37.3
37.8
Estimated
Mass Inventory
(Ibs)
2.058xl02
6.424xl03
6.630xl03
2.80xl02
S.BllxlO3
6.091xl03
Mass Flowrate
(Ib/hr x 10"6)
0.555^
1.712
0.7011
0.7011
0.6145
0.6145
0.7011
0.6145
Component
Transit Time
(seconds)
1.06
33.0
1.64
34.0
-------
TABLE 1 (Continued)
Cotrponent
Piping System - HPT to MS/MIR
Piping System - MS/RHR to LPT
a. MS/RHR to CIV
b. CIV
c. CIV to LPT
First Stage FWH and Extraction System
a. Extraction Point 4
b. Extraction Point 5
Second Stage FWH and Extraction System
Third Stage FWK and Extraction System
Fourth Stage FWH and Extraction System
Fifth Stage FWH and Extraction System
(Excluding MS Drain System)
Decay Time
at Inlet
(seconds)
3.27
5.43
5.66
5.75
6.12
6.12
6.12
6.12
6.12
3.18
Estimated
Mass Inventory
(Ibs)
3.717xl03
6.857xl02
2.852xl02
2.812xl02
1. 252x10 3
Mass Flowrate
(Ib/hr x 10~6)
13.171
10.904
10.678
10.678
0.1016
0.6017
0.6301
0.7344
0.4016
0.0126
Component
Transit Time
(seconds)
1.02
0.227
0.0962
0.0948
•
-------
TABLE 1 (Continued)
Ccmponent
Sixth Stage FWH and Extraction System
(Excluding Reheater Drain Systems)
C3ondenser
(Excluding return from FW Turbine)
Hotwell
(Excluding return from FW Heaters, etc.)
SJAE First Stage System
a. Off-Gas
b. Driving Steam Supply Line
c. First Stage Driving Steam
Recombiner System
(Second Stage Air Ejector Driving Steam)
Gland Seal System
a. From HPT
b. From Valve Stem
Feedwater Turbine System
Decay Time
at Inlet
(seconds)
3.27
6.12
"36
-7
2.09
4.33
4.33
3.27
3.18
5.66
Estimated
Mass Inventory Mass Flowrate
(Ibs) (Ib/hr x 10 ~G)
0.857
8.207
0.0016
8.207
0.0016
1. 12x10 * 0.0180
0.0080
0.0100
0.0186
0.0029
0.2259
Component
Transit Time
(seconds)
"30 (liquid)
" 1 (gas)
2.24
-------
Table 2. x)
N16 Inventories For A Standard BWR Turbine System
N-16
Inventory
Component (Curies)
Main Steam Line and Header System ' 263
High Pressure Turbine 6.3
Low Pressure Turbines (1) n.8
Moisture Separator and Reheater Shell-side Steam 53
Moisture Separator Shell-side Liquid 41
!loisture Separator Drain System 56
First Stage Reheat System (2) 33
Second Stage Reheat System (2) 32
First Stage Reheat Drain System (3) 1.4
Second Stage Reheat Drain System (3) 1.1
Intermediate Piping System - HPT to MS/HH 59
Intermediate Piping System - MS/RH to LPT 17
First Stage - FWK & Extraction System (4) 26
Second Stage - FWH & Extraction System (4) 23
Third Stage - FWH & Extraction System (4) . 27
Fourth Stage - FWH & Extraction System (4) 15
Fifth Stage - FWH & Extraction System .6
(Excluding Moisture Separator Drain
System Activity Listed Above).
Sixth Stage - FWH & Extraction System 42
(Excluding First and Second Stage Reheat
Drain System Activities Listed Above)
-------
Table 2 (Continued)
N-16
Inventory
Ocnponent (Curies)
Condenser ' 287 '
(Excluding Residual Activity Returned from
Feedwater Turbine).
Hotwell 18
(Excluding Residual Activity Returned from
Feedwater Heaters and Gland Seal System)
SJAE First Stage System (5) .6
SJAE Off-gas System .4
Gland Seal System (6) 1.0
F.W. Turbine System (6) 8.8
Total 1022.0
Ttotes
(1) 6-Flow machine.
(2) Includes inventory in liquid and steam in reheat tubes and in steam
supply line.
(3) Includes total inventory beyond reheater outlet.
(4) Includes total inventory beyound extraction point. Distribution of this
will depend on equipment arrangement and sizing.
(5) Includes inventory in steam supply line.
(6) Includes total inventory beyond inlet at steam supply line.
-------
Table 3., Turbine equipment typical total and net
16N inventories (Ci) for a 1200 Mfe plant.
TOTAL ABOVE OPERATING FLOOR
COMPONENT
Main Steam Lines
HP Turbine
HPT to MS/R Piping
MS/R
MS/R to LPT Piping
LP Turbines
FW Heaters & Extraction
Condenser
Hotwell
SJAE & Gland Seal
FW Turbine
260
6
60
220
17
10
130
290
18
2
9
GROSS
5
6
2
150
17
10
—
—
—
—
—
NET
EQUIVALENT
1.6
0.3
1.3
21
9
0.5
—
—
* •••
—
—
1022
190
34
-------
Table 4. Summary of Shielding Cost Estimates
Shield
rH
(3
(0
. v
•H
to
c
H
-
6"
6"
1'
1'
-
6"
1'
2'
2'3"
Design
H
o
0
33
CO
*
-
3"
6"
1'
1'
-
6"
1'
1'
2'
i-i
0)
§
0)
c
,0
3
H
-
-
-
_
6"
-
-
-
-
6"
Estimated dose at 500
calculational models
100%
Occupancy
&
Capacity
33
13
8.8
5.9
3.0
37
16
8.6
5.8
2.8
50%
meters (mrem/yr)
, based on:
Estimated cost
of shielding
(k$)
field measurements
100%
Occupancy Occupancy
80%
Capacity
13
5.2
3.5
2.4
1.2
15
6.4 .
3.4
2.3
1.1
&
Capacity
15
5.9
4.0
2.7
1.4
17
7.3
3.9
2.6
1.3
50%
Occupancy
80%
Capacity
6.0
2.4
1.6
1.0
0.5
6.7
2.9
1.6
1.1
0,5
Industry '-*'
1 2
'base' 'base1
720
745
890
255**
915
'base' 'base'
745
895
' 255 990
1,250
EPA
Min.
'base'
96
122
205
258
'base'
122
205
•257
348
Max.
'base'
136 '
169
271
347
'base'
169
271
327
469
* Two estimates were provided, both are shown
** This costCthoughtf to represent an option [inbetween] the final two in this category
Note: First five options for turbine perpendicular to) boundary, second five options for turbine parallel to
boundary.
-------
NM\
ir-n
x
BWR TURBINE BUILDING LAYOUT WITH
MOISTURE SEPARATORS LOCATED
BELOW THE OPERATING FLOOR
*.
J \
MOISTURE
SEPARATOR
\~*
LOW PRESSURE
FEED WATER
HEATERS-
CONDENSER
VT
\ v-
V
"A
MOISTURE
SEPARATOR
FIGURE 1. TYPICAL COMPONENT LAYOUT IN EARLY BWR TURBINE BUILDING DESIGNS.5
-------
N\l\
rr
n
-H
ROOF SLAB LOCATION
WHEN USED
INSIDE PANEL
WHEN USED
MOISTURE
SEPARATOR
// // //
LOW PRESSURE
FEED WATER
HEATERS
FIGURE 2. TYPICAL COMPONENT LAYOUT IN CURRENT BWR TURBINE BUILDING DESIGNS.5
-------
AIR-SCATTERED
BUILDING
SCATTERED
DETECTOR
DIRECT (EQUIPMENT
BE.LOW FLOOR)
FIGURE 3. CONTRIBUTIONS TO DOSE RATE FROM N-16 IN TURBINE BUILDING COMPONENTS.
-------
FIGURE 4. TOP VIEW OF TURBINE COMPONENT LAYOUT SHOWING TYPICAL "ACCESS" SHIELD
DESIGN ALONG WITH VARIOUS SHIELD OPTION.5
-------
HEATER fl 3 -°
4THPT
/HEATER
-MEZZANINE
FLOOR
EL 277'-«"
Figure 5. Transverse sectional view of [line Mile Point 2 nuclear plant turbine building,
showing shielding of moisture separators and turbines.^ '
-------
I/)
s-
o
u
-• (J
6*
z o
O 00
Q
LU<
oo:
T3
r «o
CO OC S-
•-• o
«_l +->
oo _i o
QC t-H (TJ
u-i >.
¥1 o
c
O rO
O Q.
U) =3
O
h- O
< O
to o
O LO
o
—I O)
< E
Z3 3
Z (/>
8.0-
6.0-
4.0-
2.0-
0.0
. MAX
MIN
I
200
I
400
100 200 300
COST OF ADDITIONAL SHIELDING ($ 1000)
500
Figure 6. ANNUAL DOSE AT 500 METERS VS. COST OF SHIELDING
(Turbine parallel to boundary)-
-------
SUPPLEMENT E
THE PROPOSED STANDARDS AND THE
NUCLEAR ENERGY CENTER CONCEPT
-------
EPA PROPOSED STANDARDS FOR THE URANIUM FUEL CYCLE
AND THE NUCLEAR ENERGY CENTER CONCEPT
Introduction
The Federal Register notice proposing these environmental radiation
standards for the uranium fuel cycle pointed out that "...in view of the
need to accumulate operating experience for the new large individual
facilities now under construction and the intent of the Agency to review
these standards at reasonable intervals in the future, it is considered
premature and unnecessary to predicate the standards on any siting
configurations (e.g., nuclear energy centers) postulated for the next
decade and beyond. The Agency will consider changes in these standards
based on such considerations when they are needed and justified by
experience..." (1). The proposed standard does not itself specify
standards for any specific siting configuration, nor is any siting
concept excluded from its applicability. EPA's conmitment is simply to
reconsider the standard when data is available on vhich to base an
evaluation of the nuclear energy center (NEC) concept.
A number of commenters on the proposed standards addressed the NEC
concept in somewhat general terms. They expressed two types of concerns.
The first was expressed by one commenter as follows: ''...however, the
proposed limits may discourage plans for energy parks for the following
decades. Since the (sic) energy parks may well offer reduced overall
radiation and health effects to the general public (at the expense of
-------
slightly higher individual exposures) along with possible cost savings
and safeguards improvements, the long range implications of the standards
on the parks should be explictly addressed..." (2). The second concern
seen is: "By specifically excluding nuclear parks from the standards, EP^
makes utility planning for the design, purchase and construction of
future nuclear power plants difficult" (3). None of the commenters
provide any quantitative information to support their concerns.
Background
Only a few studies of nuclear energy centers have been published.
One, titled "Assessment of Energy Parks vs. Dispersed Electric Power
Generating Facilities," and sponsored by the National Science Foundation
(4), did not treat radioactive effluents in enough detail to indicate
whether the proposed standards could or could be met. That study
referenced "Evaluation of Nuclear Energy Centers" (WASH 1288) on this
matter (5).
!
WASH 1288 provides the most complete treatment of NEC's available to
date, and evaluates two real sites in enough detail to draw some
conclusions (albeit imprecise) prior to the more detailed studies of the
NEC concept now almost completed by NRC. Appendix 1 of WASH 1288
provides a discussion of the Hanford reservation in Richland, Washington
as a potential site, which includes an evaluation of potential
radioactive effluents. The results indicate that 25 reactors and a
reprocessing plant could be sited at Hanford with a radiological impact
-------
which should be significantly less than permitted by the proposed
standards (6).
Appendix 2 of WASH 1288 provides a similar treatment of a site at
River Bend, Louisiana, and also estimates an impact less than that
permitted by the proposed standards (7). It should be noted that VASF
12R8 was written in 1973, and the authors were concerned with meeting the
then proposed Appendix I. Thus, effluent controls are assumed in the
discussions that will achieve calculated doses in accordance with
proposed Appendix I.
Appendix 5 of WASH 1288, "Radiological Impact of a Nuclear Center on
the Environment" contains a generic treatment of radioactive effluents by
Soldat. Based on his evaluation, it appears that the proposed standards
for atmospheric releases would be met if prudent site selection is made
and reasonable levels of effluent control provided.
One potential problem indicated by Soldat that would require special
attention is liquid releases. If radionuclides are released from a large
number of reactors into a single body of water, special radwaste or
operating procedures may be necessary, such as onsite receiving ponds.
This would depend on the specific characteristics of the water body for
receiving possibly large quantities of radionuclides (8).
WASH 1288 does not answer all of the concerns expressed by
commenters on the proposed standards. Existing analyses are of a scoping
nature and do not address the advantages and disadvantages of NEC's
versus dispersed siting, nor in any detail the impact of other
considerations (thermal and potential accidents, for example), which
-------
would certainly be appropriate to any decision on standards specifically
designed for NEC's.
The "Nuclear Energy Center Site Survey" (NECSS) now underway by the
NRC is expected to provide much of the data and analysis necessary to
make a sound decision on the viability of the NEC concept. A number of
surrogate sites, as well as hypothetical sites, will be analyzed and
various combinations of reactors and fuel cycle facilities will be
considered. It is EPA's understanding that the NRC staff conducting this
survey has considered these proposed standards and the associated DEIS
during its study of potential NEC's. An examination of the preliminary
results of the NRC study does not reveal any significant conflicts
between the proposed standards for the uranium fuel cycle and the
feasibility of the NEC concept. Such a preliminary finding does not, of
course, preclude a later finding, based on a more detailed study, that
some specific provisions may be required in the standards for such sites.
Piscussion
The task of completely assessing the potential impact of the
proposed standards on NEC's is beyond the scope of this discussion.
However, some of the unique aspects of NEC's that are involved can be
briefly mentioned.
There are some characteristics of NEC's that will make doses to
members of the public less than might be expected on the basis of
assessments for conventional sites. The exclusion distance or the
distance to the nearest boundary from such a large group of plants can be
expected to be greater than for smaller numbers of facilities on
conventional sites. A distance of one to one and one half miles may be
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typical versus the typical one half or less miles for conventional sites.
The sites for NEC's are likely to be quite large (50-75 square miles)
with the plants dispersed over the site in order to minimize effects from
thermal releases to the atmosphere. NEC sites may_also be relatively
remote. Economies of scale and shared systems may also make some
effluent control systems available that would not be cost-effective at
conventional sites.
The dose at the site boundary will not be the multiple of the number
of reactors times the dose from the nearest reactor to the site boundary.
Soldat (8) has calculated that the increase in dose over that due to the
nearest facility (or group) would be a factor of from two to five. A
scoping calculation carried out by EPA for thyroid doses arrives at a
factor of three. Of course this would vary depending on actual site
factors and could increase with the addition of other fuel cycle
facilities, such as fuel reprocessing. However, on a large site one
would expect that such other fuel cycle facilities would be placed well
away from the boundary of the large sites required for NEC's and not
contribute a disproportionate part of the total dose.
Before definitive conclusions can be drawn, all pathways will have
to be considered on a consistent basis; the sensitivity of doses to a
variety of site factors will require evaluation; the effect of adding
fuel cycle facilities must be quantified; quantification of the total
population dose reduction and related benefits achieved by such sites in
relation to any increased maximum individual dose will be necessary; and
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any benefits that could be achieved through shared effluent control
systems will have to be evaluated.
Based on the Information now available, the lack of any other
quantitative input from any source to the contrary, and the expectation
of prudent and sound siting decisions, it appears unlikely that nuclear
energy centers would be unable to meet the proposed standards. However,
EPA will review the entire spectrum of analyses of expected impacts and
benefits that should be provided in part by the NECSS, in part by future
more detailed assessments of specific sites, and in part by experience in
the immediate future with existing facilities, in order to arrive at a
judgment on the appropriateness of these environmental radiation
standards for nuclear power to such possible future siting
configurations. •'
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References
1. Federal Register, 4£, May 29, 1975, p. 23424.
2, Attachment to letter (H. Hollister, ERDA, to R.E. Train, EPA,
September 25, 1975) entitled "Staff comments on proposed EPA
regulation (40 CFR Part 190) 'Environmental Radiation Pro-
tection Standards for Nuclear Power Operations' and accompanying
draft environmental impact statement," p. 6.
3. Letter, W.D. Crawford, Edison Electric Institute, to Director,
Criteria and Standards Division, EPA, July 24, 1975.
4. Assessment of Energy Parks vs. Dispersal Electric Power
Generating Facilities. May 30, 1975, National Science
Foundation, NSF 75-500.
5. Evaluation of Nuclear Energy Centers, January 1974, U.S.
Atomic Energy Commission, WASH 1288.
6. Ibid., Appendix 1, p. 7.24.
7. Ibid., Appendix 2, p. 7.67 et. seq.
8. Ibid., Appendix 5, p. 13.
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SUPPLEMENT F
CONTROL OF KRYPTON AND IODINE DISCHARGES
FROM NUCLEAR FUEL REPROCESSING FACILITIES
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CONTROL OF KRYPTON AND IODINE DISCHARGES
FROM NUCLEAR FUEL REPROCESSING FACILITIES
I. Introduction
Radioactive krypton-85 and iodine-129 discharges from reprocessing
facilities have chemical and physical properties which make their
collection and retention technically difficult. Krypton is a chemically
inert gas, and iodine is volatile at normal temperatures and pressures,.
It has been the practice to discharge to the atmosphere all of the
krypton-85 present in spent reactor fuel. Iodine-129 in spent fuel has
been recognized as a potentially significant environmental contaminant,
and efforts have been made in the past to control the discharge of this
species of radioactive iodine. These efforts were only partially
successful, however, and it has become increasingly apparent that
improved control of long-lived radio-iodine discharges from fuel
reprocessing facilities is necessary (1,2). Current estimates of the
costs and control efficiencies of a variety of improved control systems
for iodine-129 and the most important options for control of krypton-85
are reviewed below. The benefits to be gained by reducing the
environmental dose commitments associated with releases of these
materials through installation of such systems are then set forth.
Finally, the level of cost-effectiveness of each of the control options
is determined.
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II. Source Terms for Iodine and Krypton
The quantities of iodine-129, iodine-131, and krypton-85 present in
spent fuel have been previously reported, based on calculations using the
computer code ORIGEN (3). These values, expressed in curies per metric
ton of heavy metal in the fuel, are:
Kr-85: 10,500 Ci/MTHM
1-129: 0.4 Ci/MTHM
1-131: 0.9 Ci/MTHM
for the following fuel parameters, used in this report:
Burnup = 33,000 MWd/MTHM
Average Specific Power = 30 MW/MTHM
Cooling Time = 160 days.
It is assumed that a light-water-cooled power reactor operates at
33% thermal efficiency, producing approximately 33 MTHM of spent fuel
with this burnup for each gigawatt-year of electric power[GW(e)-yr], and
that a typical fuel reprocessing plant has a throughput capacity of 1500
MTHM per year. Such a plant would be capable of processing the spent
fuel from about 45 such reactors each year.
If no iodine or krypton control systems were installed at a 1500 MT
plant, the number of curies discharged annually would be:
Kr-85: 16,000,000 Ci
1-129: 60 Ci
1-131: 1,400 Ci
It is assumed that these contaminants are discharged to the atmosphere,
rather than into liquid pathways, since currently projected plants use
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complete recycle of process liquids and thus no liquid discharges are
planned.
Although the source term for 1-131 could theoretically approach 1400
Ci per year, it is highly unlikely that such quantities will be available
for discharge in actual operations because of its relatively short half-
life (8.08 days). Even if all spent fuel was process at 160 days cooling
time, any delay of iodine-131 in the various inplant processes or off-gas
streams would permit additional decay and reduce the quantity available
for discharge. Other factors that would reduce the quantity of iodine-
131 available for discharge include: a) the existing large backlog of
spent fuel, which indicates there is no need, at least in the foreseeable
future, to process fuel that has been cooled for only 160 days, b)
cooling requirements for spent fuel shipping casks may be such that the
fuel cannot be loaded for shipping from the reactor to the reprocessor
until it has cooled for periods greater than 160 days, and c) for those
reprocessing plants using in-line solidification of high level waste,
cooling periods in the range of a few years may be required to permit
sufficient decay of radioactive ruthenium. Thus, it is considered highly
unlikely that the 1-131 source term at fuel reprocessing will approach
the theoretical maximum value.
III. Control Technologies for Krypton &t_ Reprocessing Plants
Since krypton is a chemically inert noble gas, it follows the
process off-gas stream in the fuel reprocessing plant and will be
discharged to the atmosphere unless specially designed air-cleaning
systems are used to capture it. Standard air-cleaning systems based on
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chemical processes are ineffective in collecting noble gases. Most of
the krypton produced by the fission process in the reactor is released to
the off-gas stream during dissolution of the spent fuel (4,5). A small
fraction is also released during the shearing operation, but this
fraction is also routed to the main off-gas stream. Thus, all of the
krypton-85 present in the spent fuel is collected in one stream, along
with other contaminants, such as oxides of nitrogen, hydrocarbons, and
other radioactive materials.
Two basic systems are in advanced stages of development for the
control of krypton-85; the the cryogenic distillation system and the
selective absorption system. These are discussed in turn, briefly,
below:
1. Cryogenic Distillation
This process is widely used in industry, where it is better
known as the "liquid air" process and is used to condense and separate
the various gaseous components of air. Heat is removed from air in the
gaseous form in a closed system until the boiling points of the various
gaseous components are reached. As the boiling point of each component
is reached, it liquifies and can be separated from the remaining gaseous
components having lower boiling points. Since krypton has a boiling
point of minus 224 °F and the two major constituents of air, nitrogen and
oxygen, have boiling points of minus 322 °F and minus 297 °F,
respectively, liquifaction and separation of the krypton poses no serious
technical problem. Several descriptions of applications of such systems
to nuclear power plants are available (6-13).
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The most serious potential difficulty associated with cryogenic
systems is the possibility of explosions due to a buildup of hydrogen,
A,
acetylene, hydrocarbons, and oxygen (or ozone) in the system (10). This
can be avoided by chemically removing all oxygen before the gas stream ia
introduced into the cryogenic apparatus (6). Thus, in order to use this
process, two additional systems are required: a) a catalytic converter
system to convert oxygen to water and carbon dioxide, followed by, b) a
system for removal of these products. While this entire system has not
yet been reduced to commercial practice through demonstration in an
operating fuel reprocessing plant, on the basis of existing laboratory
and pilot plant experience it appears feasible for such use and is
expected to be available by 1983 (10). In addition to determining that
the explosion potential of the cryogenic systems is effectively removed
by precleaning the gas stream following use of a catalytic converter, a
full assessment of the remote operation and maintenance capabilities of
this system must be completed in the interim.
The cryogenic system itself is expected to exhibit a
decontamination factor (DF) of at least 1000 (6). However, the overall
efficiency for removal of krypton from the plant is expected to be
somewhat lower because of potential leakage through the system during
startup and shutdown operations, maintenance, etc. Therefore, an
effective plant DF of between 10 and 100 has been projected for routine
3
operation of such a system (14).
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2. Selective Absorption
This process was developed at the Oak Ridge National Laboratory
(ORNL), specifically for the control of krypton-85 at fuel reprocessing
plants (15,16). The process is based on preferential dissolution of
noble gases in a fluorocarbon sorbent, such as the refrigerant freon-12.
The off-gas stream is passed through the sorbent in an absorber column at
a relatively low temperature and high pressure. Essentially all of the
krypton and xenon present are dissolved in the sorbent, along with other
components of the gas stream. The other components are then removed in a
fractionating desorption system and, essentially free of krypton and
xenon, recycled to the off-gas stream. The sorbent is then transferred
to a stripper system where a product gas concentrated in krypton and
xenon is evolved and collected. The pure sorbent is then regenerated and
returned to the absorber column.
The selective absorption process has exhibited a decontamination
factor greater than 1000 in tests with nitrogen"o^cides and carbon
dioxide (10). However, further investigations are expected to be
V
accomplished to define the relevant auxiliary systems required for
successful application of the selective absorption process. An effective
DF of between 10 and 100 has been conservatively designated for this
process. The selective absorption system is free from explosion and fire
hazards, and can be operated routinely for sustained periods with low
maintenance requirements. This system has also not been demonstrated at
an operating commercial reprocessing plant. However, it has been offered
commercially for use on the gaseous effluents from nuclear power
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reactors (17). A recent review concluded that additional testing is
required for this system using off-gas containing significant amounts of
contaminants and estimated that the process could be reduced to practice
by 1983 (10).
Estimated costs for installation and operation of these two systems
are listed in Table 1. Cryogenic distillation system costs are based on
industry estimates of equipment costs (18,19) corrected to installed
costs. Estimated costs for the selective absorption system are based on
ORNL estimates of equipment costs (10).
In order to satisfy the proposed standards, storage for 40-70 years
would be required, depending upon the degree of initial decontamination
achieved, in order to insure adequate decay. The management of krypton-
85 following its collection has been addressed by Foster and Pence (20)
and appears to present no serious problems. They reviewed the advantages
and disadvantages of long-term storage of krypton-85 in high pressure
steel cylinders and concluded that this appears to be a practical method
for fission-product noble gas storage. Final storage of krypton-85 could
take place either at the fuel reprocessing facility or at a properly
designed central waste repository.
IV. Control Technologies for Iodine at Reprocessing Plants
The control of iodine at reprocessing plants is a significant
technical challenge (7). During the last few years a number of promising
systems for control of iodine in gaseous waste streams have been
investigated and most are now in various stages of final demonstration
for commercial use. The principal remaining problem, as pointed out in
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the previous EPA report concerning fuel reprocessing (1), is that, until
recently, inadequate attention has been given to the control of iodine in
low-level liquid waste streams. Any iodine present in these liquid
streams, whether from off-gas scrubber solutions or from other sources,
can potentially be discharged to the environment because of its high
volatility. Evaporative processes are used to reduce the volume of these
low-level liquid wastes and to provide for discharge of tritium to the
atmosphere. Such processes will, of course, also drive off any iodine
present for subsequent discharge to the atmosphere, and systems developed
for removal of iodine from gaseous streams are not, in general,
applicable to evaporator discharges because of their high water content.
A simplified schematic of waste streams appropriate to the
discussion of iodine control systems for current designs of reprocessing
plants is shown in Figure 1. Most of the iodine present in spent fuel is
released to the off-gas system during the fuel dissolution and initial
processing steps. The fraction released to the off-gas has been
estimated at no less than 90% (21). The balance is collected in liquid
waste streams. The off-gas system for a specific plant will not
necessarily be designed just as shown in the schematic, since the
detailed design can vary due to the order in which contaminants are
removed. For example, it may be advantageous to remove the oxides of
nitrogen from the dissolver off-gas stream before dilution by process
off-gas inputs.
Table 2 summarizes iodine control system capabilities and costs.
The iodine control system DF's assumed are, for the most part, those used
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In a recent study of effluent controls for fuel reprocessing by
ORNL (10). It should he noted that there are differences between the
estimates of systems performance in the ORNL report and those presented
in testimony at a recent licensing hearing for the Barnwell fuel
reprocessing facility (21). In general, the ORNL analysis predicts
higher DF's for off-gas systems. For example, the DF's shown on Figure 1
yield an effective overall DF of about 100 for 1-129 and about 500 for
1-131. Those presented in the Barnwell hearing (21), in contrast, are
approximately 20 for 1-129 and about 40 for 1-131 for the plant overall.
That testimony assumes that the mercuric nitrate scrubber bottoms are
discharged into the low-level liquid stream. The management of these
scrubber bottoms is the major source of the difference between these
estimates. It will be necessary to retain the bulk of iodine in the
scrubber bottoms in order to achieve effective control of iodine.
The difference in control efficiencies for 1-129 and 1-131 shown in
Table 2 for Ag-Z and macroreticular resins are due primarily to the
differences in half-lives of these radionuclides, as discussed in detail
by Davis (22). This difference is to be expected in any system which
relies upon delay as part or all of its operating .principal". Thus, it is
essential to both isolate and contain long-lived radionuclides to insure
that they will not eventually re-enter a discharge stream.
The chemical form or species is an important characteristic of the
iodine when considering cleaning efficiencies, environmental transport,
and iodine dosimetry. In general, it is believed that iodine evolved
during the dissolution process will be in the elemental form (23).
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However, any iodine discharged to the off-gas system during or following
the separation processes is considered likely to have a large organic
component (24). The relative fractions of iodine evolved from the
dissolution process step and from the various subsequent separation
processes is not known, nor is the organic component of either fraction
(21). Estimates of these fractions vary widely (21,25) and these
differences will probably not be resolved until studies are conducted
during actual operations of a large facility (25). For the purposes of
this analysis it is assumed that 90% of iodine is discharged to the off-
gas system, with the balance going to liquid waste streams (21). The
fraction of the iodine discharged to the atmosphere following all control
systems is assumed to be about 50% organic and 50% elemental. Factors
contributing to an expectation of a significant organic component of the
final discharges are: a) iodine from the low-level liquid pathway has
passed through organic processing steps and thus can be expected to have
a significant organic component, b) iodine in the off-gas stream is
expected to contain a significant organic contribution from separation
processes, and c) most iodine cleaning systems are more efficient in
removing elemental than.organic iodine, and thus selectively allow
passage of organic iodides.
A brief description of each of the iodine control systems considered
follows:
1. Caustic Scrubbers
Caustic scrubbers are widely used in the chemical industry to
remove contaminants from off-gas streams (26). They have been used in
10
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the nuclear industry to control both ruthenium and iodine (27). Tests
have indicated that DF's of 100 and greater for elemental iodine are
attained (27), but DF's are less for organic iodine species. The
fraction of organic iodine in the primary off-gas stream is not known,
but is predicted to be low (21). It has been assumed that the organip
fraction is less than 10% and that caustic scrubbers will, therefore,
operate routinely with a removal efficiency of no less than 90%. Cost
estimates for a caustic scrubber are abstracted from the ORNL work (10).
2. Mercuric Nitrate Scrubbers
Mercuric nitrate-nitric acid scrubbers have been used at the AEC
(now ERDA) reprocessing facilities at Idaho Falls, as well as at a
commercial facility (Nuclear Fuel Services) to control the discharge of
iodine. While this type of scrubber removes both elemental iodine and
organic iodides, tests have indicated that it is also more efficient in
removing iodine in the elemental form (28). Based on the predicted
relative fractions of organic iodides present (21), it is assumed to
remove about 90% of all iodine from the off-gas stream (28,29). Costs
for mercuric nitrate scrubbers are expected to be similar to those for
caustic scrubbers (1).
3. Silver Zeolite Adsorbers
Silver zeolite adsorbers have not been used to treat
reprocessing plant off-gas, but are scheduled to be installed in future
plants. Most of the development work for this system was conducted at
Idaho National Engineering Laboratories Falls (30). Silver nitrate is
impregnated into an alumina-silica matrix and the resulting material is
11
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arranged in a relatively deep bed, since a longer residence time of the
iodine in'the adsorber appears to enhance its efficiency. High removal
efficiencies have been observed for all chemical species of iodine using
this process (30). Although considerably higher values are reported for
small-scale systems, ORNL assigned a DF of 10 for 1-129 and a DF of 100
for 1-131 for a silver zeolite adsorber, pending the-development of
additional data for plant-scale usage (31,32), and these conservative
values have been assumed here. The costs, which are abstracted from
previous EPA work (1), are subject to some uncertainty related to the
loading rate of the system and thus the quantities of silver required.
4. Macroreticular Resins
Adsorption of iodine from both neutral and slightly acidic
solutions on macroreticular resins has been shown to be about 99%
efficient in laboratory studies (33). However, performance of this
system has not been demonstrated in commercial-scale practice and, until
proven under operating conditions, a conservative DF of 10 for 1-129 and
a DF of 100 for 1-131 are assigned. Costs for this system are estimated
to 'be small (10) .
5. Suppression in Evaporator by Mercuric Nitrate
Mercuric nitrate, when added to liquid evaporators, will
suppress the evolution of iodine into the overheads. The Barnwell
Facility includes provision (34) for this method of iodine emissions
control from liquid waste streams. Yarbro has estimated a DF of 2 to 10
ac'ross the final vaporizer for this addition (21). A conservative value
12
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of 2 is assumed for this analysis. Costs are estimated to be similar to
those for a macroreticular resin system.
6. Advanced Systems
Figure 2 displays a simplified schematic of an advanced iodine
control system. The basic principle of this system is to force
essentially all of the iodine into the off-gas system so as to avoid the
difficulty of removing iodine from liquid streams, and then to use highly
efficient systems to remove and retain iodine from the off-gas. In the
schematic this objective is achieved by using an iodine evolution process
at the dissolver to drive the iodine into the off-gas, and the lodox
system to efficiently remove the iodine from the off-gas. The
voloxidation step is primarily used for tritium control. However, a
significant fraction of both the iodine and krypton present in the spent
fuel will also be driven off by this process. After tritium has been
removed from the voloxidation off-gas, this stream is routed to the
dissolver off-gas stream for subsequent krypton and iodine removal.
The lodox process itself effectively scrubs both elemental and
organic iodine from off-gas streams with concentrated (^20M) nitric
acid (23,35). Laboratory-scale studies have indicated that DF's in
excess of 10,000 for methyl iodine have been obtained in tiulti-staged
bubble-cap columns (24). The efficiency with which iodine is scrubbed
from off-gas streams with nitric acid is dependent on the oxidizing power
of the concentrated nitric acid, which converts the volatile iodine
species to the nonvolatile HI308 form. The cost estimates in Table 2 are
abstracted from the ORNL work (10); there is no provision made at this
13
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time for the additional cost of a fractionation/system to permit recovery
of the acid at low concentrations for recycle to the dissolver and lodox
systems.
The Voloxidation process effectively removes such volatile
fission products as iodine and krypton from sheared fuel, by heating the
fuel to about 550 °C in air or oxygen to release these fission products
by thermal evolution or by oxidation (36,37). The process equipment
would consist of: a) a rotary kiln to oxidize the fuel, b) a recombiner
to form tritiated water, and c) a drier to collect the water and separate
it from iodine and krypton which then flow to the lodox equipment
(36,37). Laboratory-scale tests with highly-irradiated sheared fuel show
that up to 75% of the iodine and 45% of the krypton are volatilized. The
costs shown are based on the ORNL work (10).
ORNL is currently conducting development work on these advanced
systems. Cost estimates and projected DF's are abstracted from their
recent summary (10). ORNL has projected that these systems will be
demonstrated and available for installation in new reprocessing plants by
about 1983.
V. Cost Evaluations
Estimated capital costs and annual operating costs for the various
krypton and iodine control systems described are listed in Tables 1 and
2. For those systems for which only equipment cost estimates were
available, a factor of 1.49 times the equipment cost was applied to
estimate total capital cost. This factor includes engineering,
14
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construction, installation, quality assurance, miscellanous, contingency
and some interest costs (10).
The total annual cost listed in Tables 1 and 2 is the sum of the
annual operating costs and the annualized cost of capital. An annual
fixed charge rate of 18% was used to calculate annual fixed charges.
This rate is based on the following assumptions:
Plant (equipment) lifetime 20 years
Capital investiment in bonds 30% , ^ *? 6~-^ 7
Capital investment in equity 70% /.3~-f //.^
Interest rate on bonds 5%
Rate of return on equity (after taxes) 16%
Local property tax rate 3.2%
Annual cost of replacements 0.35%
Annual property insurance rate 0.25%
The annual fixed charge rate was calculated as:
Return on Investment = 12.7%
Sinking Fund Factor at 10% = 1.75%
Miscellaneous = 3.8%
Annual Fixed Charge Rate = 18.25%
This value is lower than that calculated by ORNL (10), which was based on
a series of earlier cost evaluations (38,39). A review of these
evaluations indicates that economic conditions have changed sufficiently
to warrant the use of the revised rate caOculated above (41). In
particular, the debt-equity ratio has significantly increased during the
past decade and equity returns have decreased proportionally thus
15
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producing lower annual fixed charge rates, although interest rates on
bonds may be higher than the rate used above. Further, the investment
tax credit for pollution control equipment allows for increased benefits
to the reprocessing industry which installs the various krypton and
iodine control systems described in Tables 1 and 2 (41). Other tax
advantages that industry receives by investing in such systems are fast
tax writeoffs for depreciation (42), all incurred state and local taxes
(42), some expense-oriented outlays such as insurance (44), and favorable
treatment for adjustments to the capital basis of equipment (45) . These
tax considerations, recent trends toward low rates of return on equity
v(considerably less than 16%) for those industries that have nuclear
reprocessing interests, and their higher leverage investment status
(higher debt-equity ratios) would also tend to reduce the annual fixed
charged rate below the 18% rate calculated above, which ignores these
additional factors. Finally, a discount rate of 10% was used for both
sinking fund and present worth calculations.
VI. Doses and Potential Health Impact Attributable tx? Krypton and Iodine^
Discharges from Fuel Reprocessing
Cumulative environmental dose commitements to the whole body, lungs,
and the gonads, as well as estimated potential health effects
attributable to release of krypton-85 from a model 1500 MTHM/yr plant are
given in Table 3 for a variety of levels of control efficiency. Plant
startup in 1983 and a useful lifetime of control equipment of 20 years is
assumed. A simple model for krypton transport which assumes immediate
and uniform dispersion into the world's atmosphere was used to estimate
16
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worldwide doses. Total doses calculated using this simple model agree
with results from a more detailed multicompartment treatment described by
Machta, Ferber, and Hefter (46,47) within a few percent, although the two
models do differ regarding the regional distribution of doses delivered
immediately following release. Other parameters, such as population
growth and distribution, dosimetry, and dose-effect relationships, were
handled as described in the previous analysis (1).
Partial cumulative environmental dose commitments to the thyroid and
estimated potential health effects attributable to discharges of iodine-
129 from a model 1500 MTHM/yr plant were calculated using the specific
activity method (1), and are presented in Table 4. These values
represent a partial assessment of the total potential dose and health
impact of iodine-129 in that the period of assessment following release
of this extremely long-lived material (17 million years half-life) is
limited to 100 years. Dose commitments were cumulated for releases over
an assumed control equipment lifetime of 20 years commencing in 1983.
These partial cumulative environmental dose commitments and their
associated health impacts are shown for representative values of overall
plant decontamination factors obtainable using the control methods
described above. The dose-effect assumptions used were derived from more
recent values (47,49) than those used in the original analysis (1): a
population age weighted value of 60 thyroid cancers per million rems to
thyroid was used.
Health effects may also result from exposure of local populations
immediately following release of both iodine-131 and iodine-129, in
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addition to the long-term effects described above. Using methods
described previously (1) and short term pathway parameters noted below,
it is estimated that uncontrolled release of 1400 Ci/yr of 1-131 could
result in 35 health effects and the release of 60 Ci/yr of iodine-129
could result in 15 health effects over a 20-year period of plant
operation commencing in 1983. These values should be added to those
listed in Table 4 to obtain a complete estimate of potential health
effects attributable to uncontrolled release of radioactive iodines for
the first 100 years following release.
In addition to the population doses and impacts calculated above,
maximum potential thyroid doses to individuals may also be significant.
Tables 5 and 6 list calculated maximum individual thyroid doses from
iodine-129 and iodine-131 discharges for a variety of age groups and
release fractions. The values for iodine-131 were calculated using dose
conversion factors previously described (48). Dose conversion factors
for iodine-129 were based upon those used for iodine-131, corrected for
differences in pathway and dosimetry dependent upon half-life and
effective energy of decay products (1). It is assumed that 50% of the
iodine released is in elemental form and 50% is in organic form, and that
X/Q is equal to 5 x 10"8 sec/m3. Although specific sites could vary
significantly from this assumption, it is expected that site selection
criteria for fuel reprocessing facilities will reflect particular
attention to minimization of the possibility of dose to the thyroid of
nearby individuals.
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VII. Cost-effectiveness Considerations
Table 7 displays the estimated cost-effectiveness of risk reduction
of the various options considered for both krypton and iodine control.
The cost-effectiveness of both options for krypton control is high,
compared to that for typical control systems currently in use in the
nuclear power industry, and satisfies the criteria used in judging the
reasonableness of the proposed standards (50).
Analysis of the options available for control of iodine is
complicated by a) the multitude of alternatives available, and b) the
variability of the current stage of development of the different
processes. It is clear that iodine evolution and the iodox cleanup
process represent the most effective improvements over the basic cleanup
of gas streams by scrubbers (with or without backup by Ag-Z) and the
cleanup of liquid waste streams by macroreticular resins characteristic
of current design practice. Unfortunately, reduction to commercial
practice of these systems has not been projected to be completed until
1983. Thus, for the few facilities projected to go into operation prior
to that date, utilization of less efficient (and, in the case of Ag-Z,
less cost-effective) systems will be necessary. However, with the
exception of some secondary systems for liquid cleanup (HgN03 suppression
and, in the case of iodine evolution, macroreticular resin), all of the
options display acceptably high levels of cost-effectiveness. It should
also be noted that although a second scrubber has apparently greater
cost-effectiveness than does Ag-Z, use of the latter system may be
preferable due to its anticipated higher level of performance for removal
of organic iodines.
19
-------
Although Table 7 does not display overall plant decontamination
factors, it can be seen from Tables 2, 5, and 6 that conformance with the
proposed thyroid dose limit of 75 mrem/yr can be readily achieved through
use of a variety of combinations of systems exhibiting DF's of 100 or
more. However, conformance with the proposed limit of 5 mCi/GW(e)-yr or
1.4 kg/yr for iodine-129 (0.225 Ci/yr from a 1500 MTHM facility) by 1983
will require a plant DF of no less than 300. This would be readily
achieved by utilization of iodine evolution followed by the iodox
process. Successful achievement of this level of cleanup without use of
the iodox process will depend to some extent upon future operating
experience with less sophisticated systems. Present estimates of their
performance are quite conservative because of a paucity of operating
experience, especially regarding their performance with iodine-129.
However, it is anticipated and highly probable that DF's greater than 300
for iodine-129 could be achieved by 1983 using appropriate combinations
of scrubbers and Ag-Z, since a variety of options are available for
improving, if necessary, the conservative levels of performance currently
projected. These include a) tandem operation of systems, b) additives,
such as thiosulfate to caustic scrubbers, to improve their efficiency
(51) c) use of iodine evolution to reduce the fraction of iodine in the
liquid waste stream and increase the efficiency of scrubbers by reducing
the organic content of the gas streams, and d) demonstration of more
refficient cleanup of liquid streams than currently assumed.
20
-------
REFERENCES
1. "Environmental Analysis of the Uranium Fuel Cycle, Part III- Nuclear
Fuel Reprocessing," U.S. Environmental Protection Agency, EPA-520/9-
73-003-D, October 1973.
2. Magno, P.J., et. al., "Liquid Waste Effluents from a Nuclear Fuel
Reprocessing Plant," BRH/NERHL 70-2, November 1970.
3. Oak Ridge National Laboratory Plants and Waste Management
Facilities," ORNL-4451, July 1970.
4. Cochran, J.A., et. al., "An Investigation of Airborne Radioactive
Effluent from an Operating Nuclear Fuel Reprocessing Plant,"
BRH/NERHL 70-3, July 1970.
5. Goode, J.H., "Hot Cell Evaluation of the Release of Tritium and
Krypton-85 during Processing of ThO - UO Fuels," ORNL-3956, June
1966.
6. Davis, J.S., and J.R. Martin, "A Cryogenic Approach to Fuel
Reprocessing Gaseous Radwaste Treatment," in "Noble Gases," Stanley,
R.E., and Moghissi, A.A., Editors, U.S. Environmental Protection
Agency, CONF-730915, Las Vegas, September 1973.
7. Schmauch, G.E., "Cryogenic Distillation - An Option for Off-Gas
Treatment," ASME, 74-WA/NE-2.
8. Feibush, A.M., "Cryogenic Distillation, Separation Process for Power
Reactor Gaseous Radwaste," Airco/BOC Cryogenic Plants Corp., Murray
Hill, N.J.
9. Thrall, G.M. and D.F. Pilmer, "A Cryogenic System for Processing
Waste Gas From a PWR Generating Station," 19th Annual Meeting of the
Institute of Environmental Sciences, Anaheim, April 1973.
10. Finney, B.C., et. al., "Correlation of Radioactive Waste Treatment
Costs and the Environmental Impact of Waste Effluents in the Nuclear
Fuel Cycle for Use in Establishing "As Low as Practicable" Guides -
Nuclear Fuel Reprocessing," ORNL-TM-4901, May 1975.
11. Bendixsen, C.K., and F.O. German, "Operation of the ICRP Rare Gas
Recovery Facility at the Idaho Chemical Processing Plant," Idaho
Nuclear Corp., IN-1221, April 19, 1969.
12. Bendixsen, C.L. and F.O. German, "Operation of the ICRP Rare Gas
Recovery Facility During Fiscal Year 1970," Allied Chemical Corp.,
ICRP-1001, October 1971.
21
-------
13. Nichols, J.P., and F.T. Binford, "Status of Noble Gas Removal and
Disposal," ORNL-TM-3515, August 1971.
14. Buckman, James A., "Second Supplement to the Direct Testimony of
James A. Buckman," Barnwell Hearings, AEC Docket No. 50-332.
15. Merriman, J.R. and J.H. P'ashley, "Engineering Development oz an
Absorption Process for the Concentration and Collection of Krypton
and Xenon," Union Carbide K-1770, March 1969.
16. Stephenson, et. al., "Experimental Investigation of the Removal of
Krypton and Xenon from Contaminated Gas Streams by Selective
Absorption in Fluorocarbon Solvents," Union Carbide K-1780, August
1970.
17. Hogg, R.M., "New Radwaste Retention System," Nuclear Engineering
Institute 17. (189), 1972.
18. Personal communication, J.S. Davis, Union Carbide Corp., Linde Div.,
to J.L. Russell, U.S. EPA., March 1974.
19. Personal communication, Dr. A.M. Feibush, Airco Cryoplants Corp., to
J.L. Russell, U.S. EPA, July 1974.
20. Foster, B.A. and D.T. Pence, "An Evaluation of High Pressure Steel
Cylinders for Fission Product Noble Gas Storage," TID-4500, February
1975.
21. Yarbro, O.O., "Supplementary Testimony Regarding the State of
Technology for and Practicality of Control and Retention of Iodine
in a Nuclear Fuel Reprocessing Plant," Barnwell Hearings, AEC Docket
No. 50-332, October 1974.
22. Davis, W., Jr., "Models for Calculating the Effects of Isotopic
Exchange, Radioactive Decay, and of Recycle in Removing Iodine from
Gas and Liquid Streams," ORNL-5060, September 1975.
23. Yarbro, O.O., J.C. Mailen, and W.S. Groenier, "Iodine Scrubbing From
Off-Gas With Concentrated Nitric Acid," 13th AEC Air Cleaning
Conference, 1974.
24. Groenier, W.S., "An Engineering Evaluation of the lodix Process:
Removal of Iodine from Air Using a Nitric Acid Scrubbing in a Packed
Column," ORNL-TM-4125, August 1973.
25. Newman, R.I., "Fourth Supplement to Direct Testimony of Robert I.
Newman," Barnwell Hearings, AEC Docket No. 50-332.
22
-------
26. U.S. Public Health Service, "Air Pollution Engineering Manual," 999-
AP-40, 1967.
27. Staff of the Chemical Technology Division, Aqueous Processing of_
LMFBR Fuels - Technical Assessment and Experimental Program
Definition. ORNL-4436, June 1970.
28. Staff of the Chemical Technology Division, Aqueous Fuel Reprocessing
Quarterly Report for Period Ending June 30^ 1973, ORNL-TM-4301,
August 1973.
29. Staff of the Chemical Technology Division, Aqueous Fuel Reprocessing
Quarterly Report for Period Ending March 31, 1973, ORNL-TM-4240,
June 1973.
30. Pence, D.T., et. al., "Application of Metal Zeolites to Nuclear Fuel
Reprocessing Plant Off-Gas Treatment," ANS Trans. _15, 1, Las Vegas,
1972.
31. Ackley, R.D. and R.J. Davis, "Effect of Extended Exposure to
Simulated LMFBR Fuel Reprocessing Off-Gas on Radioactive Trapping
Performance of Sorbates," ORNL-TM-4529.
32. Allied-Gulf Nuclear Services, Barnwell Nuclear Fuel Plant -
Environmental Report, Docket No. 50-332, November 1971.
33. Unger, W.E., et. al., LMFBR Fuel Cycle Studies Progress Report for
August, November and December 1970, ORNL-TM-3281, ORNL-TM-3127, and
ORNL-TM-3250.
34. Allied-General Nuclear Services, Barnwell Nuclear Fuel Plant Final
Safety Analysis Report. October 1973.
35. Staff of the Chemical Technology Division, Aqueous Fuel Reprocessing
Quarterly Report for Period Ending March 31. 1974, ORNL-TM-4587,
June 1974.
36. Staff of the Chemical Technology Section, "Voloxidation"Removal of
Volatile Fission Products from Spent Fuels," ORNL-TM-3723, January
1973.
37. Staff of the Chemical Technology Division, "Voloxidation-Removal of
Volatile Fission Products from Spent LMFBR Fuels," ORNL-TM-3723,
January 1973.
38. U.S. Atomic Energy Commission, "Reactor Fuel Cycle Costs for Nuclear
Power Evaluation," WASH-1099, 1971.
23
-------
39. Salmon, R., "A Review of Computer Code POWERCO to Include Breakdowns
of Power Cost and Fixed Charge Rates," ORNL-4116, August 1969.
40. Stauffer, C.H., "Position Paper on Tax Relief and Other Federal
Subsidies for Pollution Control Costs," Office of Planning and
Evaluation, U.S. Environmental Protection Agency, April 3, 1973.
41. Complete Internal Revenue Code of 1954 (June 1, 1975 Edition)
Prentice-Hall Inc., Sections 38, 46, 48 (Credit for Investment in
Certain Depreciable Property).
42. Complete Internal Revenue Code of 1954 (June 1, 1975 Edition)
Prentice-Hall Inc., Section 169 (Amortization of Pollution Control
Facilities).
43. Complete Internal Revenue Code of 1954 (June 1, 1975 Edition)
Prentice-Hall Inc., Section 164 (Deductions for Taxes).
44. Complete Internal Revenue Code of 1954 (June 1, 1975 Edition),
Prentice-Hall Inc., Section 162 (Trade or Business Expenses).
45. Complete Internal Revenue Code of 1954 (June 1, 1975 Edition),
Prentice, Hall Inc., Section 1016(la) (Adjustments of Basis).
46. Machta, L., Ferber, G.J., and Heffter, J.L., "Regional and Global
Scale Dispersion of Krypton-85 for Population Dose Calculations," in
Physical Behavior of Radioactive Contaminants in the Atmosphere,
International Atomic Energy Agency, Vienna, 1974.
47. "Krypton-85 in the Atmosphere-Accumulation, Biological Significance,
and Control Technology," National Council on Radiation Protection
and Measurements, Report No. 44, July 1975.
48. "Environmental Analysis of the Uranium Fuel Cycle, Part II - Nuclear
Power Reactors," U.S. Environmental Protection Agency, EPA-520/9-73-
003-C, November 1973.
49. United Nations Scientific Committee on the Effects of Atomic
Radiations, "Ionizing Radiation: Levels and Effects," Vol. II,
United Nations Publication E.72.IX.18, New York, 1972.
50. "Environmental Radiation Protection Requirements for Normal
Operations of Activities in the Uranium Fuel Cycle," Draft
Environmental Statement, U.S. Environmental Protection Agency, May
1975.
51. Cederberg, G.K. and O.K. MacQueen, "Containment of Iodine-131
Released by the RALA Process," IDO-14566, October 1961.
24
-------
n—— T-TST = — — — ri?E — ™
1 IIODINE
1 1
1 1
• I IODINE
1 1 STORAGE
1 1
PRODUCT
'SHFAR — ni'J'jm vr — PROCFSS STEPS —
SYSTEM
(T) CAUSTIC
SCRUBBER
(2) MERCURIC
NITRATE SCRUBBER
AgZ
@ MACRORETICULAR
RESIN
© MERCURIC
NITRATE
SUPPRESSION
(*)
^ MACRORETICULAR
RESIN
®
DF '
LOW LbVbL LIUUIU
10 10o/o * tWVPUKAIOR "
10 OF IODINE
10(1-129) HIGH LEVEL
100(1-131) WASTE OPTION
10(1-129) STORAGE "*~
100(1-131) ,
2. i.. i i i i .ii -
INTERMEDIATE
s LEVEL WASTE
. STORAGE
--•4 STACK
I
t
I
I
I
I
I
I
I
-, I
I I
FINAL
VAPORIZER
-
FIGURE 1. SIMPLIFIED SCHEMATIC OF CURRENT IODINE CONTROL SYSTEMS AT REPROCESSING PLANTS
-------
re)
OFF-GAS 99%
r T °F~~
1 IODINE
1
1
1
I
|
TRITIUM
CONTROL
»
! •
-------
Table 1. Krypton control cost summary.
a)
Annual
Capital Operating
Process DF Cost (M$) Cost (M$)
1.
2.
Cryogenic 10-100 3.4 0.12 "
Distillation
Selective 10-100 3.9 0.40
Adsorption
Present
Worth
Total at
Annual ized 10% and
Cost (M$) 20 years (M$)
0.73 6.2
1.1 9.4
a) All costs are expressed in millions of 1974 dollars,
-------
Table 2. Iodine control cost summary.
Process
1.
2.
3.
4.
5.
6.
A.
B..
Caustic Scrubbing
Mercuric Nitrate
Scrubbing
Silver Zeolite Beds
Adsorption on Macroreticular
Resins
Mercuric Nitrate Suppression
lodox
Vol oxidation5)
Iodine Evolution
DF
10
10
10 (1-129)
10 (1-131)
10 (1-129)
100 (1-131)
2
10,000
4C)
200 C^
Capital
Cost (M$)
0.34
0.31
0.44
0.14
0.14
2.07
2.74
0.75
Annual
Operating
Cost (M$)
0.04
0.12
0.15
0.04
0.04
0.22
0.29
0.08
'Total Present Worth
Annual i zed .@"10%"& 20 years
Cost (M$) (M$)
0.10
0.18
0.13
0.065
0.065
0.59
0.78
0.21
0.82
1.5
2.0
0.56
0.56
5.0
6.6
1.8
a) All costs are expressed in millions of 1974 dollars'.
b) This system is not installed, primarily, to facilitate iodine control, and is listed only for
completeness.
c) These values do not represent actual DF's, but represent a process efficiency factor. .
-------
Table 3. Comulative environmental dose commitment and potential health effects attributable
to Kr-85 discharges from a 1500 MTHM/year reprocessing plant3)
Source Term (Ci/yr)
1.6 x/107
1.6 x 106
1.6 x 105
DF Exposed Organ
1 whole body
lungs
gonds
10 whole body
lungs
gonads
100 • whole body
lungs
gonads
Dose Commitment (person-kilorems)
150
300
82
15
30
82
1.5
3.0
0.82
Health Effects
60
14
25
total 99
6.0
1.4
2.5
total 9T9~
0.60
0.14
0.25
total 0799"
a) Dose commitments are displayed for a plant operating life of 20 years beginning in 1983.
-------
Table 4. 100-year cumulative environmental dose commitment and estimated health effects
attributable to release of 1-129 from a 1500 MTHM/yr-reprocess ing plant. a»b)
Source Term (Ci/yr)
$0
6
1.2
0.6
0.3
.06
DF
1
10
50
100
200
1000
Thyroid Dose Commitment (person-kilorems)
1700
170
34
17
8.6
1.7
Health Effects
100
10
2
1
0.5
0.1
a) Partial environmental dose commitment and health effects are-calculated for 100 years following release
only and for a plant operating life of 20 years.
b) Doses and health effects do not include short term, local impact of either iodine-129 or iodine-131.
These are estimated to be 15 and 35 health effects, respectively, for a DF or 1.
-------
Table 5. Maximum Individual thyroid doses from 1-129 discharged from a 1500 MTHM/year
reprocessing plant .
DF
1
10
50
100
200
1000
a)
Source Term (Ci/yr)
60
6
1.2
0.6
0.3
0.06
6 month old
1100
110
22
11
5.5
1.1
1-129 Thyroid
4 year old
1600-
160
32
16
8
1.6
Dose (mrem/yr)
14 year old
600
60
12
6
3
0.6
b)
adult
140
14
2.8
1.4
0.7
0.14
a) The elemental Iodine fraction is assumed to be 50.
Q
b) Atmospheric dispersion coefficient equals 5 x 10 seconds per cubic meter; only the milk pathway is
considered.
-------
Table 6. Maximum individual doses from 1-131 discharged
from a 1500 MTHM/year reprocessing plant.
a)
DF Source Term (Ci/yr)
6 month old
1-129 Thyroid Dose
4 vear-old
(mrem/yr)
14 year old
adult
1 1400 1900
10 140 190
100 14 19
200 7 9.5
500 2.8 3.8
1000 1.4 1.9
10000 0.14 ' 0.19
2300
230
23
12
4.6
2.3
0.23
430
43
4.3
2.2
0.86
0.43
0.043
110
11
1.1
0.54
0.22
0.11
0.011
a) Fuel cooled for 160 days before processing; the elemental iodine fraction is assumed to be 50%.
b) Atmospheric dispersion coefficient equal 5 x 108 seconds per cubic meter; all pathways.are considered.
-------
Table 7. Cost effectiveness of krypton and iodine control system.
System
1 . Krypton
2. Iodine (for off -gas without
iodine evolution)
3. Iodine (for off -gas with
iodine evolution)
4. Iodine (for liquid streams
without iodine evolution)
5. Iodine (.for liquid streams
with evolution)
Cost
Sncrement
Equipment (M$)
(a) Cryogenic
distillation
(b) Selective
absorption
(a) Scrubber (HgN03 )
(b) lodox (no scrubbers)
(c) Second caustic scrubber
(d) Silver zeolite (one
(one scrubber)
(a) Scrubber a)
(b) lodox (no scrubbers)3'
(c) Second scrubber
(d) Silver zeolite
(one scrubber)
(a) Macroreticular resin
(b) Mercuric nitrate suppression
(a) Macroreticular resin
(b) Mercuric nlitrate suppression
6.2
9,4
1.5
5.0
0.82
2.10
3.3
6.8
0.82
2.06
0.56
0.56
0.56
0.56
Health
Effects
Adverted
89-98
89-98
121
135
12
13
135
150
13
14
14
0.75
. Q.68
. 0.038
Cost/Health
Effect (M$/HE)
0.063-0.070
0.096-0.106
0.012
0.037
0.068
0.154
0.024
0.045
0.063
0.143
0.040
0.75
0.82
14.7
a) Add incremental iodine evolution cost
-------
SUPPLEMENT G
TRANSURANIUM EFFLUENTS FROM RE-ENRICHING
OR REFABRICATING REPROCESSED URANIUM
-------
TRANSURANIUM EFFLUENTS FROM RE-ENRICHING OR REFABRICATING REPROCESSED
URANIUM
Uranium feed material, either to an enrichment plant or to a
fabrication plant, which has been previously used as fuel in a
nuclear power plant may still contain trace amounts of radioactive
impurities after decontamination at fuel reprocessing.
Spent reactor fuel is typically allowed to decay either at the
reactor plant site or at the chemical reprocessing plant site a
minimum decay time of 150 to 180 days. The fuel is then dissolved
in nitric acid and processed by solvent extraction
The UF product from chemical reprocessing will contain small
6
quantities of fission products and transuranium isotopes. Specifications
have been published by the Atomic Energy Commission ' which indicate
the maximum acceptable limits for radioactivity resulting from these
impurities. These are: gross alpha due to transuranium isotopes --
1500 dis/min/(g of U); gross beta due to fission products and
transuranium isotopes -- 10% of the beta activity of aged normal
uranium; and gross gamma due to fission products and transuranium
isotopes -- 20% of the gamma activity of aged normal uranium.
Such processed uranium may then be sent to the enriching plant.
The above maximum acceptable limit for gross alpha radioactivity can
be translated into the following typical distribution (assuming total
(2)
solvent extraction plus conversion decontamination factors for
-------
-2-
3 7 g
neptunium of 10 , plutonium - 10', and transplutonium - 10 ):
2 2
neptunium - 9 X 10 alpha dis/min/(g of U), plutonium - 5 X 10 alpha
2
dis/min/(g of U) and transplutonium - 1 X 10 alpha dis/min/ (g of U).
The actual alpha activity distribution will depend on reactor type,
fuel irradiation history, type•of chemical process, and the additional
conversion and purification operations used in converting uranyl
nitrate hexahydrate to UF., but should not vary significantly from these
6
typical values.
The above beta-gamma radioactivity limits are based on gross
radioactivity measurements related to the background of aged normal
uranium. The beta activity limit is based on direct measurement of the
beta counting ratio, and therefore depends upon the variation of counting
efficiency with energy. The gamma specification is based -on a
comparative measurement using aged natural uranium and a high pressure
ion chamber. A reasonable gamma comparison with natural uranium can
therefore be equated to 20% of the gamma power of aged normal uranium.
The gamma power of aged normal uranium can be calculated to be 269 MeV/
sec/(g of U), which results in a gamma specification of approximately
54 MeV/sec/(g of U).
Typical reactor return material has shown the fission product
gamma radioactivity distribution given in Table 1. Technetium and
uranium beta and uranium and transuranium alpha radioactivity levels
i
found are also indicated.
-------
-3-
TABLE 1
CALCULATED GAMMA RADIOACTIVITY DISTRIBUTION OF FISSION PRODUCTS, GAMMA
AND BETA RADIOACTIVITY OF ALL FISSION PRODUCTS, AND ALPHA RADIOACTIVITY
OF TRANSURANIUM AND URANIUM ISOTOPESa(2)
Isotope
of Gamma
Ru-106
Zr-95-Nb-95
Cs-137
Ce-144
75
22
1
1
Other fission products 1
Tc-99
U-237
c
Transneptunium
Np-237
U-232
U-233
U-234
U-235
U-236
U-238
Typical distribution
based on
gamma specification
(Y MeV/sec/g U)
Radioactivity
(Ci/g U)
Y Radioactivity
40.0
12.0
0.054
0.054
0.054
42.2 X 10
-10
9.3 X 10
^6.9 X 10
-10
^6.9 X 10
-11
-11
^6.9 X 10
-11
Radioactivity
3.16 X 10
2.41 X 10
a Radioactivity
2.43 X 10
4.32 X 10
9.01 X 10
4.70 X 10
7.59 X 10
1.71 X 10
2.88 X 10
3.14 X 10
-8
-6
-10
-10
-9
-11
-7
-8
-7
-7
reactor returns are based on an initial feed of 3.2% U-235,
specific power 30 MW/metric ton uranium, exposure 33,000 MW day/metric
ton, decay 180 days.
These fission products consist principally of Sr, Sb, Sn, and Te.
cPu-238, Pu-239, Pu-240, Pu-241, Pu-242, Am-241, Cm-242, Cm-244
-------
-4-
These radioactivities can be used to determine the annual
inputs and system equilibrium concentrations at an enrichment plant
(Table 2). The technetium-99 beta will contribute the remaining
beta radioactivity and is also included. Plutonium and neptunium
concentrations are based on the above specifications for transuranium
isotopes in the reactor return material.
Gaseous diffusion operating experience, although of almost 30
years duration, has been very limited in terms of large throughputs
of power reactor returns. Although there has been considerable produc-
tion reactor material returned to the cascade, irradiation exposure
of that material has been ten- to twenty-fold less than that for power
(2)
reactors. Experience to date has indicated the following:
1. A significant quantity of all non-uranium radioactivity
(neptunium, plutonium, and fission products) is retained in the
feed cylinder (UFg tank) and will be removed when and where the
returned cylinder is washed.
2. PuF, and NpF, are easily reduced and therefore removed by
trapping with CoF- MgF~, NaF, Cryolite, etc.
3. Fission product removal (except technetium) by these traps may
also be significant. However, good data based on low-level radio-
activity feed materials have not been obtained.
4. Technetium, compared to other fission or alpha emission
products, is less likely to be removed by any process. Experience at
ORGDP* indicates that technetium release to the environment would be
10% of feed to the liquid effluent and 1% of feed to the gaseous
effluent.
*0ak Ridge Gaseous Diffusion Plant
-------
TABLE 2
CALCULATED FISSION PRODUCT AND TRANSURANIUM ISOTOPE*
ANNUAL INPUTS AND EQUILIBRIUM SYSTEM6 CONCENTRATIONS
Annual Input Equilibrium System
Isotope (Ci/year) burden
(Ci)
Ru-106 9.3 13.5
Zr-95-Nb-95 2.0 0.5
Cs-137 0.16 (
0.16 fl-e
0.0266
Ce-144 0.16 0.17
_c
Other fission products 0.16 0.7
Tc-99 (g only) -70.0 70.0Td
Np-237 0.9 -0.9Td
Transneptunium - 0.5 0.5T
aBased on fuel specifications of Table 1.
Not an equilibrium condition since Cs-137 has a 26-year half-life
and true equilibrium would only be approached in 130 years. Therefore,
activity depends on time, T (years of operation).
°Assuming an average effective half-life of 3 years.
"Very long half-life, never reaches equilibrium.
e8.75 MSWU
-------
-6-
5. Experience also indicates that other fission products and
alpha radioactivity release fractions should be no more than one tenth
of that for technetium. Measurements of gaseous and liquid effluents
have failed to identify any other fission products. However release
fractions of 1% to the liquid effluent and 0.1% to the gaseous
effluent for other fission products will be used below to estimate
environmental releases.
6. Cobaltous fluoride traps exhibit decontamination factors of
400 for neptunium and 10 for plutonium prior to feeding to the
cascade or conversion facility. Releases for the system after
trapping can then be proportioned to those exhibited for uranium in
ORGDP release data. Thus, alpha release fractions will be 4 X 10
to the liquid and 2 X 10~ to the gaseous effluents for neptunium
and 1.6 X 10" to the liquid and 8.0 X 10"10 to the gaseous effluents
for plutonium.
7. A large portion of the radioactivity entering a settling pond
will be entrained in the sludge of the pond.
Releases to the environment can occur in three physical states
(gas, liquid, and solid). The bulk of the radioactivity will be
released as solids, either entrained on adsorbate or equipment
removed from service for disposal. Liquid waste will be generated
by rinsing (decontamination) of recycled equipment. The first rinse
solution, which contains the bulk of the radioactivity, are saved to
be used as the dilute acid wash solution. Subsequent rinses are sent
to the primary holding pond.
-------
-7-
Gaseous wastes can result from purge system venting, venting of
evaporator overheads at the uranium recovery facility, and venting of
decontamination hoods in the recycle facility. However, the exact
breakdown for retention and release factors for each step is not known.
One can only make assumptions based on experience with gaseous diffusion.
The limited experience available was used to arrive at the following
estimates (see Table 3) about gaseous, liquid, and solid discharges
for non-uranium radioactivity.^ J
TABLE 3
ASSUMED DISTRIBUTION OF FISSION PRODUCTS AND TRANSURANIUM ISOTOPES
TO ATMOSPHERE, PRIMARY HOLDING POND, AND BURIAL GROUND
Fraction released •
Isotope
Np-237
Other Transuranium
Tc-99
Fission Products
Fraction released
to atmosphere
2 X 10~
8 X ICf10
0.01
0.001
to primary
holding pond
4 X 10"6
1.6 X 10"8
0.10
0.01
Fraction input
to burial ground
VI. 0
M.O
0.89
0.989
-------
-8-
Primary enrichment plant sources of gaseous radioactive wastes
are the product and waste purge systems. Uranium particulates are
removed from these process streams by the high-efficiency-particulate
absolute (HEPA) filter, which has an efficiency greater than 99.95%.
Removal of gaseous uranium is achieved through the use of two chemical
traps in the product and waste withdrawal systems, in series, between
the cold trap and point of discharge into the air.
The first trap contains sodium fluoride that provides for the
adsorption of uranium and certain fission or alpha emitting products.
Through heating and proper valving, the trapped uranium may be
desorbed and subsequently returned to the cascade. The second trap
in the series contains alumina that is used for further removal of
uranium prior to discharbe of the gas stream to the atmosphere. This
trap is nonreversible and uranium recovery is accomplished by leaching
with nitric acid.
The fraction of the feed made up of reactor returns is passed
(2)
through cobaltous fluoride traps prior to being fed into the cascade ;
the traps remove plutonium, neptumium, and a major fraction of the
fission products. These products are removed from the gas stream
by-reduction with Cof^ to the tetraflouride forms that, being particulates,
are entrained within the traps.
Quantification of potential gaseous effluents is difficult because
of uncertainties about the behavior of certain fission products in
feed cylinders, traps, piping, and equipment. In attempting to analyze
-------
-9-
possible releases to the environment, all assumptions, where necessary,
have been made so as to overestimate the magnitude of the source term.
Uranium and technetium releases were estimated by comparison with
operating experience and extrapolated to higher operating levels.
Fission product releases were based on current fission product
specifications, with releases being assumed proportional to that of
technetium, with the exception that a decontamination factor (DF) and/or
retention factor 10 times that for technetium was assumed. This
assumption is very conservative, since current experimental, investigations
(2)
indicate that the actual factor might be as high as 100 to 1000.
Releases of the alpha emitters, neptunium and plutonium, were estimated by
assuming an alpha specification of 1500 dis/min/Cg of U) in reactor returns,
5
with a neptunium DF of 400 and a plutonium DF of 10 through Cop2 traps.
Once fed into the cascade, neptunium and plutonium are assumed to be
released to the environment in the same proportions as uranium.
The estimated constituents of an effluent under the above assumptions
are listed in Table 4.
It may be concluded that recycled uranium which has been re-enriched
will present no particular problem at the fabrication plant because most
of the impurities of higher isotopes have been taken out in the enriching
process, and could not make a significant contribution to an industry
limit of 0.5 mCi/GW(e) for alpha-emitting transuranics of half-life
greater than one year.
-------
-10-
TABLE 4
ESTIMATED RADIOACTIVITY RELEASED TO THE ATMOSPHERE FROM
AN ENRICHMENT PLANTd
(Transuranic alpha specification = 1,500 dis/min/g U)
Isotope Radioactivity
(Ci/year)/Gw(e)
U-232 2.75 X 10"8
-10
U-233 1.5 X 10
U-234 3.25 X 10~5
U-235 1.25 X 10"6
U-236 0.92 X 10"6
U-238 5.3 X 10"6
Transneptunium 3.3 X 10~
c -10 .
Np-237 1.7 X 10
Tc-99 4.5 X 10"4
Ru-106 6.0 X 10~6
Zr-95-Nb-95 1.25 X lo"6
Cs-137 0.92 X 10"7
Ce-144 0.92 X 10".
Other fission products 0.92 X 10
aRelative to Tc-99, the retention of all fission
products in equipment or traps is greater by a factor of 10.
Cobaltous fluoride trap decontamination factor for
Pu-239 = 10 .
cCobaltous fluoride trap decontamination factor for
Np-237 = 400.
d8.75 MSWU Plant
-------
-11-
If, however, recycled material goes directly from reprocessing
to fabrication, cleanup systems will have to be designed and installed
to collect the impurities as the material is converted from UFg to
UCL for blending and/or pelletizing. These systems should have
efficiencies and decontamination factors similar to those described
above for the enrichment plant. They would, therefore, be expected
to also reduce transuranium isotopes in the U0_ to levels resulting in
negligible releases compared to the proposed standard of 0.5 mCi/GW(e).
REFERENCES
(1) 32 FR 16289, November 29, 1967.
(2) "Environmental Statement - Expansion of U.S. Uranium Enrichment
Capacity," U.S. Energy Research and Development Administration,
DRAFT ERDA-1543, June 1975.
-------
SUPPLEMENT H
ENVIRONMENTAL ANALYSIS OF THE URANIUM FUEL CYCLE,
PART I (FUEL SUPPLY): URANIUM MILLING - REVISED
-------
CONTENTS
Page
1.0 Introduction 1
2.0 General Description of the Milling Process 2
3.0 Releases of Radioactive Effluent from Uranium Mills 6
3.1 Airborne Releases 6
3.2 Waterborne Releases 9
4.0 The Model Uranium Mill 15
5.0 Radioactive Effluents from a Model Uranium Mill 17
6.0 Radiological Impact of a Model Mill 20
7.0 Health Effects Impact of a Model Mill 23
8.0 Control Technology for Uranium Milling 24
i
8.1 Airborne Effluent Control Technology 24
8.2 Waterborne Effluent Control Technology and Solid Waste
Control Technology 2&
9.0 Effluent Control Technology for the Model Mill 31
10.0 Retrofitting Control Technology to Operating Uranium Mill... 33
References 34
TABLES
Section 2
2.0-1 Uranium Mills in Operation as of March 1975 3
Section 3
3.1-1 Predicted Airborne Releases of Radioactive Materials from
the Highland Uranium Mill 8
3.2-1 Concentrations of Radioactive Effluents in Waste Liquor
from the Highland Uranium Mill 10
3.2-2 Estimates of Quantities of Radionuclides Seeping Through
the Impoundment Dam of a Uranium Mill Initially and at
2-1/4 Years 12
-------
CONTENTS (CONTINUED)
Page
3.2-3 Analysis of Plant Tailings Effluents from the Humeca
Uranium Mill (Alkaline Leach Process) 14
Section 5
5.0-1 Discharge of Radionuclides to the Air from Model Uranium
Mills and Tailings Piles with Base Case Controls 18
Section 6
_ • t
6.0-1 Radiation Doses to Individuals Due to Inhalation in the
Vicinity of a Model Mill with Base Case Controls 21
6.0-2 Collective Dose to the General Population in the Vicinity
of a Model Mill with Base Case Controls 22
Section 8
8.1-1 Cost and Efficiencies of Control Technology for Mills.... 26
Section 9
9.0-1 Radiological Impact of Airborne Effluents versus Control
Costs for a Model Uranium Mill 32
ii
-------
1.0 Introduction
The EPA recently completed a technical review (1) of the
uranium milling industry as part of an overall analysis of the
uranium fuel cycle (2) (3). This review included a description of
the milling process, estimations of radioactive effluent releases,
radiological impact, health effects impact, and the costs and
effectiveness of control technologies for mills. An analysis of
the tailings piles associated with mills was also included. This
review was prepared in support of EPA's proposed standards for the
nuclear fuel cycle, 40 CFR Part 190 (4_).
Since publication in 1973, considerable new information on the
uranium milling industry has become available (JL»Ji»Z».§)» ^n Particular,
the engineering survey report (j6), "Correlation of Radioactive Waste
Treatment Costs and the Environmental Impact of Waste Effluents in
the Nuclear Fuel Cycle for Use in Establishing 'as Low as Practicable'
Guides - Milling of Uranium Ores," has been prepared by Oak Ridge National
Laboratory for the Nuclear Regulatory Commission (NRC). This report con-
tains an extensive review of the costs and the effectiveness of various
control technology systems for uranium mills and mill tailings piles.
The EPA believes it to be worthwhile to revise its previous
technical review of the milling industry, taking into account these
new sources of information. Because radon-222 releases from fuel
cycle facilities have been specifically excluded from EPA's proposed
standard, analysis of radon-222 releases from uranium mills and
uranium mill tailings piles has been omitted from this document.
Rador-222 will be the subject of separate regulatory actions at a
later date.
1
-------
2.0 General Description of the Milling Process
A uranium mill extracts uranium from ore. The product is a
semi-refined uranium compound (UQ00) called "yellowcake" which is
J O
the feed material for the production of uranium hexafluoride (UFg).
As of March 1975, seventeen mills (_7_) were operating in the United
x_,
States (table 2.0-1) with nominal capacities ranging from 250 to
\
7,000 tons of ore per day. These mills are characteristically
located in arid, isolated regions of the west. Areas with significant
high grade ore reserves are (6): Wyoming, 55 million tons; New Mexico,
50 million tons; Texas, 11 million tons; Colorado - Utah, 6 million
tons; all other areas combined, 7 million tons.
Eighty percent of yellowcake is currently produced by a process
that uses sulfuric acid to leach the uranium out of the ore; the remainder
is produced by a sodium carbonate, alkali leach process. Exact details
vary from mill to mill, but, as an example, the principal steps in an
acid leach process mill are as follows:
a. Ore is blended and crushed to pass through a 2.5 cm (1 inch)
screen. The crushed ore is then wet ground in a rod or ball mill
and is transferred as a slurry to leaching tanks.
b. The ore is contacted with sulfuric acid solution and an
oxidizing reagent to leach uranium from the ore. The product liquor
is pumped to the solvent-extraction circuit while the washed residues
(tailings) are sent to the tailings pond or pile.
c. Solvent extraction or ion exchange is used to purify and
concentrate the uranium.
-------
Table 2.0-1 (7)
URANIUM MILLS IN OPERATION AS OF MARCH 1975
COMPANY
LOCATION
YEAR OPERATIONS
INITIATED
NOMINAL CAPACITY
(Tons of Ore/Day)
Anaconda Company
Atlas Corporation
Conoco & Pioneer
Nuclear, Inc.
Cotter Corporation
Dawn Mining Company
Exxon, U.S.A.
Federal-American
Partners
Kerr-McGee Nuclear
Petrotomics Company
Rio Algom Corp.
Union Carbide Corp.
Union Carbide Corp.
Grants, New Mexico
Moab, Urah
Falls City, Texas
Canon City, Colorado
Ford, Washington
Powder River Basin, Wyoming
Gas Hills, Wyoming
Grants, New Mexico
Shirley Basin, Wyoming
La Sal, Utah
Uravan, Colorado
Natrona County, Wyoming
1953
1956
1961
1958
1957
1971
1959
1958
1962
1972
1950
1960
3000
800-1500
220-1750
150-450
0-400
2000
500-950
3600-7000
525-1500
500
0-1300
1000
-------
Table 2.0-1 (Continued)
COMPANY
LOCATION
YEAR OPERATIONS
INITIATED
NOMINAL CAPACITY
(Tons of Ore/Day)
United Nuclear-
Homestake Partners
Utah International,
Inc.
Utah International,
Inc.
Western Nuclear, Inc.
TVA (Mines Develop-
ment , Inc.)
Grants, New Mexico
Gas Hills, Wyoming
Shirley Basin, Wyoming
Jeffrey City,x^yoming
Edgemont, South Dakota
v.
\
1958
1958
1971
1957
1956
1650-3500
750-1200
1200
400-1200
250-500
-------
d. The uranium is precipitated with ammonia and transferred
as a slurry.
e. Thickening and centifuging are used to separate the
uranium concentrate from residual liquids.
f. The concentrate is dried at 400°F and is sometimes
calcinated at 750 to 950°F.
g. The concentrate or yellowcake is packaged in 208 liter
(55 gallon) drums for shipment.
Large amounts of solid waste tailings remain following the
removal of the uranium from the ore. A typical mill may generate
1,800 metric tons per day of tailings solids slurried in 2,500
metric tons of waste milling solutions. Over the lifetime of the
mill, 100 to 200 acres may permanently be committed to store this
material. These "tailings piles" will have a radiological impact
on the environment through the air pathway by continuous discharge
of radon-222 gas (a daughter of radium-226), through gamma rays given
off by radium-226, radon-222 and daughters as they undergo radioactive
decay, and finally through air and water pathways if radium-226 and
thorium-230 are blown off the pile by wind or are leached from the pile
into surface waters.
-------
3.0 Releases of Radioactive Effluent from Uranium Mills
The radioactivity associated with uranium mill effluents comes
from the natural uranium and its daughter products present in the
ore. During the milling process, the bulk of the natural uranium
is separated and concentrated, while most of the radioactive daughter
products of uranium remain in the uranium-depleted solid residues that
are pumped to the tailings retention system. Liquid and solid wastes
from the milling operation will contain low level concentrations of
these radioactive materials, and airborne radioactive releases include
radon gas and particles of the ore and the product uranium oxide.
External gamma radiation levels associated with uranium milling
processes are low, rarely exceeding a few mrem/hr even at surfaces
of process vessels.
3.1 Airborne Releases
Airborne releases from uranium milling operations include both
particulate matter and gases. Dusts containing uranium and uranium
daughter products (thorium-230 and radium-226) are released from ore
piled outside the mill. Dusts containing uranium and uranium daughter
products are released from the ore crushing and grinding ventilation
system, while a dust containing mostly uranium without daughters is
released from the yellowcake drying and packaging operations. These
dusts are discharged to the atmosphere by means of low stacks.
Because uranium is discharged to the air pathway as ore dust
and as calcinated yellowcake, it will be considered as an insoluble
aerosol. Radium-226 and thorium-230 discharged as ore dust will also
-------
be considered insoluble aerosols.
The air flow through a typical crushing and grinding ventilation
system is about 27,000 cfm; that through the yellowcake drying and
packaging ventilation system is about 6,000 cfm. Because of the
different air flows, dust characteristics, and locations within the.
plant, separate air cleaning equipment systems are usually required;
a mill is therefore usually considered to have two separate airborne
effluent release streams, each with its own control systems, costs,
and source terms.
Radon gas is released from the leach tank vents, ore piles,
tailings retention system, and the ore crushing and grinding ventila-
tion system. There is no practical method presently identifiable
that will prevent the release of radon gas from uranium mills.
As an example, table 3.1-1 gives the estimated maximum release
rates and conservative estimates of site boundary concentrations
considering all potential sources of airborne dust fumes and mists
as predicted for the Highland Uranium Mill in Wyoming (9_,10) . The
capacity of the Highland Mill is about 1,200 MT/yr of yellowcake.
Toward the end of the operating lifetime of a tailings retention
system, some of the tailings will no longer be under water and will
dry out to form a beach (6) . Wind erosion can then carry off tailings
material as airborne particulate matter unless control measures are
taken to prevent such erosion.
-------
00
-Table 3.1-1 (1,10)
Predicted airborne releases of radioactive materials from the Highland Uranium Mill
Release rate Site boundary A a
Radionuclide
Uranium-natural
Thorium-230
(insoluble)
Radium-226 -
(insoluble)
(Ci/yr)
•»
0.1
.06
.06
4
Air concentration
(pCi/m3)
0.003
.001
.001
Site boundary B
Air concentration
(pCi/m3)
0.0004
.0001
.0001
aDistance to site boundary A assumed to be 800 m (2,600 ft) west of mill.
Distance to site boundary B assumed to be 5,200 m (12,700 ft) east of mill.
-------
3.2 Waterborne Releases
The following discussion refers to the best of current procedures
of handling mill liquid wastes,in which these wastes plus tailings
are stored in a tailings retention pond system which uses an
impervious clay-cored earth dam combined with local topographic features
of the area to form an impoundment.
The liquid effluenc from an acid-leach process mill consists
of waste solutions from the leaching, grinding, extraction and washing
circuits of the mill. These solutions, which have an initial pH of
1.5 to 2, contain the unreacted portion of the sulfuric acid used
as the leaching agent in the mill process, sulfates, and some
silica as the primary dissolved solids, along with trace quantities of
soluble metals and organic solvents. This liquid is discharged with
the solids into the tailings pond.
Concentrations of radioactive materials predicted in the 2,500
MT/day of waste liquor from the Highland milling plant are shown in
table 3.2-1 (9_, 10) . Radioactive products of radon decay may also be
present in small concentrations. Since the concentrations of radium-226
and thorium-230 are about an order of magnitude above the specified
limits to 10 CFR 20, considerable effort must be exerted to prevent any
release of this material from the site. The waste liquor is, therefore,
stored in the tailings retention pond which is constructed to prevent
discharge into the surface water system and to minimize percolation
into the ground. This is a continuing potential problem requiring
monitoring programs to insure that there is no significant movement
of contaminated liquids into the environment.
-------
Table 3.2-1
Concentrations of radioactive effluents in
waste liauor from the Highland uranium mill (9.,1_0)
Concentration
Radionuclide (pCi/1)
Uranium-natural 800a
Radium-226 350
Thorium-230 22,000
aAbout 0.001 g/ml.
10
-------
If an earth-fill, clay-cored dam retention system serves as
a collection and storage system for the liquid and solid process
wastes generated in the mill, it will permit the evaporation of most
of the contained waste liquids and serve as a permanent receptacle
for the residual solid tailings. However, after the initial
construction of the retention system, it is to be expected that
there will be some seepage of radionuclides through and around the
dam (9_,_10_) and downward into the soil beneath the impoundment area.
It has been estimated that this seepage will diminish over a period
of about 2 years because of the sealing effect from accumulation
of finer particles between the sandstone grains. On the other hand,
sealing may not occur. Examples of the total quantities of radionuclides
that are estimated to be released through and around the dam are shown
in table 3.2-2. Radium-226 is a radionuclide of concern in this case.
Radium-226 levels as high as 32 pCi/1 (11) have been found in seepage
from current operating mills. Assuming a seepage rate of 300 liters
per minute, the concentration of radium-226 seeping into a stream of
140 liters per second (5 cubic feet per second) is approximately 1 pCi/1
which is 1/5 of EPA's proposed interim Primary Drinking Water Regulation
for radium-226 (12). In the applicant's environmental report for the
Highland Uranium Mill (9_,10) , a seepage concentration of 350 pCi/1
radium-226 was assumed, bringing the concentration of radium in such an
offsite stream up to 12 pCi/1. The Highland Uranium Mill is also esti-
mated to release to the tailings pond 22,000 pCi/1 thorium-230 and trace
quantities of short-lived radon daughter products.
11
-------
Table 3.2-2
Estimates of quantities of radionuclides seeping through the
impoundment dam of a uranium mill initially and at 2-1/4 years (9.,JO_)
Initial seepage Seepage per day(a'
Radionuclide per day after 2-1/4 years
Uranium 350 yCi 35 yCi to 3.5 yCi
Thorium-230 9,600 yCi 960 yCi to 96 yCi
Radium-226 150 yCi 15 yCi to 1.5 yCi
U)$eepage assumed to be inhibited due to seal ings effect from
accumulation of fines between sandstone grains.
12
-------
As an additional example, the analysis of plant tailings
effluents for the Humeca Uranium Mill, which uses an alkaline lead
process, is given in table 3.2-3 (13).
The radiological significance of seepage from tailings ponds
will depend on the location of the pond. In arid regions, the
seepage may evaporate before leaving the site, leaving the radio-
activity entrained and absorbed on soil. Should the tailings pond
be located near a river, minor leakage might be diluted sufficiently
by the additional river water to meet relevant drinking water standards.
Discharge of pond seepage into streams providing insufficient dilution
and not under the control of the licensee would not be acceptable. In
such cases, a secondary dam may be built below the primary dam to
catch the seepage which may then be pumped back into the tailings ponds.
13
-------
Table 3.2-3 (13.)
Analysis of plant tailings effluents
from the Humeca Uranium Mill
(alkaline leach process)
Radionuclide pCi/1
Radium-226 10 to 2,000
Thorium-230 0.1
Uram'um-238 4,000
14
-------
4.0 The Model Uranium Mill
A model plant has been assumed in order to achieve a common
base for the comparison of radiation doses, committed health effects,
and radioactive effluent control technology.
The model mill is defined in terms of contribution to the
nuclear fuel cycle that is consistent with current designing and
projected commercial industry practice (6). However, it is not
necessarily representative of presently operating facilities.
Characteristics of the model mill are assumed to be:
a. 600,000 MT ore milled per year,
b. 1,140 MT U-jOg as yellowcake produced per year,
c. use of the acid leach process,
d. a tailings retention pond system which uses a clay-core earth
dam and local topographic features of the area to form the* impoundment,
e. collection and return of any seepage through the dam to the
tailings pond, and
f. location in a western State in an arid, low-populated density
region.
While Reference 1 considered the radiological impact of seepage
through a model clay core impoundment dam, it is now believed to be
standard practice (6) to collect and return any such seepage to the
tailings pond so that there are no routine liquid discharges of radio-
nuclides to water pathways from mills. The cost of a seepage control
15
-------
system is nominal compared to the cost of the tailings impoundment
system itself.
Radiation dose rates and health effects that might result from
the discharges of airborne radioactive effluents from the model mill
were calculated using standard x/Q values, dose conversion factors, •
model pathways, and health effect conversion factors that are similar
to those for other facilities in the previous discussion of the fuel
supply cycle. These factors and assumptions are discussed in Appendix
A of Reference 1.
The operating lifetime of a uranium mill is commonly from 12
to 15 years, depending upon the local ore supply and the demand for
uranium. In a few instances, the operating lifetime may be longer,
and allowances are sometimes made for that possibility if it appears
feasible. For the model mill, an operating lifetime of 20 years has
been selected.
16
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5.0 Radioactive Effluents from a Model Uranium Mill
Because regulations have not required uranium mills to report
the total amounts of each radionuclide discharged per year, the
source terms chosen for model mills are based on somewhat limited
operational information (6). Source terms listed in table 5.0-1
for model mills are believed, however, to be reasonably accurate
estimates of the quantities of radioactive materials discharged to
air pathways with base case controls. The controls assumed as the
base case consist of an orifice scrubber on the crusher and fine ore
bins, and a wet impingement scrubber in the yellowcake drying and
packaging areas. The milling procedures are so similar for acid and
alkaline leach processes that source terms for the two types of mills
are considered identical, except that the alkaline leach process does
not remove thorium from the ore so that, in this case, there is very
little thorium-230 as an impurity in the yellowcake dust.
The model mill is also assumed to use clay-core dam impoundment
technology for tailings with a catch basin if required to contain
seepage through the dam. Unless the impoundment area is lined with an
impervious material, considerable quantities (as much as 10 percent)
of the liquid effluent from the mill will leak out through the bottom
of the pond. However, because of the ion-exchange properties of most
soils, radionuclides dissolved in this effluent will attach to soil
particles and will not reach offsite locations or ground water. The
model mill is considered, therefore, to deliver no radiation exposure
to members of the general population through liquid pathways.
17
-------
Table 5.0-1
Discharge of Radionuclides to the Air from Model Uranium Millsa' and Tailings Piles (6_)
With Base Case Controls
Radionuclide
Chemical or
Physical State
Acid Leach Hill
i
Source Term
(mCi/yr)
Alkaline Leach" Mill
Source Term
(mCi/yr)
00
Uranium-238 and 234
Radium-2 26
Thorium-230
Uranium-238 and,234
Radium-226
Thorium-230
Uranium-238 and 234
Radium-226
Thorium-230
ore dust (oxides)
ore dust
ore dust
yellow cake (oxides)
yellow cake
yellow cake
tailings sand (0-10 y)
tailings sand (0-10 y)
tailings sand (0-10 y)
9.0
4.5
4.5
170.
0.2
4.7
0.2 - 0.8
1.3 - 4.2
1.4 - 4.5
9.0
4.5
4.5
170.
1.7
0.3 - 2.2
2.3 - 1.5
2.4 - 1.5'
(a)
6% moisture ore, radon-222 releases excluded
-------
Each site must be evaluated individually. If the ground
water table is high and the soil is low in ion exchange capacity
so that it becomes likely that radium-226 and thorium-230 will escape
from the tailings impoundment into underground waters, then the pond
area could be lined with an impervious membrane of asphalt to minimize
seepage. Acid wastes would have to be neutralized beforehand to
prevent damage to this type of liner.
The amount of radioactive particulate material removed from the
tailings beach by wind erosion is believed to depend on the area of the
beach, the wind velocity, and particle size distribution of the tailings
(j>). Estimates of this source term are included in table 5.0-1. Par-
ticles greater than lOy in diameter are not considered to be respirable
particles and are not included in the inhalation source term pathway.
Historically, windblown tailings have caused elevated gamma exposure
levels around piles, however, the inhalation pathway has been determined
to be the critical pathway. Levels of control sufficient to limit radi-
ation exposure through the inhalation pathway will also prevent, to a
significantly greater degree, exposures through the ground deposition,
whole body exposure pathway.
19
-------
6.0 Radiological Impact of a Model Mill
Estimates of the radiation doses to individuals through the air
pathway in the vicinity of an acid leach model mill using base case
controls from routine emissions are shown in table 6.0-1. The esti-
mated collective lung doses to the population in the vicinity of an
acid leach mill are given in table 6.0-2. The collective lung dose
is determined by summing the average individual radiation dose equiva-
lent to individuals living within 80 kilometers of the mill over the
total population within 80 kilometers of the mill. The models for the
dispersion and dose calculations are discussed in detail in Appendix A
of Reference (1). Based on the information available at the time that
analysis was performed, an effective half-life of 1,000 days was used for
insoluble class Y compounds in the pulmonary region of the lung in cal-
culating the lung doses from mill emissions. In accordance with what
is now becoming accepted practice, in this report all dose conversion
factors are calculated using a 500-day effective half-life (18), and
are, therefore, reduced by a factor of two from the previously used
values.
The dose conversion factor used to calculate the lung dose is
believed to be an order of magnitude more conservative than the dose
conversion factor used in Reference (6). Reasons for this difference
which relate to assumptions regarding lung model parameters, are dis-
cussed elsewhere. It is also assumed that food consumed by individuals
living near the mill is not produced locally so that exposure through
20
-------
Table 6.0-1
Radiation Doses to Individuals due to Inhalation
in the Vicinity of a Model Mill with Base Case Controls
Dose Equivalent to Critical Organ
Radionuclide
Uranium-234
and 238
Thorium-230
Radium-226
Source
Term
(mCi/yr)
180
15
10
Critical
Organ
Lung
Lung
Lung
Individual at Plant
Boundary
(mrem/yr)
170
15
15
Average Individual
Within 80 kms
(mrem/yr)
3.9 x 10~2
3 . 4 x 10~3
2.2 x 10~3
Total 205
200
4.5 x 10~2
-------
Table 6.0-2
Collective Dose to the General Population in the
Vicinity of a Model Mill with Base Case Controls
a
^O11 T"f* P*
n ,. . . , „ _ ., Critical Collective Critical Organ Dose
Radionuclide Term Pathway &
(mCi/yr) (person' rem/yr)
Uranium-234 and 238 180 Air Lung 2.2
Thorium-230 15 Air Lung 0.2
Radium-226 10 Air Lung 0.1
Total '2.5
aReleases to water pathways assumed equal to zero, and doses from radon-222 are
not included.
22
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food chains is not significant compared to lung exposures resulting
from the direct inhallation of radioactive particulate matter. The
radon exposure pathway was excluded from this report.
Because there are no liquid releases from the model mill, there
is no projected radiological impact through water pathways.
7.0 Health Effects Impact of a Model Mill
Potential health effects to members of the general population
in the vicinity of a model mill using base case controls are esti-
mated to be 0.0002 lung cancers per year of operation, or 0.005
such effects for 30 years of operation. The models used for the cal-
culation of health effects are given in Appendix A of reference (1).
23
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8.0 Control Technology for Uranium Milling
8.1 Airborne Effluent Control Technology
Hazardous airborne gaseous and particulate wastes are generated
in the milling operation from a number of different sources. The
major areas of the milling operations in which gaseous and particulate
matter effluents must be controlled are the ore crushing area, the
fine ore bins, and the yellowcake drying and packaging areas. Mills
often prefer to use multiple dust collection systems rather than design
a single, more elaborate system. There will usually be two or more ore
dust collectors and separate systems for the yellowcake dryer and for
the yellowcake packaging rooms.
Dust collector systems that are currently used or that can be
adapted for use by uranium mills are discussed in reference (j>) .
They are for the most part control technologies that have been proven
and are standard industrial equipment.
Briefly, these treatment methods are:
a. Orifice Scrubbers - The dusty air flows through a stationary
baffle system coated with a sheet of water. The dust particles
penetrate the water film and are captured.
b. Wet Impingement Scrubber - The dusty air carrying water
droplets added by preconditioning sprays passes through perforated
plates to atomize the water and to wet the dust. Particles are then
collected by impingement on baffle plates and a vaned demister.
c. Venturi Scrubber - The dusty air is passed through a venturi,
increasing its velocity. Water is added which atomizes in the gas
stream and collects the dust by impingement. The wetted dust is
24
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removed by demisters. Raising the pressure drop across the
venturi increases the collection efficiency, but this requires
higher energy levels and raises the costs.
d. Bag Filters - These filters are made of woven or felted
fabric and have high collection efficiencies provided the air
being filtered is cool and dry.
e. HEPA Filters - These filters are made of fiber glass.
They have very high efficiencies but have a number of limitations;
in particular, they can only be used in conjunction with a
prefilter and on dry air streams.
Current practice involves the use of wet dust control systems,
although several mills use bag filters for air flows from ore
handling and from the yellowcake packaging area. The costs and
percent effluent reduction for the various control systems suitable
for effluent streams of the model mill are given in table 8.1-1.
Particulate material can be prevented from being windblown off
the tailings pile beach by back filling with overburden and,as an
interim measure, by chemical stabilization by spraying with
petroleum derivatives. Chemical stabilization lasts about a year and
must be repeated on a regular schedule.
Other sources of gas and dust which can be controlled are the
open pit mine haul roads and the ore storage and blending piles. In
some instances, the liquid content of the ore as mined may be
sufficiently high to eliminate most dust formation in the ore
storage and blending area; due to insufficient information, this case
25
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Table 8.1-1
Cost and Efficiencies of Control Technology for Mills
Control Method
A. Gaseous (Crusher and Fine Ore Bins)
1. Orifice Scrubber
2. Wet Impingement Scrubber
3. Low Energy Venturi Scrubber
4. Bag Filters
B. Gaseous (Yellowcake Drying and Packaging)
1. Wet Impingement Scrubber 'c'
2. Low Energy Venturi Scrubber (c)
3. High Energy Venturi Scrubber
4. High Energy Venturi Scrubber + HEPA
Filters
Capital Cost
(dollars)
101,000
116,000
173,000
300,000
(35,000)
(35,000)
46,000
106,000
Annual
Operating Costs
(dollars)
7,200
$,600
17,000
21,000
(3,500)
(6,900)
15,000
22,000
Present Worth(b)
(dollars)
172,000
200,000
340,000
506,000
(69,000)
(103,000)
193,000
322,000
Percent
Effluent
Reduction
(%)
93.6
97,9
99.5
99.9
97.9
99.5
99.9
>99.99
C. Liquids, Solids, and Windblown Particulate
Matter
1. Clay Core Dam Retention System with 2,250,000
Seepage Return and 0.6 Meters (2 feet)
of Earth Cover Plus Rock Stabilization^6)
2. Chemical Control of Windblown Dust from 63,000
Tailings Pond Beach
3. Asphalt Liner for Tailings Pond^6) 800,000
50,000
8,000
0
(d)
2,750,000
142,000
800,000
100.00
100.00
(a)l974 dollars, radon-222 emissions not included.
(^Present Worth = Capital Cost + (Annual Cost x 9.818); 8% Discount Rate, 20 yr. Plant Lifetime.
^c'Costs for all yellowcake effluent control are shown for completeness. In actual practice, the value of
recovered product more than compensates the cost of control options Bl and B2.
^Includes investment to provide for perpetual care.
'e'160 acre tailings pile.
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will not be considered at present beyond stating that the problem
appears potentially significant and, that it can be controlled in
principle through sprinkling and by use of wind breaks. Dust genera-
tion on ore haul roads can also be controlled by sprinkling.
27
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8.2 Waterborne Effluent Control Technology and Solid Waste Control
Technology
New mills in the Rocky Mountains area are using impoundment
technology in order to approach zero liquid discharge levels. Recent
practice for treatment of solid and liquid wastes is to select a
natural ravine which has three basic qualifications for waste
storage: (a) limited runoff, (b) dammable downstream openings,
and (c) an underlying impermeable geologic formation. Diversion
systems (dams and canals) are used to limit the runoff area
emptying into the storage basin to prevent flooding of the ravine
during a postulated 50-100 year maximum rainfall occurrence. The
tailings dam, which should be clay-cored, is keyed into the underlying
impermeable formation, which, in one example, is a low porosity
shale. Tailings solids slurried in waste process liquids are
pumped to the impoundment reservoir for storage and liquid reduction.
Liquid reduction is accomplished primarily by evaporation, but
also by seepage through the dam, the reservoir walls and floor.
By filling a dammed natural depression with tailings, a relatively
flat, stable contour is achieved.
Two methods for seepage collection and return are being
considered for new mills. Seepage has been estimated to occur from
a clay-core retention dam at a rate of 300 liters per minute. In
that situation when an impermeable geological formation underlies
the retention system, seepage can be collected in a catch basin
located at the foot of the dam. The collected seepage can be pumped
28
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back into the retention 'pond thus eliminating release to the offsite
environment. In that situation where either an underlying imper-
meable geological formation is not existent or is not continuous,
vertical seepage may occur to the underlying ground water formation.
Wells may be drilled downstream of the retention system into the
subsurface formations where seepage will collect, and this water
is pumped back to the retention system. Such a system requires
specific favorable subsurface conditions. In both cases, these
control costs are small compared to the cost of the clay core dam
retention system (1) .
Impoundment of solids is being accomplished in older mills
merely by construction of a dike with natural materials and
filling the diked area with slurried tailings. When full, the
height of the dike is increased with dried tailings to accommodate
even more waste material. Process liquids which overflow the tailings
dike or seep through the dike are sometimes routed through a treat-
ment system and discharged to the environment. The diking procedure
which is less costly initially, creates an above-ground pile of
tailings which is difficult and costly to stabilize. While the
mill is operating, this type of pile is also subject to wind and water
erosion. Field studies at tailings piles after mill shut-down have
shown high gamma radiation levels in the vicinity of such piles,
elevated radium-226 levels in water supplies, and high airborne levels
of thorium-230 and radium-226 due to wind blown tailings (14,15,16,17).
For these reasons, new mills are not likely to be built using this type
of solid waste control.
29
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Stabilization of tailings piles requires grading of the tailings
area to lessen side slopes, establishing drainage diversion, covering
with nonradioactive material, and revegetating the area. In
semiarid regions it may be necessary to initially irrigate the pile
to achieve vegetation growth. Other types of stabilization may also
be feasible. One method involves the covering of the tailings
with large aggregate gravel from a river bottom. Silt fines which
accompany the river gravel will blow away in a short time leaving
what is affectively a wind-proof rip rap, thus significantly
reducing or eliminating migration of the tailings outside the
controlled area. The costs of such stabilization has recently been
estimated (6) at $350/acre-ft for earth, and $2,000/acre-ft for rock.
The cost associated with stabilizing a diked surface pile is sig-
nificantly higher and probably less effective because of difficulties
faced in grading, covering, and revegetating the potentially steep
side slopes.,
Uranium mill tailings piles are long half-life, low-level
radioactive wastes. As such, they will require perpetual care. This
will include occasional inspection and maintenance to insure integrity
of the stabilizing cover, fencing, and of the warning signs around
the pile. An annuity should be included as part of the cost of the
control technology to pay for this care. The maintenance associated with
perpetual care of a stabilized dike system would probably be higher
than that for the depression fill system, since there is tendency toward
collapse of side slopes and possibly inadequate drainage of precipition
from the pile.
30
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9.0 Effluent Control Technology for the Model Mill
Typical current effluent control systems were assumed for
the model mill. They were:
a. Ore Crusher and Ore Bin Dust - Orifice Scrubber.
b. Yellowcake Dryer and Packaging Dust - Wet Impingement
Scrubber.
c. Liquid and Solid Waste - Clay-core dam retention system
(160 acres) with seepage return and exposed beach. To be stabilized
with 2 feet of earth cover and 6 inches of rock cover.
The radiological impact of total airborne effluent versus
successively more effective control systems for a model uranium mill
are listed in table 9.0-1. Each improvement in control is the most
cost-effective available at that level of control.
The output of the model plant using base case controls is 1,140
MT U^Og of which approximately 1% is recovered by the wet impingement
dust collector system during drying and packaging operations (6). The
value of 11,00 kilograms (24,000 Ibs) of recovered yellowcake more than
compensates for the cost of this control system. The low energy venturi
scrubber is 1.6% more efficient that the wet impingement scrubber and
will recover an estimated additional 200 kilograms (440 Ibs)
of yellowcake per year. The value of this additional recovered yellow-
cake is approximately equal to the increased annual operating costs of
the low energy venturi scrubber as compared to the wet impinger. The
present worth of these systems are, therefore, not included as a control
cost for the model mill.
31
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Table 9.0-1
Radiological Impact of Airborne Effluents versus Control Costs for a Model Uranium Mill
u>
NJ
Controls
(Table 8.1)
None
Al; Bl(c)(d)
Al; B2(d)
Al; B3
A2; B3
A2; B3; C2
A2; B4; C2
A3; B4; C2
A4; B4; C2
cKO
Source Term'3'
(mCi/yr)
>20,000
205
75
35
25
15
6
1.5
0.3
0
Maximum Lung
Dose to an
Individual 20,000
200
73
34
24
15
6
1.5
0.3
0
Present Worth
(1974 $/facility)
0
172,000
172,000
262,000
290,000
432,000
561,000
701,000
867,000
2,750,000
(a)Alpha emitting radionuclides as insoluble, respirable particulate matter.
('b)For the assumed worst case of an individual permanently occupying a location exhibiting
a x/Q of 6 x 10~6 s/m3.
(c)Assumed current level of controls for new mills.
sts for control options Bl and B2 not included, since they are more than compensated by
the value of product recovered.
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10.0 Retrofitting Control Technology to Operating Uranium Mill
The cost and practicality of retrofitting control technology
systems to an operating uranium mill was not included in Reference
(j6). The cost is judged to be approximately the same order of
magnitude as the cost to install the same control system in a new
mill.
The cost and practicality of retrofitting control measures
to operational tailings piles that do not use clay core dam impound-
ment technologies must be considered on an individual basis.
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REFERENCES
1. U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Analysis of
the Uranium Fuel Cycle, Part I - Fuel Supply, EPA-520/9-73-003-B,
Office of Radiation Programs, Environmental Protection Agency,
Washington, D.C. 20460 (October 1973).
2. U.S. ENVIRONMENTAL PROTECTION AGENCY, Environmental Analysis of
the Uranium Fuel Cycle, Part II - Nuclear Power Reactors, EPA-
520/9-73-003-C, Office of Radiation Programs, Environmental Pro-
tection Agency, Washington, D.C. 20460 (November 1973).
3. U.S. ENVIRONMENTAL PROTECTION AGENCY. Environmental Analysis of
the Uranium Fuel Cycle, Part III - Nuclear Fuel Reprocessing,
EPA-520/9-73-003-D, Office of Radiation Programs, Environmental
Protection Agency, Washington, D.C. 20460 (October 1973).
4. U.S. ENVIRONMENTAL PROTECTION AGENCY, Environmental Radiation
Protection for Nuclear Power Operations, 10 CFR Part 190, Federal
Register, Vol. 40 No. 109 (Thursday, May 29, 1975).
5. U.S. ATOMIC ENERGY COMMISSION. Draft Environmental Statement
Related to the Utah International, Inc. Shirley Basin Uranium
Mill, Shirley Basin, Wyoming, Docket No. 40-6622, Fuels and
Materials Directorate of Licensing, U.S. Atomic Energy Commission,
(June 1974).
6. SEARS, M.B. et.al. "Correlation of Radioactive Waste Treatment Costs
and the Environmental Impact of Waste Effluents in the Nuclear Fuel
Cycle for Use in Establishing 'as Low as Practicable1 Guides -
Milling of Uranium Ores," ORNL-TM-4903, Two Volumns, Oak Ridge
National Laboratory, Oak Ridge, Tennessee 37830 (May 1975).
7. TEKNEKRON, INC. "Scopping Assessment of the Environmental Health
Risk Associated with Accidents in the LWR Supporting Fuel Cycle -
Draft Report"-EPA Contract No. 68-01-2237, Teknekron, Inc.
Washington, D.C. 20036 (September 2, 1975).
8. "Controlling the Radiation Hazard from Uranium Mill Tailings" Report
of the Congress by the Comptroller General of the United States
RED-75-365 (May 21, 1975).
9. HUMBLE OIL AND REFINING COMPANY. Applicant's Environmental Report,
Highland Uranium Mill, Converse County, Wyoming. Minerals Depart-
ment, P.O. Box 2180, Houston, Texas 77001 (July 1971).
34
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10. HUMBLE OIL AND REFINING COMPANY. Supplement to Applicant's Environ-
mental Report, Highland Uranium Mill, Converse County, Wyoming.
Minerals Department, P.O. Box 2180, Houston, Texas 77001 (January
1972).
11. U.S. ENVIRONMENTAL PROTECTION AGENCY. Evaluation of the Impact of
the Mines Development, Inc. Mill on Water Quality Conditions in the
Cheyenne River. EPA Region VIII, Denver, Colorado 80203 (September
1971).
12. U.S. ENVIRONMENTAL PROTECTION AGENCY. Interim Primary Drinking
Water Regulations - 40 CFR Part 141 - Federal Register, Volumn 40,
No. 158 (Thursday, August 14, 1975).
13. U.S. ATOMIC ENERGY COMMISSION. Draft Detailed Statement on the
Environmental Considerations Related to the Proposed Issuance of
a License to the Rio Algom Corporation for the Humeca Uranium
Mill, Docket No. 40-8084. Fuels and Materials Directorate of
Licensing, U.S. Atomic Energy Commission, Washington, D.C. 20545
(December 1972).
14. SNELLING, R. N. and SHEARER, S. D., Jr. Environmental Survey of
Uranium Mill Tailings Pile, Tuba City, Arizona. Radiological Health
Data and Report 10:475-487 (November 1969).
15. SNELLING, R. N. Environmental Survey of Uranium Mill Tailings Pile,
Monument Valley, Arizona. Radiological Health Data and Report 11:511-
517 (October 1970).
16. SNELLING, R. N. Environmental Survey of Uranium Mill Tailings Pile,
Mexican Hat, Utah. Radiological Health Data and Report 12:17-28
(January 1971).
17. U.S. ENVIRONMENTAL PROTECTION AGENCY. Radium-226, Uranium, and Other
Radiological Data from Water Quality Surveillance Stations Located
in the Colorado River Basin of Colorado, Utah, New Mexico, and Arizona,
January 1961 through June 1972. 8SA/TIB-24, EPA Region VIII, Denver,
Colorado (July 1973).
18. International Commission on Radiological Protection, The Metabolism
of Compounds of Plutonium and Other Actinides, Adopted May 1972, ICRP
Publication 19, Pergammon Press, New York (1972).
GPO 900-1 90
35
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