40CFRpart61
National Emission Standards
for Hazardous Air Pollutants
EPA402-R-96-021
NESHAPS RULEMAKING ON NUCLEAR
REGULATORY COMMISSION AND AGREEMENT STATE LICENSEES
OTHER THAN NUCLEAR POWER REACTORS
BACKGROUND INFORMATION DOCUMENT
December 1996
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
Washington, DC 20460
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PREFACE
The Environmental Protection Agency (EPA) is rescinding 40 CFR 61, Subpart I,
National Emission Standards for Radionuclide Emissions From Facilities licensed by the
Nuclear Regulatory Commission and Federal Facilities Not Covered by Subpart H as it
applies to licenses other than commercial nuclear power reactors. This Background
Information Document (BID) has been prepared in support of rulemaking proceedings for
EPA's action. This BID contains an introduction, descriptions of Nuclear Regulatory
Commission (NRC) source subcategories, estimates of doses from both designated and
randomly selected NRC facilities, a comparison of NRC and EPA regulations governing
emissions of radioactive material, estimates of the number of NRC facilities that are in
compliance with Subpart I, and a description of quality control measures used in this BID.
Copies of this BID, in whole or in part, are available to all interested persons. An
announcement of the availability appears in the Federal Register. For additional information,
contact Julie Rosenberg at (202) 233-9154 or write to:
Director, Radiation Protection Division
Office of Radiation and Indoor Air (6602J)
Environmental Protection Agency
401 M Street, SW
Washington, DC 20460
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DISCLAIMER
Mention of any specific product or trade name in this report does not imply an
endorsement or guarantee on the part of the Environmental Protection Agency.
LIST OF PREPARERS
Various staff members from EPA's Office of Radiation and Indoor Air contributed to
the development and preparation of the BID.
Albert Colli
Craig Conldin
Dale Hoffmeyer
Larry Gray
Byron Bunger
Chief, Air Standards and
Economics Branch
Health Physicist
Health Physicist
Environmental Scientist
Economist
Reviewer
Writer/Reviewer
Reviewer
Reviewer
Reviewer
An EPA contractor, S. Cohen & Associates, Inc., McLean, VA, provided significant
technical support in the preparation of the BID.
IV
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Contents
Pas
Preface ...:......... v iii
Disclaimer iv
list of Preparers . . . : . -. • • • • iy
List of Tables ...................... viii
List of Figures . . x
Section
1. Introduction and Summary 1-1
LI History of Standards Development 1-1
1.2 Update Methodology ... . . . '. ....... 1-4
1.3 Purpose of this Background Information Document .............. 1-6
1.4 Summary . 1-7
2. Description of Regulatory Programs 2-1
2.1 The EPA's Regulatory Program under the Clean Air Act 2-1
2.1.1 Requirements 2-1
2.1.2 Methods for Demonstrating Compliance . 2-1
2.2 The NRC's Regulatory Program under the Atomic Energy Act . . 2-4
2.3 Comparison of the NRC's Requirements with the NESHAP . . . .-...". . . 2-6
2.4 NRC-Licensed Facility Program Analysis 2-9
i
3. Results of Designated Survey of NRC-Licensed Facilities ............... 3-1
3.1 Uranium Fuel Cycle Facilities ... r.................. 3-2
3.1.1 Uranium Mill Tailings . . . 3-2
3.1.2 Uranium Conversion Facilities 3-5
3.1.3 Fuel Fabrication Facilities 3-8
3.1.4 Interim Spent Fuel Storage Facilities 3-12
3.2 Test and Research Reactors . . , . 3-12
3.2.1 Previous Evaluations . . 3-13
3.2.2 Evaluations of Specific Facilities Made During the Reconsideration
Period 3-14
3.2.3 Results of the Designated Survey of Test and Research Reactors . 3-16
3.3 Radiopharmaceutical and Radiolabeled Compound Manufacturers 3-16
3.3.1 Previous Evaluations 3-16
3.3.2 Evaluations of Specific Facilities Made During the Reconsideration •
Period .............•............-...._...:.'... 3-16
3.3.3 Results of the Designated Survey for Radiopharmaceutical and
Radiolabeled Compound Manufacturers 3-19
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Contents (Continued)
Section
3.4 Hospitals and Medical Research Facilities 3-20
3.4.1 Previous Evaluations 3-20
3.4.2 Evaluations of Specific Facilities Made During the Reconsideration
Period . . 3-21
3.4.3 Results of the Designated Survey for Hospitals and Medical
Research Facilities . . . 3-25
3.5 Manufacturers of Sealed Sources 3-26
3.5.1 Previous Evaluations . . 3-26
3.5.2 Evaluations of Specific Facilities Made During the Reconsideration
Period . ; 3-26
3.5.3 Results of the Designated Survey for Manufacturers of Sealed
Sources 3_2g
3.6 Testing of Depleted Uranium Munitions . . 3-29
3.6.1 Previous Evaluations 3-30
3.6.2 Evaluations of Specific Facilities Made During the Reconsideration
Period 3-30
3.6.3 Results of the Designated Survey for Testing of Depleted Uranium
Munitions 3_31
3.7 Rare Earth and Thorium Processors (Source Material) 3-32
3.7.1 Previous Evaluations • . '.....' 3-32
3.7.2 Evaluations of Specific Facilities Made During the Reconsideration
Period . . . 3-33
3.7.3 Results of the Designated Survey for Rare Earth and Thorium
. Processors .- 3-36
3.8 Commercial Low-Level Radioactive Waste Disposal and Incineration . . 3-37
3.8.1 Previous Evaluations 3-38
3.8.2 Evaluations of Specific Facilities Made During the Reconsideration
Period 3-39
3.8.3 Results of the Designated Survey for Waste Disposal and
Incineration 3-39
3.9 Summary of Results 3.40
4. Results of Random Survey of Licensees 4-1
4.1 Purpose of the Random Survey 4-1
4.2 Methods for Selecting the Random Sample and Data Requirements 4-2
4.2.1 Selection Criteria . 4-2
4.2.2 Data Requirements ............ 4-2
4.3 Methods for Evaluating Data .4-5
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Contents (Continued)
4.4 Raw Results of the Survey ...... .... .............. 4-6
4.4.1 Results 4-6
4.4.4 Translation from Dose to Risk 4-7
4.4.2 Assumptions . . . 4-8
4.4.3 Population Dose Estimates 4-8
4.5 Statistical Interpretation of the Results . . ... 4-9
4.5.1 Frequency Distribution Analysis 4-14
4.5.2 Cumulative Distribution Analysis . 4-17
5. Quality Control . . 5-1
References R-l
Appendices
A - NRC's Organization, Regulations, and Controls ................ A-l
B - Selected NRC Regulatory Guides . . B-l
C - Description of Licensed Activities . C-l
D - Description of Facilities Evaluated D-l
E- Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants . E-l
F - NRC Agreement States and State Directors ................... F-l
G - Random Survey Questionnaire . . . . G-l
H - Dose Calculation Assumptions H-l
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Tables
Number page
1-1 Summary of estimated doses 'm 1-9
2-1 Summary of regulatory requirements ;-.... 2-8
3-1 COMPLY code input data for uranium mills 3-4
3-2 Atmospheric radioactive emissions assumed for reference dry and wet process
uranium conversion facilities -. 3-7
3-3 Light water reactor commercial fuel fabrication facilities licensed by the
Nuclear Regulatory Commission as of January 1988 . . ,3-9
3-4 Light water reactor commercial fuel fabrication facilities reported annual
uranium effluent releases for 1983 through 1987 in mCi/yr 3-11
3-5 Atmospheric radioactive emissions assumptions for reference fuel fabrication
facility 3-n
3-6 Licensed test reactors in the United States as of August 1991 ............ 3-13
3-7 Effluent release rates (Ci/yr) for test and research reactors 3-14
3-8 DuPont Boston emission data 3.17
3-9 DuPont Billerica emission data 3-18
3-10 Mallincrodt emission data . . . 3-18
3-11 Hospital and medical research facilities effluent release rates 3-23
3-12 Effluent release rates (Ci/yr) for sealed source manufacturers ........... 3-27
3-13 Source term used for Aberdeen Proving Ground 3-31
3-14 Distances to receptors at Aberdeen Proving Ground 3-31
3-15 Rare earth processors' annual release rates ? .- 3.35
3-16 Summary of Designated Survey doses . . . . 3.41
vru
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Tables (Continued)
Number
4-1 Summary of Random Survey responses • • • • • • • 4-3
4-2 Number of facilities having doses in various ranges 4-7
" * , ' " - - "
4-3 Population dose estimates ; 4-9
4-4 Estimated distribution of maximum individual doses 4-11
4-5 Estimated distribution of maximum individual doses for radioiodine . 4-12
4-6 Estimated percentage and number of facilities exceeding specified dose
using the lognormal and hybrid-lognormal models ,. 4-31
rx
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Figures
Number .
4-1 Frequency Distribution of Dose for 367 Facilities, with Fitted Model . 4-15
4-2 Frequency-Distribution of Iodine Dose for 290 Facilities, with Fitted Model . . 4-16
4-3 Frequency Distribution of Iodine Dose for 290 Facilities, with Fitted Models . . 4-18
4-4 Cumulative Dose Distribution 4-19
4-5 Extreme Tail of Sample Distribution 4-20
4-6 Cumulative Iodine Dose Distribution . 4-21
4-7 Extreme Tail of Iodine Distribution 4-23
4-8 HLN-Probability Plot with Rho=0.14 4-24
4-9 HLN-PrpbabiKty Plot with Rho=7.7 ................ 4-25
4-10 Cumulative Dose Distribution 4-26
4-11 Extreme Tail.of Sample Distribution 4-27
4-12 Cumulative Iodine Dose Distribution 4-29
4-13 Extreme Tail of Iodine Distribution .'.....' 4-30
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1. Introduction and Summary
1.1 HISTORY OF STANDARDS DEVELOPMENT
Pursuant to the 1977 amendments to the Clean Air Act (CAA), on December 27,
1979, the administrator listed radionuclides as a hazardous air pollutant under Section 112 of
the Act (44 FR 76738). The Administrator then initiated studies to determine what source
categories of facilities emit radionuclides to the air in quantities sufficient to warrant
establishing a NESHAP (National Emission Standard for Hazardous Air Pollutants) to limit
emissions to levels providing an ample margin of safety to protect the public health.
On April 6, 1983, EPA published a FR notice proposing radionuclide NESHAPs for
four source categories and announced its finding that NESHAPs were not required for seven
of the source categories that it had investigated (48 FR 15076). NESHAPs were proposed to
limit emissions of radionuclides from elemental phosphorus plants, Department of Energy
(DOE) facilities, certain non-fuel cycle facilities licensed by NRC, and underground uranium
mines. Uranium fuel cycle facilities were one of the seven source categories that the
Administrator determined did not require a NESHAP.
In October, 1984, acting pursuant to a court order to take final action on the proposed
NESHAPs, the Administrator published a FR notice announcing that the proposed standards
for elemental phosphorus plants, DOE facilities, and certain NRC-licensed facilities were
being withdrawn (49 FR 43906). The decision to withdraw the proposed standards was
based on the Administrator's finding that control practices already in effect for those source
categories provide an ample margin of safety. The FR notice also made final the
Administrator's decision not to issue NESHAPs for the other seven source categories.
The decision to withdraw the proposed NESHAPs was immediately challenged in .
court, and on December 11, 1984, the U.S. District Court for the Northern District of
California found the Administrator in contempt of its earlier order directing the
Administrator to promulgate final standards or to make a finding that radionuclides are not a
hazardous air pollutant. EPA complied with the court's December decision by issuing
NESHAPs for elemental phosphorus plants, DOE facilities, and certain NRC-licensed
facilities on February 6, 1985 (50 FR 7280).
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The Environmental Defense Fund (EDF), the Natural Resources Defense Council
(NRDC), and the Sierra Club filed petitions with the court to review the final decisions not
to regulate certain source categories (including the uranium fuel cycle) and the February
1985 standards. On July 28, 1987, while these petitions were pending, the U.S. Court of
Appeals for the District of Columbia remanded to the Agency the NESHAP for vinyl
chlorides (a nonradioactive hazardous air pollutant). In that decision, the court concluded
that the Agency had improperly considered cost and technological feasibility in determining
the level of the standard without first making a determination based exclusively on the risk to
the public. *
Given the court's decision on vinyl chloride, EPA determined that its radionuclide
NESHAPs should also be reconsidered and petitioned the court for a voluntary remand of
standards. In its petition, EPA also moved that the pending litigation on all issues relating to
its radionuclide NESHAPs be placed in abeyance during the rulemaking and .agreed to
reexamine all issues raised by the parties to the litigation. The court granted EPA's petition
on December 8, 1987, and EPA began to revisit its earlier decision under a court-imposed
schedule.
The Administrator's final decisions on the radionuclide NESHAPs were published in
the Federal Register on December 15, 1989 (54 FR 51654). The final NESHAP for the
NRC-licensed facilities (40 CFR 61, Subpart I) included facilities that are part of the uranium
fuel cycle and established a standard of 10 mrem/yr effective dose equivalent (ede) to any
member of the public, with no more than 3 mrem/yr ede caused by emissions of
radioiodines. In explaining his decision to promulgate a NESHAP that included the uranium
fuel cycle facilities previously unregulated under the CAA, the Administrator explained that
the standard would insure that the current levels of emissions do not increase.
Simultaneously with promulgating the NESHAPs, EPA granted reconsideration of 40
CFR Part 61, Subpart I, National Emissions Standards for Radionuclide Emissions From
Facilities Licensed by the Nuclear Regulatory Commission and Federal Facilities Not
Covered by Subpart H. The reason for the reconsideration was to allow assessment of
information received late in the rulemaking process from NRC and the National Institutes of
Health (NIH) regarding the impacts of duplicative regulations on licensees and the potential
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for the NESHAP to discourage the use of radioisotopes in medical and experimental
therapies. The Agency also stayed the effective date of Subpart I. Over the next year, EPA
continued to stay Subpart I in its entirety.
While Subpart I was under reconsideration, Congress passed the Clean Air Act
Amendments (CAAA) of 1990. Section 112(d)(9) of these amendments states, in part:
No standard for radionuclide emissions from any category or subcategory of
facilities licensed by the Nuclear Regulatory Commission (or an Agreement
State) is required to be promulgated under, this section if the Administrator
determines, by rule, and after consultation with the Nuclear Regulatory
Commission, that the regulatory program established by the Nuclear
Regulatory Commission pursuant to the Atomic Energy Act for such category
or subcategory provides an ample margin of safety to protect the public health.
EPA reviewed the information provided to it during the reconsideration of Subpart I
concerning radionuclide emissions from one subcategory of NRC-licensed facilities,
commercial nuclear power reactors. In light of the new authority provided by CAAA
Section 112(d)(9), EPA made an initial determination that the NRC's regulatory program
limiting these emissions protects public health with an ample margin of safety. On March
13^ 1991, EPA issued an Advance Notice of Proposed RulemaMng (ANPR), announcing its
intention to proceed with the rulemaking, pursuant to Section 112(d)(9), to rescind Subpart I
of 40 CFR 61, as it applies to nuclear power reactors (56 FR 10524). Concurrent with the
ANPR, EPA published a FR notice proposing to stay the effectiveness of Subpart I for power
reactors until the conclusion of the rulemaking on recision (56 FR 10523). On August 5,
1991, EPA issued a Notice of Proposed Rulemaking (NPR) announcing the Agency's
intention to rescind Subpart I as it applies to nuclear power reactors (56 FR 37196). At, the
same time, EPA stayed subpart I for these facilities pending completion of the recision
rulemaking.
For all other categories of NRC licensees, EPA concluded that it lacked adequate
information to characterize the facilities' emissions and embarked on the information "
collection survey (under Section 114) that is described in detail in this Background
Information Document (BID). On April 15, 1991, EPA stayed the effectiveness of Subpart I
for all NRC-licensed facilities other than nuclear power reactors until November 15, 1992, or
until such earlier date that EPA is prepared to make an initial determination under CAAA
Section 112(d)(9) and conclude its reconsideration (56 FR 18735, April 24, 1991). On
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December 1, 1992, the Administrator determined that the NRC regulatory program for
licensed facilities other than commercial nuclear reactors should provide an ample margin of
safety to protect public health, and proposed to rescind Subpart I as it applies to such
licensees. .
1.2 UPDATE METHODOLOGY
In previous evaluations (EPA83, EPA84, EPA89), EPA used both actual and model
facilities to characterize the doses and risks caused by airborne emissions of radionuclides
from. NRC-licensed facilities. In those assessments, the doses caused by activities judged to
have the greatest potential for relatively large airborne emissions were evaluated primarily on
the basis of actual facilities. Most of these large facilities had emissions data providing a
basis for reasonable estimates of doses and risks to public health. Emissions from model
facilities were used in an attempt to bound the doses and risks from the thousands of facilities
engaged in activities judged to have less potential to cause exposures. These estimates were
based on available data and conservative assumptions to ensure that doses and risks were not
understated.
The most recent evaluation of emissions from NRC-licensed facilities, conducted for
the 1989 promulgation of the Subpart I NESHAP, used the methodology that EPA developed
to meet the approach that the U.S. Court of Appeals for the D.C. Circuit set out in the Vinyl
Chloride decision. That approach requires two steps in setting standards: first, determine an
"acceptable" level of risk that considers only health factors, and second, set a standard that
provides an "ample margin of safely," in which cost, feasibility, and other relevant factors in
addition to health may be considered. The Agency's methodology utalizes a multifactor
approach which focuses on three measures of risk:
• Maximum Individual Risk (MIR) - an estimate of the risk incurred by the
individuals most exposed to the effluent from a given facility. For
radionuclide NESHAPs, EPA estimated the lifetime fatal cancer risk that
would result from continuous exposure over the individual's entire lifetime. A
lifetime MIR of approximately 1 in 10,000 (1E-04) is judged to be
presumptively acceptable.
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• Incidence - an estimate of the total number of health effects in the population
residing within 80 kilometers of the facilities in the source category.
Incidence is considered with other health risk information in judging
acceptability.
• Risk Distribution - an estimate of the number of persons at a given level of
MER and the estimated fraction of the total number of health effects expected
to be incurred in, the population within each range of risks. As a goal, EPA
seeks to assure that as many individuals as possible are at an MIR of 1 in
1 million (1E-06) or less.
Using these criteria, EPA found that the risk from all actual facilities evaluated (both
NRC-Eeensed facilities covered under Subpart I and uranium fuel cycle facilities that were
not at that time covered by a NESHAP) were acceptable. The evaluations based on
conservatively modeled facilities also met these criteria. While the highest doses estimated
for any actual facility in those assessments were within the range that the Administrator has
determined to be safe, the total number of facilities and the diversity of the activities in
which they are engaged resulted in some uncertainty that the facilities causing the highest
individual doses had actually been identified and evaluated.
\
To provide the Administrator with enough information to determine whether the
NRC's regulatory program protects public health with an ample margin of safety, the Agency
has performed additional dose estimates to provide a "snapshot" of current emissions and
doses. EPA has also analyzed the NRC's regulatory program to determine if the program
can ensure that future emissions provide for the public health with an ample margin of
safety.
For the large facilities previously evaluated by EPA, updated emissions,
meteorological, and population information was obtained and new dose estimates made to
. better account for previous limitations. Dose estimates were also performed for facilities that
had not been studied earlier but where concerns remained that radioactive emissions could
present significant risks. These analyses are called the Designated Survey.
For a more accurate characterization of the doses attributable to the many smaller
licensees that previously had been evaluated by using model facilities; EPA has taken a
statistical approach, based on a random sample of a subset of NRC and Agreement State
licensees. Facilities that had already been evaluated by EPA and facilities that are only
licensed to use sealed sources of radioactive materials were excluded from the subset of
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licensees to be surveyed. The random sample of licensees was selected from lists provided
by NRC and the Agreement States. As these lists included some facilities licensed to use,
.only sealed sources, over-sampling was employed and all selected facilities that only use
sealed sources were excluded from the analysis. From this random sample, information from
367 users of radioactive material was evaluated. The data needed to evaluate doses were
obtained by a survey form mailed to each randomly chosen facility, and doses were estimated
using the COMPLY computer code. These.analyses are called the Random Survey.
1.3 PURPOSE OF THIS BACKGROUND INFORMATION DOCUMENT
* - '
This BID provides background information from the Designated and Random Surveys
to assist the Administrator in determining whether the NRC's regulatory program maintains
radioactive emissions sufficiently low to protect the public health with an ample margin of
safety. This BID also includes a comparison of NRC and EPA regulations governing
airborne radioactive emissions and a detailed description of the Agency's procedures and
methods for estimating radiation dose due to radioactive emissions to the air. This material
is presented as follows: ,
• Chapter 2 - A description of the EPA regulations that limit the effective dose
equivalent to members of the general public and the method for determining
compliance with that dose limit. This chapter also summarizes the
organizational and administrative controls imposed by NRC on materials
licensees.
• Chapter 3 - A description of the annual doses resulting from emissions from
designated NRC licensees (the Designated Survey). This chapter also provides
the reasons for the selection of the designated facilities.
• Chapter 4 - A description of the annual doses resulting from emissions from
randomly selected NRC licensees (the Random Survey). This chapter also
describes the methods for ensuring the selection of a statistically significant
random sample, data requirements for performing realistic dose estimates, and
the statistical methods used to evaluate the raw data.
• Chapter 5 - A description of the quality control measures instituted to ensure a
high level of confidence in the results of the BED. .
This BID also contains several appendices. Appendix A describes the organization of
NRC and its regulations. Appendix B describes various NRC Regulatory Guides pertinent to
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radioactive emissions and exposure control. Appendix C describes the various licensee
activities for which an NRC or Agreement State license is required. Appendix D identifies
the types of facilities selected for the Random Survey. Appendix E describes quality
assurance requirements for nuclear power plants and fuel reprocessing plants. Appendix F
lists the NRC Agreement States and contact persons. Appendix G contains a copy of the
questionnaire sent to the randomly selected facilities to obtain site-specific information.
Appendix H discusses the assumptions used in the dose calculations performed.
1.4 SUMMARY
The major findings of this BID for NRC licensed facilities other than nuclear power
reactors include:
1. The highest dose found in the Random Survey was 8 mrem/yr from all radionuclides
and 0.7 mrem/yr from radiOiodines. The highest dose found in the Designated
Survey was 8 mrem/yr from all nuclides and 1 mrem/yr from radioiodines. This
indicates that, in general, the doses being received by the members of the public at
greatest risk are lower than the NESHAP standard established by the Administrator
(10 mrem/yr ede with not more than 3 mrem/yr ede caused by radioiodines).
2. Because the doses received by members of the public vary from year to year for any
given facility, a trend for all facilities could not be established from the available
data. However, NRC regulatory requirements have become more stringent over time,
and it may be inferred that this will result in a downward trend in future airborne
• releases.
3. NRC begins to consider doses received by the public from radioactive effluents at the
time of license application and continues to evaluate the potential for effluents to
cause doses in excess ';of regulatory limits throughout a facility's lifetime. The
stringency of NRC's requirements varies with the potential of licensed facilities to
place the health and safety of the public at risk. However, all facilities must comply
with the limits established in 10 CFR Part 20, Standards for Protection Against
Radiation, and fuel cycle facilities must also meet the requirements of 40 CFR 190,
f
Environmental Radiation Protection Standards for Nuclear Power Operations.
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4. NRC has recently amended the requirements in 10 CFR Part 20. The amendments,
consistent with Federal guidance and the International Commission on Radiological
Protection (ECRP), establish a risk-based system of dose limitations. For members of
the general public, the amendments lower the maximum permissible dose to
100 mrem/yr total effective dose equivalent (tede) from direct radiation and exposure
to gaseous and liquid effluents. The derived air concentrations (DACs) that may be
used to demonstrate compliance with the 100 mrem/yr tede limit are based on
50 mrem/yr tede to account for multiple pathways.
5. The Part 20 amendments also establish the requirement, previously just guidance, that
all licensees conduct operations in a manner such that doses to both workers and
members of the public are as low as is reasonably achievable (ALARA).1 Revised
Part 20, although still allowing a higher maximum permissible dose than the
NESHAP, is more restrictive than the regulations that have resulted heretofore in ,
actual doses to members of the public below the NESHAP limits. For this reason,
the NRC program should result in future emission levels no higher than current
emissions levels.
Based on the results obtained from the Random Survey of NRC-licensed facilities, the
vast majority of facilities (over 99.5 percent) are not causing doses greater than the NESHAP
standards of 10 mrem/yr ede from all radionuclides with not more than 3 mrem/yr ede from
radioiodJnes. In fact, the majority (>95 percent) have emissions that result in doses of less
than 1 mrem/yr ede. Based on statistical considerations, EPA expects that 14 facilities out of
approximately 6,000 may cause doses in excess of the NESHAP standard.
Estimated doses from the Designated Survey and the Random Survey are summarized
in Table 1-1.
J Per 10 CER 20, ALARA is an acronym for "as low as is reasonably achievable" and means making every
reasonable effort to maintain exposures to radiation as far below the dose limits in 10 CER Part 20 as is
practical, consistent with the purpose for which the licensed activity is undertaken. The requirement takes into
account the state of technology, the economics of improvements in relation to the state of technology, the
economics of improvements in relation to benefits to the public health and safety, and other societal and
socioeconomic considerations, and the value of utilizing nuclear energy and licensed materials in the public
interest. ,
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Table 1-1. Summary of estimated doses.
Survey
Designated
, Survey .
Random Survey2
licensee
U-Fuel Cycle
Category "
U-Mills & Tailings
UF6 Conversion
(Wet Cycle)
Fuel
Fabrication
Test and Research Reactors
Radiopharmaceutical Manufacturers
Hospitals and Medical
Research Facilities
Manufacturers of Sealed Sources
Depleted Uranium Munitions
Rare Earth Processors
Commercial Low-Level Radioactive Waste
Disposal and Incineration1
Maximum
Estimated Dose
rarem/yrede
2
7
6E-02
4
5
8
4
6E-04
2
7E-01
8
\
Maximum
Estimated
Jfadjas JDose
mrem/yr ede
N/A
N/A
N/A
N/A
2E-01
1
N/A
N/A
N/A '
7E-01
7E-01
1. This value is estimated for a facility not yet designed or built. The highest dose from an operating^
facility was 7E-03 mrem/yr.
2. With 95 percent assurance, the 99.6th percentile. of the distribution of doses from these facilities
does not exceed 8 mrem/yr, where 8 mrem/yr ede is the highest dose estimated for all the facilities
in the sample. Radioiodines contributed a very small fraction to the effective dose equivalent of
the maximally exposed individuals.
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2. Description of Regulatory Programs
This chapter briefly summarizes the organizational and administrative controls
imposed by EPA and NRC on licensees for establishing emissions controls and for assuring
that emissions are not likely to increase in the future. A much more detailed description of
the NRC program can be found in Appendix A, and supplementary information is contained
in Appendices B, C, and E.
2.1 THE EPA'S REGULATORY PROGRAM UNDER THE CLEAN AIR ACT
2.1.1 Requirements
EPA regulations limit the effective dose equivalent (ede) to any member of the public
to 10 mrem/yr from all airborne radionuclides with no more than 3 mrem/yr from
radioiodine. EPA does not license facilities; instead, each facility is required to prepare an
annual report evaluating the doses from its emissions to the most exposed member of the
public. To minimize the burden on small users of radioisotopes, EPA does not require the
report to be filed with EPA if the estimated dose is less than 10 percent of the standard.
EPA has provided a number of methods for the user to demonstrate compliance with
the standard. They range from very simple to fairly complicated. They are all based upon
the methods developed by the National Council on Radiation Protection and Measurements
(NCRP).
2.1.2 Methods for Demonstrating Compliance
In 1986, the NCRP published Commentary No. 3, "Screening Techniques for
Determining Compliance with Environmental Standards," in response to a need indicated by
EPA for simple methods to assess compliance with the NESHAPs (NCRP86). Commentary
No. 3 Was revised in January 1989. EPA-approved methods for demonstrating compliance
with the NESHAPs are all based upon the January 1989 revision. In addition, EPA allows
the use of other methods for demonstrating compliance, provided they have been approved
by the Agency. -
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The EPA-approved methods form a tiered set of procedures, ranging from very
simple to moderately complex. They are intended to suit the needs of all types of facilities,
ranging from those with simple operations involving only small amounts of radioactivity to
those having complex operations involving large amounts of radioactivity.
The simplest procedures can be carried out using only a hand calculator; the most
complicated one requires a computer. All of the procedures have been put into the computer
program COMPLY, which is available from EPA. COMPLY has been designed to be user-
friendly and even at the highest level (the most complex method) requires a minimum amount
of input.
If the licensee is unable to demonstrate compliance using one of the simpler
procedures, the licensee is allowed to go to a more complicated one. If the licensee cannot
demonstrate compliance at the highest level, the licensee must report that fapt to EPA.
Facilities in compliance must file an annual report with EPA unless their estimated doses are
less than 10 percent of the limits. The EPA-approved compliance procedures are as follows:
• Level 1 - Possession Tables. This is the simplest method and is intended for
use by licensees who do not monitor their emissions. The licensee computes
the ratio of the annual amount of each radionuclide used to a standard value
for the radionuclide. The licensee then sums these ratios, and if the sum is
less than one, compliance with the dose standard is demonstrated.
• Level 1 - Concentration Tables. This simple method can be used by licensees
who measure their stack concentrations. The licensee computes the ratio of
the measured stack concentration of each radionuclide to a standard value for
that radionuclide. Compliance with the dose standard is demonstrated if the
sum of the ratios is less than one.
• Level 2. This corresponds to Screening Level 2 of NCKP Commentary No 3.
It requires such information as the release rate of each radionuclide, the
release height, the building dimensions, and the distance from the point of
release to the nearest receptor. It may also require some information about the
size of the stack or vent. If the release rates are not measured, EPA has
provided simple methods to estimate them. If desired, the licensee may supply
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the annual wind speed or use the default value of 2 meters/second. If the dose
is less than 10 mrem/yr from all radionuclides and 3 mrem/yr from
radioiodine, the licensee is in compliance with the dose standard.
• Level 3. This corresponds to Screening Level 3 of NCRP Commentary No. 3.~
In addition to the information needed at Level 2, it requires the user to supply
the distances to the nearest farms producing vegetables, milk, and/or meat. If
the dose is less than 10 mrem/yr from all radionuclides and 3 mrem/yr from
radioiodine, the licensee is in compliance with the dose standard.
j
• Level 4. This is the highest level. It is based upon the methods of NCRP
Commentary No. 3, but with some differences, the principal one being the
optional use of a wind rose. At the other levels, it is assumed that the wind
blows from the source toward the receptor 25 percent of the time; if the
licensee supplies a wind rose, the actual frequencies for the 16 sectors are used
along with the actual wind speed in each sector. The licensee must supply
distances to receptors in each of the 16 sectors. COMPLY then determines
which of these receptors receives the highest dose. If the dose is less than
10 mrem/yr from all radionuclides and 3 mrem/yr from radioiodine, the
licensee is in compliance with the dose standard.
Levels 1-3 are simple enough to carry out with a hand calculator; instructions are
contained in EPA89a. Level 4 must be carried out using the COMPLY code on an IBM-
compatible computer.
The COMPLY code considers four pathways of exposure: inhalation, ingestion,
immersion, and external exposure to surface contamination. Because it accounts for building
wake effects, it is suitable for close-in distances. At distances beyond the recirculation zone
near a building, it uses a modified Gaussian plume model. It accounts for decay and in-
growth of daughter radionuclides during transit from the release point to the receptor and the
farms, after being deposited on vegetation and soil, and after harvest, milking, or slaughter.
It also accounts for deposition of radioactivity upon food crops and forage and for uptake
from the soil. At Levels 1-3, these processes are handled by pathway factors developed by
the NCRP; at Level 4, the calculations are done explicitly.
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2.2 THE NRC'S REGULATORY PROGRAM UNDER THE ATOMIC ENERGY ACT
The regulatory programs established by NRC are intended to satisfy its statutory
obligations under the Atomic Energy Act of 1954, as amended, to protect the health and
safety of both workers and members of the public. NRC implements its programs either
directly through licensing and inspection of facilities, or through the Agreement State
program, in which NRC relinquishes its regulatory authority for most facilities to the states.
Agreement States perform the licensing and inspection functions.
Facilities are regulated under Chapter 1 of Title 10 of the Code of Federal
Regulations (10 CFR) and are licensed by NRC according to the type of radioactive material
that they use or possess and/or the type of activity in which they are engaged. The five
major types of licenses affected by Subpart I are: Parts 30-39 Licenses for specific uses of
byproduct material;2 Part 40 Licenses for source material (unenriched uranium or thorium);
Part 50 Licenses for production and utilization facilities (reactors and reprocessing plants);
Part 61 Licenses for land disposal of low-level radioactive wastes; and Part 70 Licenses for
special nuclear material (plutonium and enriched uranium). Licensees are subject to the
specific requirements established by the CFR part under which they are licensed and the
generally applicable requirements established in other parts of Chapter I of Title 10 of the
Code of Federal Regulations, such as 10 CFR 20. The vast majority of licenses are for
activities using byproduct material regulated under Parts 30-39.
The NRC's regulations limiting routine radionuclide airborne emissions are contained
in 10 CFR Part 20, Standards for Protection Against Radiation, which applies to all ,
licensees. For members of the public, the basic dose limit was recently amended to limit
individual exposures to 100 mrem/yr total effective dose equivalent (tede). The 100 mrem/yr
limit includes direct radiation and doses from both gaseous and liquid effluents. In addition,
recent amendments to Part 20 require that all licensees implement a radiation protection
2 Byproduct materials are man-made radioactive materials (except special nuclear material) produced or
made radioactive by exposure to the radiation incident to the process of producing or utilizing special nuclear
materials such as in a nuclear reactor. Byproduct material does include activation jproducts from nuclear
reactors and from plutortium-beryllium (Pu-Be) neutron sources, but does not include activation products from
other neutron sources such as Cf-252 or accelerators. Byproduct Material licenses are issued to educational
institutions, medical facilities, industrial facilities, and individuals for the possession and use of byproduct
materials and radionuclides for teaching, training, research and development, manufacturing, equipment
calibration, medical research and development, medical diagnosis and/or therapy.
2-4
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program that keeps exposures as low as is reasonably achievable (ALARA). Formerly, this
stated that other facilities "should" attempt to maintain exposures and effluents ALARA.
The revisions to Part 20 must be implemented by all licensees prior to January 1994.
In support of the Part 20 amendments, NRC issued draft Regulatory Guide DG-8013,
"ALARA Levels for Effluents from Materials Facilities," in October 1992. This document
provides guidance to licensees on designing and implementing an acceptable program for
establishing and maintaining ALARA levels for gaseous and liquid effluents at materials
facilities. The guide also states that, based on practical experience, an ALARA goal of about
10 millirem per year should be achievable by most licensees.
In addition to complying with the 10 CFR Part 20 limits, licensees must comply with
license conditions, which are often tailored to individual facilities. Also, licensees that are
part of the nuclear fuel cycle must comply with the FJPA standard established in 40 CFR Part
190, Fjivironmental Protection Standards for Nuclear Power Operations. Part 190 requires
that the doses to real individuals from all uranium fuel cycle sources, considering all gaseous
and liquid effluent pathways and direct radiation, not exceed 25 mrem/yr to the whole body
or any organ except the thyroid, for which the dose limit is set at 75 mrem/yr.
The NRC's licensing program can be best understood as a "tiered" or "graduated"
program based on the potential hazards associated with the types and quantities of radioactive
materials used and the activities authorized. The greater the potential hazard, the more
stringent the requirements. In general, the licensing procedures require the applicant for a
license to:
• list the activity or activities for which a license is sought;
• identify the facility or portions of the facility where the licensed materials will
be used, including a description of all engineered controls;
• identify the training and qualifications of the persons authorized to use the
material, and/or of the radiation safety officer designated to oversee licensed
activities; ,
• describe the procedural controls to be employed to assure containment and
physical protection of the radioactive materials;
• establish the limiting conditions for operations; and
• implement confirmatory monitoring and/or radiation surveys.
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The degree of specificity in the license application, the extent of application review,
and/the extent of license conditions imposed are all related to the potential hazards associated
with the activity. Fuel cycle and other "large" facilities must meet the most stringent
requirements (NRC does not define "large" licensees, but in general, large licensees are
those required to submit the data needed to prepare an Environmental Impact Statement or
Assessment at the time of license application or renewal). Other licensees (predominately
research and medical facilities holding byproduct licenses issued under Parts 30-39) must
meet somewhat less detailed,obligations but must still provide the basic information listed
above.
In the case of these "other" licensees, where the potential for airborne releases of
radioactive materials is small, continuous effluent monitoring requirements are usually not
imposed, but periodic confirmatory measurements must be made. If the potential for releases
is more substantial, requirements will include both stack monitoring and confirmatory
environmental sampling and analysis. The recent amendments to Part 20 include
requirements detailing how licensees are to demonstrate compliance with the annual dose
limit. These amendments also require that all licensees retain the records needed to confirm
that dose limits have not been exceeded until the license is terminated. Periodic onsite
inspections are conducted to confirm that the licensee has operated the facility in full
compliance with the applicable regulations and license conditions. For byproduct material
licensees using non-sealed sources, inspections are conducted approximately every one to
seven years, depending on the quantity of material possessed, the type of activity conducted,
and the priority assigned by NRC. Priority seven licensees receive an initial inspection and
are only inspected again if a particular problem arises.
2.3 COMPARISON OF THE NRC'S REQUIREMENTS WITH THE NESHAP
The NESHAP established in 40 CFR 61, Subpart I, requires NRC-licensed facilities
to determine compliance with the 10 mrem/yr (no more than 3 mrem/yr from radioiodines)
dose limit annually. Facilities are required to maintain records of their calculations and
supporting data; if the calculated doses exceed 10 percent of the standard, they must file an
annual report with EPA. Facilities seeking to build a new source must prepare and submit
an application for construction approval if the estimated doses from the source equal or
exceed 10 percent of the standard. Facilities seeking to modify an existing source must
prepare and submit an application for construction approval if the doses from the proposed
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modification are equal ito or greater than 1 percent of the standard or if doses from the entire
facility, including the modification, are equal to or greater than 10 percent of the standard.
Because the Designated Survey (Chapter 3) and the Random Survey (Chapter 4)
evaluate facilities whose operations are restricted by the pre-existing 10 CFR 20, it is
appropriate to compare the NESHAP to the pre-existing standard as well as the revised
version. Table 2-1 compares the NRC's requirements for both fuel cycle and other large
facilities and other licensees and the requirements of the NESHAP.
As detailed in Appendix A, the pre-revision Part 20 required large licensees to
develop and report extensive data on their effluent releases and to be subject to extensive
confirmatory inspections. However, of the approximately 6,000 facilities in the study
population, only a tiny fraction are large licensees. In addition to 150 fuel cycle facilities,
there are perhaps another 50 large materials licensees. Other NRC licensees are not required
to estimate doses to members of the public, nor are they required to calculate routinely and
report their compliance with the derived air concentrations (DACs), formerly called
maximum permissible concentrations (MPCs), of radionuclides for unrestricted areas.
However, these small licensees do develop and maintain most of the data needed to
determine compliance with the limits imposed by the NESHAP and to prepare an application
for approval to construct or modify.
Discussions with personnel at medical and research facilities indicate that seldom, if
ever, will an applicant propose DACs greater than those established by 10 CFR Part 20,106.
Thus, the DACs for unrestricted areas are the de facto limits for these licensees. However,
because licensees typically assess these concentrations in the stack, the concentrations in open
t - ,
areas are lower due to dispersion.
This notwithstanding, the basic limits imposed by NRC via either the old or new Part
20 are less restrictive than those imposed by the EPA's NESHAP. Other differences
between the NESHAP and the pre-existing Part 20 are primarily due to two differences in the
methodologies used by NRC and EPA to estimate dose. The first is that the NRC's MPCs
are based primarily on the inhalation pathway. By contrast, EPA concentrations consider
doses received via the immersion, inhalation, ingestion, and ground-surface pathways. The
second difference is that NRC's MPCs are based on ICRPII recommendations, and EPA's
are based on ICRP 26 and 30. With the revision to Part 20, the differences between NRC
and EPA are lessened since the new Part 20 uses ICRP 26 and 30 methodology.
2-7
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Table 2-1. Summary of regulatory requirements.
Regulatory
Activity
Licensing or
Approval
Dose Limit
Records
Reports
Inspections
Enforcement
NRC Requirements for
Large NRC Licensees
Environmental report,
safety analysis report,
ALARA design
review, technical
specifications
Per technical
specifications. For
fuel cycle facilities 25
mrem/y whole body or
any organ (75 mrem/y
thyroid).
Results of surveys,
effluent monitoring,
environmental
measurements, dose
calculations for 40
CFR 190 compliance.
Quarterly or semi-
annual source terms,
and environmental
monitoring results,
annual dose report for
40 CFR 190
compliance.
Annual or resident
inspectors, follow-up
on previous violations.
Five violation levels
based on safety-
implications, corrective
actions, fines, orders,
license revocation;
citizens may petition
NRC to enforce, but if
the EDO does not
agree, no action is
taken.
NRC Cutreat Pat* 20
for Other Licensees:
Facility design
handling/use
procedures, possession
limits.
Per license condition
or limits in 20.105 &
MFCs in 20.106
Results of surveys,
material receipts,
ventilation rates.
Exposures or releases
greater man 10 times
20.105 or 20.106.
Once, or once every 1
to 7 years, -depending
on type of license and
activities conducted.
Same as for large
facilities.
EPA's NBS8AP
Facility design, effluent
controls, quantities of
material by chemical &
physical form, 'dose
estimate, but only if _>_
10 percent of limits
10 mrem/y, not more
man 3 mrem/y due to
radioiodines.
Effluent monitoring
data or annual
possession of materials
data used to determine
compliance.
Annual dose
calculations if greater
man 10 percent of
limits.
Under development.
Monthly reports for
facilities not in
compliance; citizens
may take legal actions
(CAA, Section 304) to
compel compliance.
NRC' Revised Part 20
No change.
100 mrem/y total ede
to any member of the
public. Doses from
direct radiation, liquid
and gaseous effluents
must be counted.
Dose rate must be less
man 2 mrem/hr.
Licensees subject to 40
CFR 190 must comply
with mat standard in
addition to NRC
limits.
All licensees must
retain records needed
to demonstrate
compliance with dose
limits until license is
terminated.
As before, except any
exceedence of dose
limits must be reported
within 30 days.
No change.
No change.
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2.4 NRG-IICENSED FACILITY PROGRAM ANALYSIS
The NRC's programs for "Fuel Cycle and Other Large Facilities" provide regulations
to limit airborne radionuclide emissions to the atmosphere. As described in Appendix A, the
NRC's requirements for facility design, environmental impact assessment, and safety analysis
(10 CFR 30, 32, 33, 35, 39, 40, 50, 70), together with a comprehensive enforcement and
inspection program, provide reasonable assurance for the protection of the public.
In reviewing the regulatory program for "other facilities," several observations are
notable. First, other facility licensees, although required to evaluate their compliance with
the DACs established by the revised 10 CFR Part 20 (Appendix B to 10 CFR 20.1001-
20.2401), need not submit their calculations to NRC for review, even at the time of initial
application. Second, ALARA requirements apply to all licensees and to emissions from the
facility as well as to workers. Third, with respect to releases of radioactive materials, the
only reporting requirement imposed on these licensees is to notify NRC if concentrations
exceed 10 times the allowable DACs.
Although the monitoring and inspection process for these facilities are relatively
infrequent, NRC requires the facility to keep the concentrations of radioactive materials in
effluent air at or below the levels of the DACs. The NRC or Agreement States often
recommend that the license applicant use a more conservative approach in calculating
potential airborne effluent concentrations released in the exhaust system or at the stack. In
general, a "10 percent at the stack" rule is recommended as the starting point of the
estimation (NRC84a). This approach lowers the total effective dose equivalent to individual
members of the public residing close to the institution. In addition, it reduces the potential
of exceeding the regulatory limits set forth in Table n of Appendix B to 10 CFR Part 20
even in the event of minor operational errors.
x The airborne effluent concentrations at the release point of the emission are used to
estimate the total effective dose equivalent to the public at the receptor locations which are
farther away (ranging from several hundred feet to several miles). However, the estimation
does not usually take into account the effluent dispersion and dilution factors in the
atmosphere. These factors will make the dose lower. On the other hand, the NRC's DACs
only calculate the dose from inhalation and immersion and do not take into account the dose
from ingestion or ground deposition. In some cases, these may be the major pathways.
2-9
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The NEC's and the Agreement States' regulatory programs control the use of
radioactive byproduct material. The programs also provide a regulatory mechanism to limit
airborne radionuclide emissions to the atmosphere from research and development facilities,
manufacturing facilities, and medical institutions. However, because effluents are not
actually measured by stack instruments, NRC must rely on the licensees' administrative
programs for assurance that concentrations of radioactive materials in effluent air do not
exceed the levels of the DACs.
Given the lack of monitoring requirements for these facilities, the lack of guidance on
appropriate assumptions for releases of materials that are not handled as gases or aerosols,
and the infrequent inspections at these facilities, EPA decided to conduct an analysis of the
doses caused by these facilities to judge the adequacy of the NRC's program.
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3. Results of Designated Survey of NRC-Licensed Facilities
This chapter updates the emissions and doses from a small group of actual NRC-
licensed facilities (those included in the Designated Survey) that are currently subject to the
Subpart I NESHAP. These facilities belong to the following source categories: uranium fuel
cycle (Section 3.1), test and research reactors (Section 3.2), radiopharmaceutical and
radiolabeled compound manufacturers (Section 3.3), large hospitals and medical research
facilities (Section 3.4), manufacturers of sealed sources (Section 3.5), depleted uranium
munitions test sites (Section 3.6), rare earth and thorium processors (Section 3.7), and
commercial low-level radioactive waste disposal and incineration facilities (Section 3.8).
Most of the facilities in the Designated Survey were analyzed by EPA (EPA73a, EPA73b,
EPA78, EPA79, EPA82, EPA83, EPA84, EPA86, EPA89) prior to the reconsideration
period. In this study, several of the source categories are evaluated in greater detail than in
previous studies.
The facilities in the Designated Survey were selected based on expert opinion that
they had the greatest potential for causing the highest doses. It was believed that if the
evaluation of these facilities demonstrates that the public health and safety is protected with
an ample margin, the same can be concluded about smaller facilities. To be certain that it
has identified those facilities causing the greatest dose to a member of the general public,
EPA designed and conducted the Random Survey, the results of which are presented in
Chapter 4. Appendix D describes in more detail the types of facilities evaluated in the
random survey.
The Designated Survey updates previous analyses to improve the accounting for
(a) the wide diversity of facilities, (b) the limitations in the available database, and (c) the
limitations of dispersion models for evaluating certain facilities. This analysis draws upon
and updates previous evaluations and incorporates revisions to the estimates based on new
information developed during the public comment period. Because this BID draws upon all
past work and provides new information, it represents EPA's most recent and comprehensive
analysis of the doses caused by these facilities.
Current information used in evaluating doses was obtained through research of
licensee dockets contained in the NRC Public Document Room (PDR); responses to formal,
written questionnaires; EPA studies conducted since EPA89 (EPA91, EPA92); and telephone
3-1
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interviews with the licensees. In all cases, available facility information was reviewed to
ascertain the potential for significant airborne emissions. Potential doses were evaluated for
activities where potential existed for a facility to exceed the Subpart I NESHAP dose limits.
The results demonstrate that all NRC licensees examined as part of the Designated Survey
are currently meeting the standard. The dose estimates are summarized in Section 3.9.
For each source category, this chapter presents the results of prior studies (up through
and including .the 1989 NESHAP studies), studies that have been undertaken since the 1989
NESHAP studies but before this study, and the results of this study, the Designated Survey.
3.1 URANIUM FUEL CYCLE FACILITIES
Uranium fuel cycle facilities consist of: mills that extract uranium from ore and their
accompanying tailings piles; conversion facilities that chemically convert uranium feed from
the mills (yellowcake) to uranium hexafluoride; the enrichment plants (owned by DOE but
not regulated by NRC) that enrich uranium in the uranium-235 isotope; fuel fabrication
facilities that convert uranium from the hexafluoride to an oxide form, pelletize the uranium,
and incorporate it into fuel rods for power reactors; pressurized-water and boiling-water
power reactors; spent reactor fuel storage and disposal facilities; and, although none are
currently operating or envisioned, fuel reprocessing plants that recover residual fissile
material (uranium and plutonium) from spent fuel. This section presents the results of the
current evaluation of airborne emissions from uranium mills, uranium hexafluoride
conversion facilities, light water reactor fuel fabricators, and spent fuel storage. Dose
estimates are presented for each source category analyzed and are summarized in Section
3.9. Emissions from power reactors are covered in another analysis for a separate
rulemaking (EPA91).
3.1.1 Uranium Mill Tailings
Uranium mills extract uranium from ores which contain only 0.01 to 0.3 percent
U3O8. The product of the mills is shipped to conversion plants where it is converted to
volatile uranium hexafluoride (UF6) which is used as feed to uranium enrichment plants.
Emissions of radon from this process are regulated by separate NESHAP standards,
Subpart T for disposal of tailings and Subpart W for operating mill tailings.
3-2
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3.1.1.1 Previous Evaluations. EPA's most recent analysis (EPA89) of uranium mills
focused on mills with dry tailings piles that were either operating Or on standby. The study
also included analysis of a generic model mill to assess the dose and risk from tailings piles
at mills that are either decommissioned or undergoing decommissioning. The maximum dose
calculated for an operating mill was for the Homestake Mill. The dose from process exhaust
was 12.8 mrem/yr ede and the dose from the tailings pile was 0.95 mrem/yr ede.3 The dose
obtained for the model mill's tailings pile was 25.8 mrem/yr effective dose equivalent (ede).
3.1.1.2 Evaluations of Specific Facilities Made During the Reconsideration Period. Since
the dose reported in EPA89 for the model mill's failings exceeded the Subpart I NESHAP
dose limits and since the schedule for remediation of mill sites may change, EPA decided to
look at all mills with exposed piles. EPA, NRC, and affected Agreement States have entered
into a Memorandum of Understanding (MOU) (56 FR 67564) addressing the schedule for
remediation of non-operational tailings piles. The objective of the MOU is to assure the
installation of an earthen cover at all current disposal sites by the end of 1997, or within
7 years of when, the existing operating and standby sites enter disposal status.
Doses from mill process exhausts have not been re-evaluated because Homestake Mill
has ceased operations, and the dose from all other operating mills evaluated in EPA89 were
all less than 0.3 mrem/yr ede which is below the Subpart I NESHAP dose limit.
Table 3-1 lists all NRC-licensed mill sites that currently have exposed tailings. This
evaluation utilizes the most recent information on dry tailings areas and radium-226
concentrations. These data were obtained from the NRC Public Document Room, from
NRC's Uranium Recovery Field Office, and from conversations with cognizant personnel in
EPA's Regions 6 (Phil Shaver), 8 (Ed Kray), and 10 (Leo Wainehouse) between July and
August 1991 ^ Demographic and meteorological data were taken from EPA89. Based on the
demographic data, assumptions were made concerning the placement of farms. These
assumptions are consistent with those made in the Random Survey portion of the study.
3 The dose of 12.8 mrem/yr was estimated prior to Homestake's commitment to install yellowcake drying
and packaging scrubbers. Given a decontamination factor of 10 for scrubbers, the prospects were good for
future emissions to be below the NESHAP limit of 10 mrem/yr. Homestake has since ceased operations and is
, being decommissioned.
3-3
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Source Term Determination
EPA derived airborne source terms for exposed tailings using site meteorology taken
from EPA89 and the methodology suggested by NRC in Regulatory Guide 3.59 (NRC87).
Table 3-1 presents these source terms. Thorium-230 is assumed to be in equilibrium with
radium-226, lead-210, and polonium-210. Uranium-238 is assumed to be in equilibrium with
uranium-234.
Meteorologic. Demographic, and Agricultural Data
Table 3-1 presents the source of meteorological data used as input to the calculations
and the distances to the nearest residents that were used as input to COMPLY^ These data
were taken froin EPA89. The stability array meteorological data were converted to wind
roses for use by the COMPLY code.
Demographic data were taken from EPA89. If these data placed the nearest resident
within 2,000 m of the site, vegetables were assumed to be grown at home. Otherwise, the
distance to the vegetable farm used for the dose analysis was 2,000 m. Meat- and milk-
producing farms were placed at 2,000 m. These assumptions were used to maintain
consistency with the Random Survey portion of this study.
3.1.1.3 Results of the Designated Survey for Uranium Mill Tailings. The results show that,
using updated estimates of windblown releases from dry tailings piles, the maximum ede
calculated using COMPLY is 2 mrem/yr. This dose is primarily from the inhalation and
irigestion pathways. This dose is calculated for the resident exposed to the highest offsite
concentration around the Petrotomics facility in Medicine Bow, Wyoming. The results for
other facilities with dry tailings piles range from 0.008 to 1 mrem/yr ede. Results for all
evaluated facilities are presented in Table 3-16, located at the end of this chapter.
3.1.2 Uranium Conversion Facilities :
A uranium conversion facility converts uranium oxide (U3O8 or yellowcake) to
purified uranium hexafluoride (UFe)^ Uranium hexafluoride, which is volatile at slightly
elevated temperatures, is the chemical form in which uranium enters the enrichment plant.
3-5
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3.1.2.1 Previous Evaluations. Currently, two commercial uranium hexafluoride (UF6)
production facilities are operating in the United States, the Allied Chemical Corporation
(Allied-Signal) facility at Metropolis, Illinois, and the General Atomics facility in Sequoyah,
Oklahoma (formerly owned by Kerr-McGee Nuclear Corporation). Both facilities were
evaluated in EPA89. The doses calculated for the Sequoyah and Metropolis facilities were
3.6 and 2.2 mrem/yr ede, respectively.
3.1.2.2 Evaluations of Specific Facilities Made During the Reconsideration Period. Both
the Sequoyah and Metropolis uranium hexafluoride production facilities were included in the
Designated Survey. In support of this evaluation, the licensees supplied information on the
location of the closest receptor in each of 16 compass directions and the distance to the
nearest vegetable-, meat-, and milk-producing farms. All other parameters used in this study
are the same as those used in EPA89.
Source Term Determination
Source terms and solubility classes used in this study and in EPA89 are averages of
the measured releases for each facility for 1984 through 1987. These data, which were
originally reported in semi-annual environmental monitoring reports to NRC, are presented in
Table 3-2.
The plant parameters used in this study and originally in EPA89 were taken from
NRC84 and NRC85b. For Allied-Signal, the stack height used, 24 m, is an average of all
release points for that plant. The same stack height was used for the Sequoyah facility. A
stack diameter of 0.16 m was used for both facilities.
Meteorologic. Demographic, and Agricultural Data
The stability array meteorological data used in EPA89 were converted to wind roses
for use by the COMPLY code. •
Site-specific demographic data locating the closest receptor in each of 16 directions
were obtained from the licensee for each facility. The nearest individual at both facilities is
assumed to produce vegetables at home. In both cases, the nearest milk-producing farm is
located at greater than 2,000 m. Therefore, to be consistent with the assumptions used in the
Random. Survey study, both milk-producing farms were placed at 2,000 m.
3-6
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Table 3-2. Atmospheric radioactive emissions assumed for reference dry and wet
process uranium conversion facilities
Facility
Radionacltde
Emissions
Solubility Class
W
Reference
Allied Corp.
Metropolis, EL
U-Natural2
Tli-2302
Ra-2262
0.10000
0.00050
0.00001
56
0
0
30
0
100
14
100
0
NRC84
Sequoyah. Fuels
Sequoyah, OK
U-Natutal3
Th-2303
Ra-2263
0.050
0.005
0.005
65
0
0
5
0
100
30
100
0
NRC85b
1.
2.
3.
Solubility classes D, W, and.Y refer to the retention of inhaled radionuclides in the lungs;
representative half-times for retention are less than 10 days for class D, 10-100 days for class
W, and greater than 100 days for class Y.
Particle size 3.4 /im.
Particle size f/uri)
4.2 to 10.2
2.1 to 4.2
1.3 to 2.1
0.69 to 1.3
0.39 to 0.69
0.00 to 0.39
% (Average: 1980-1984->
9.3
9.7
5.5
6.5
13.5
55.3
Data taken from NUREG-1157 (NRC85b).
The nearest meat-producing farm is located more than 2,000 m from the Allied-Signal
facility. Therefore, the meat-producing farm was placed at 2,000 m for the COMPLY
analysis. However;, Sequoyah Fuels maintains a "stacker operation" in which cattle are
rotated through different pastures to achieve a desired weight gain prior to being shipped to a
feed lot. The nearest pasture used in this stacker operation is located 244 m from the nearest
plant stack.
3.1.2.3 Results of the Designated Survey for the Uranium UF6 Conversion Facilities. The
maximum ede calculated using COMPLY and current detailed demographic data is
7 mrem/yr for the Allied-Signal wet process uranium conversion facility. This dose is
primarily from the inhalation pathway. The maximum ede calculated for the dry process
uranium conversion facility (Sequoyah Fuels Corporation) is 3 mrem/yr from the inhalation
and ingestion pathways. In both cases, the most exposed individual is a resident located
approximately 700 m from the facility.
3-7
-------
3.1.3 Fuel Fabrication Facilities
There are two basic types of fuel fabrication plants: those that produce fuel
assemblies for light water reactors and those that produce fuel assemblies for test and
research reactors. In either case, the raw material is pelletized, encased with metal, and
formed into assemblies. .
3.1.3.1 Previous Evaluations. .l
Non-Light Water Reactor (LWR) Fuel Fabrication Facilities. None of the facilities in
this category were estimated to cause doses greater than 1 mrem/yr ede to nearby individuals
(EPA89).
LWR Fuel Fabrication Facilities. Table 3-3 lists the seven licensed uranium fuel
fabrication facilities in the United States that fabricate commercial LWR fuel. Of the seven,
only five had active operating licenses as of January 1, 1988. Of those five facilities, two
use enriched uranium hexafiuoride to produce completed fuel assemblies and two use
uranium dioxide. The other facility converts UF6 to UO2 and recovers scrap materials
generated in the various processes of the plant.
In EPA89, the site characteristics used in the assessment of the reference fuel
fabrication facility were drawn from a combination of the Westinghouse (Columbia, South
Carolina) and General Electric (Wilmington, North Carolina) facilities. This is appropriate
since all phases of fuel fabrication (i.e., both ammonium diuranate wet process and direct-
conversion dry process conversion of UF6 to UO2, mechanical fabrication of fuel assemblies,
and scrap recovery) take place at these sites. The dose calculated for this model fuel
fabrication facility was 0.27 mrem/yr ede.
3.1.3.2 Evaluations of Specific Facilities Made During the Reconsideration Period.
Non-Light Water Reactor (LWR) Fuel Fabrication Facilities. For non-LWR fuel
fabricators, the doses were found to be very low (EPA89). Consequently, evaluations of
these facilities were not updated. -
LWR Fuel Fabrication Facilities. The EPA89 study emissions data were developed
so that the model fuel fabrication facility assessed would represent the bounding case for
3-8
-------
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•8
13
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If
W OJ
1
s
I
t
o
p
I
(!) O
61
I
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fill!
s o a- n
lllfl
i«"r^ Ev /^
a. 3
2 §
1S1 3
1 = -2B'
U o g &i
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.a
J6
3-9
-------
LWR fuel fabricators. However, past evaluations of the "worst case" model facility did not
utilize detailed close-in, site-specific demographic data. During the reconsideration period,
EPA obtained and used updated demographics that located the closest receptor in each of 16
compass directions for the Westinghouse fuel fabrication facility. The distance to the nearest
vegetable-, meat-, and milk-producing farms was also obtained as part of the Designated
Survey. AU omer data utilized in this study were taken from EPA89.
Source Term Determination
*
Table 3-4 presents reported uranium effluents from 1983 through 1987 for each of the
fuel fabrication facilities with current operating licenses. These data, taken from EPA89,
were originally reported in the semi-annual environmental monitoring reports submitted by
the facilities to NRC. The data in Table 3-4 show that the Westinghouse and General
Electric facilities have releases 10 to 100 times those of the Babcock and Wilcox and
Combustion Engineering facilities. This is expected because the Westinghouse and General
Electric plants produce substantially more fuel and start the process with uranium
hexafluoride, while the other two facilities begin the fuel fabrication process with UO2.
The atmospheric radioactive emissions estimated to be released each year by the
reference fuel fabrication facility analyzed in EPA89 are presented in Table 3-5. With the
exception of uranium-236, these values represent the geometric mean of the reported effluent
releases for the Westinghouse fuel fabrication facility for 1983 through 1987. The geometric
mean best represents the radioactive emissions, since the sample distribution is lognormal.
The value for uranium-236 is based on release data for, 1983 through 1987 as reported
in the semi-annual environmental monitoring reports submitted to NRC by the General
Electric facility at Wilmington, North Carolina. The effluent release height used in this
analysis is 10 m (EPA89).
* , " •'
Meteorologic. Demographic, and Agricultural Information
The climatological data used originally in EPA89 are based on measurements taken at
the U.S. Weather Bureau Station at Columbk Metropolitan Airport in South Carolina
(NRC85a). Sets of hourly meteorological data obtained from the airport for 1984 through
1986 were used to develop wind frequency distributions for stability classes A through F.
Those same stability arrays were converted to a wind rose for use with the COMPLY code.
3-10
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Table 3-4. Light water reactor commercial fuel fabrication facilities reported annual
uranium effluent releases for 1983 through 1987 in /iCi/yr.(1)
licensee
Babcock and Wilcox
Lynchburg, VA
SNM-116
70-1201
Combustion Engineering
Windsor, CT.
SNM-1067
70-1100
Combustion Eng
Hematite, MO
SNM-33
70-36
General Electric
Wilmington, NC
SNM-1097
70-1113
Westinghouse
Columbia, SC
SNM-1107
70-1151
.Year
1983
1984
1985
1986
1987
1983
1984
1985
1986
1987
1983
1984
1985
1986
1987
1983
1984
1985
1986
1987
1983
1984
1985
1986
1987
U-234
4.7E+00
5.6E+00
4.6E+00
5.7E+00
3.9E+00
NA®
NA
NA
NA
NA
NA
NA
NA
NA
NA
3.1E+02
4.0E+02
4.1E+02
1.2E+03
1.6E+02
1.2E+03
1.5E+03
1.2E+03
1.1E+03
l.OE+03
U-235
2.1E-01
2.5E-01
2.1E-01
2.5E-01
L7E-01
NA.
NA
NA
NA
NA'
NA
NA
NA
NA
NA
2.0E+01
2.6E+01
2.7E+01
7.1E+01
l.OE+01
5.3E+01
1.2E+02
7.2E+01
5.3E+01
5.6E+01
U-23&
2.1E-02
2-3E-02
2.1E-02
2.6E-02
1.7E-02
NA
NA
NA
NA
NA
NA
NA
NA
NA
NA .
4.5E+02
5.7E+00
5.7E+00
1.6E+01
2.0E+00
NR(*
NR
NR
NR
NR
,U-23S
1.1E+00
1.3E+00
1.1E+00
1.3E+00
9.1E-01
NA
NA
NA
NA -
NA
NA
NA
NA
NA
NA
1.3E+02
1.7E+02
1.5E+02
3.5E+02
5.6E+01
2.5E+02
3.2E+02
3.1E+02
3.4E+02
3.1E+02
Total
6.0E+00
7.2E+00
5.9E+00
7.3E+00
5.0E+00
3.9E+01
2.7E+01
4.9E+01
5.5E+01
4.7E+01
2.1E+02
4.2E+01
7.3E+01
6.7E+02
2.8E+02 -
4.6E+02
6.0E+02
5.9E+02
1.6E+03 .
2.3E+02®
1.5E+03
1.9E+03
1.6E+03
1.5E+03
1.4E+03
1. Taken from semi-annual licensee environmental monitoring reports submitted to NRC.
2. Not available; only total curies of uranium released reported to NRC.
3. Release data for the second half of 1987 were not available but were assumed to be the
same as first half s. ,
4. NR denotes not reported. Values are small and not included in total.
Table 3-5. Atmospheric radioactive emissions assumptions for
reference fuel fabrication facility.
ItadionftcBde '
U-234
U-235
U-236
U-238
< Eatfssjtofts (Ci/yr)
1.2E-O3
6.7E-05
1.6E-O5
3.0E-04
3-11
-------
Site-specific demography locating the closest receptor in each of 16 directions and the
distance to the nearest vegetable-, meat-, and milk-producing farms was obtained from the
licensee for the Westinghouse facility. The nearest vegetable-producing farm is located
240 m from the source. Milk- and meat-producing farms are located more than 2,000 m
from the stack. To be consistent with assumptions used for the Random Survey, residents
were assumed to grow all their vegetables at home, and meat- and milk- producing farms
were placed at 2,000 m for this analysis.
3.1.3.3 Results of the Designated Survey for Fuel Fabrication Facilities. The maximum ede
calculated using COMPLY and current detailed demographic data for the Westinghouse
CNFD fuel fabrication facility in Columbia, South Carolina, is 0.06 mrem/yr. This dose is
primarily from the inhalation pathway. The dose occurs to a resident located approximately
1,000 m from the facility.
3.1.4 Interim Spent Fuel Storage Facilities
The only commercial spent fuel storage facility licensed in the United States is the
General Electric facility in Morris, Illinois. It is currently operating. However, the vast
majority of spent fuel is stored at nuclear power reactor sites.
Interim spent fuel storage facilities were not examined separately in past evaluations
but were included in the evaluation of power reactors (EPA89, EPA91). All reactor sites
have wet pool storage capability, and some have additional out-of-pool capacity. EPA89
found that the overall emissions from power reactors, of which spent fuel storage was one of
four sources of emissions, were well within regulatory limits. A more recent EPA study
(EPA91) also found that total airborne emissions from reactor sites are very low, causing
doses of less than 1 mrem/yr ede to the most exposed individual. On this basis, EPA
concludes that a separate evaluation of the Morris facility is not necessary.
3.2 TEST AND RESEARCH REACTORS
As of August 1988, there were 76 non-power research and 8 test reactors licensed by
NRC in the United States (NRC88a). -
3-12
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The majority of the research reactors are located at universities where they are used
for teaching and research: to study reactor designs, to conduct research on the effects of
radiation on materials, and to produce radioactive materials used by sealed source and
radiopharmaceutical manufacturers. Approximately 37 percent of these are of the TRIGA
design. These reactors have thermal power levels ranging from essentially zero to 10,000
kilowatts.
Table 3-6 lists the NRC docket number, thermal power level, location, and present
licensing status of the eight test reactors. Two are operational. The remainder are in safe
storage... Their thermal power levels range from 6 to 60 megawatts thermal (Mwt).
Table 3-6. Licensed test reactors in the United States as of August 1991.1
NRC ,
^Docket
Nbv ^
50-22
50-30
50-70
50-146
50-184
50-183
50-200
50-231
Test Reactor Name
Westinghouse
NASA Plum Brook
General Electric
Saxton PWR
MBS
GEEVESR
Exp. Superheat
B&WBAWTR
SEFOR Sodium Cooled
Ifeerntal
^Pcwer(Mw)
60
60
50
28
10
17
6
20
Location
Waltz Mill, PA
Sandusky, OH
Alameda County, CA
Saxton, PA
Gaithersburg, MD
Alameda County, CA
Lynchburg, VA
.1 • ,
Striclder, AR
Present Status
Safe Storage
Dismantling Order
Issued May 26,
1981
Operational
(currently shut
down)
Safe Storage
Operational
Safe Storage
Safe Storage (NRC)
Safe Storage (State)
1. List taken from NRC82; status verified August 1991.
3.2.1 Previous Evaluations
Previous evaluations (EPA79? EPA84, EPA89) show that the emissions from these
facilities are a function of power level and duty cycle.
3-13
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In EPA89, doses resulting from test and research reactors were bounded on the basis
of the four actual reactors with the largest emissions as identified by Corbit (Co83). These
included three university research reactors (Massachusetts Institute of Technology, University
of Missouri, and University of Rhode Island) and one government test reactor (the National
Bureau of Standards4). Emissions data from Corbit were supplemented by information
presented in the facilities' annual operating reports (e.g., MIT87). The principal nuclide
emitted is argon-41. Tritium is also emitted, although in lesser amounts. The emissions
result in a maximum estimated dose of 0.7 mrem/yr ede (EPA89).
3.2.2 Evaluations of Specific Facilities Made During the Reconsideration Period
Of the four reactors that were evaluated in EPA89, only three are currently
operational (Massachusetts Institute of Technology [MIT], University of Missouri, and the
National Institute of Standards and Technology [MIST] reactors). These three remaining
reactors were included in the Designated Survey. As part of the revaluation, detailed
demographics were obtained from the licensees. All other parameters used were taken from
the EPA89 assessment.
Source Term Determination
The current study used the same effluent release data as EPA89. These data are
shown in Table 3-7.
Table 3-7. Effluent release rates (Ci/yr) for test and research reactors.
Facility
University of Missouri
National Institute of Standards & Technology
Massachusetts Institute of Technology
Radionuelide
H-3
1.6E+01
1.6E+02
-
Ar-41 ,
2.5E+03
4.7E+02
4.2E+03
4 NBS is now known as the National Institute of Science and Technology.
3-14
-------
The effluent releases occur from stacks 33 m, 33 m, and 50 m high, respectively, for
the University of Missouri, NIST, .and MIT reactors. ,
Meteorologic, Demographic, and Agricultural Data
s. -
As part of the Designated Survey, site demographic data used for the assessments
presented in EPA89 were updated to incorporate information obtained from the licensees on
the distance to the closest receptors in each of 16 directions. The distance to the nearest
meat, milk, and vegetable farms was also obtained.
The meteorological data used in this study for the University of Missouri, NIST, and
MET reactors are for Columbia, Missouri; Fort Meade, Maryland; and Boston,
Massachusetts; respectively (EPA89). For this study, the stability array data used in EPA89
were converted to wind roses for use with the COMPLY code.
Based on the COMPLY run for the University of Missouri, the receptor exposed to
the highest concentration is a resident located approximately 700 m from the source. For
NIST, the receptor exposed to the highest concentration is also a resident, in this case
approximately 480. m distant. The COMPLY run using detailed demography showed that,
for MIT, the receptor exposed to the highest concentration is a nonresident. This individual
is located approximately 100 m from the source.
Agricultural data obtained from the University of Missouri indicated that a vegetable-
producing farm is located 600 m from the source. The vegetable farm was placed at this
location for this study. No milk- or meat-producing farms were reported within 2,000 m of
the reactor. Therefore, in order to maintain consistency with the Random Survey
assumptions, the milk and meat farms were.placed at 2,000 m.
Agricultural data supplied by NIST and MIT indicated no farms within 2,000 m of
either reactor. To be consistent with the Random Survey assumptions, the vegetable, milk,
and meat farms were placed at 2,000 m.
3-15
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3.2.3 Results of the Designated Survey of Test and Research Reactors
The immersion pathway is the dominant contributor to the dose for all three facilities.
The maximum ede calculated using COMPLY and current detailed demographic data is
4 mrem/yr. This dose is calculated for the individual exposed to the highest offsite
concentration around the Massachusetts Institute of Technology research reactor. This dose
is to a nonresident in an office; therefore, an occupancy factor of 0.3 was applied. The
value of 0.3 is based upon 10 hours per day, 5 days per week, 52 weeks per year
(10x5x52/8760=0.3).
The ede calculated for the University of Missouri research reactor is 2 mrem/yr. The
ede calculated for the receptor exposed to the highest offsite concentration around the
National Institute of Standards and Technology test reactor is 0.8 mrem/yr. In both cases,
the dose is to an offsite resident. Refer to Section 3.9 for a summary of all dose estimates.
3.3 RADIOPHARMACEUTICAL AND RADIOLABELED COMPOUND
MANUFACTURERS
Of the approximately 120 radiopharmaceutical suppliers, distributors, and nuclear
pharmacies (Ce81), 15 are large firms. These firms handle large amounts of radionuclides in
hot cells, while smaller firms change the chemical form of the nuclides, and the pharmacies
repackage the material into convenient amounts.
3.3.1 Previous Evaluations
The four largest firms (DuPont Boston, DuPont Billerica, Amersham, and Cintechem)
were previously evaluated (EPA89). The maximum dose to nearby individuals was estimated
to be 9 mrem/yr ede.
3.3.2 Evaluations of Specific Facilities Made During the Reconsideration Period
The previous evaluation of Amersham was judged to be adequate; therefore, it was
not re-evaluated as part of this study. Because Cintechem has shut down and is
decommissioning its production reactor, it was not included in the current evaluation. In
March 1991, DuPont Boston and DuPont Billerica were re-evaluated using updated
information obtained from the licensees (SCA91). MaUincrodt's Maryland Heights,
3-16
-------
Missouri, facility, a large facility not analyzed in EPA89, was also included in the March
1991 study. All data and results presented here for these facilities were taken from SCA91.
Source Term Determination
Operations at the DuPont Boston facility are housed in several multi-story buildings
on two city blocks, across the street from each other. Each block contains several buildings
and a large parking lot. The first group of buildings handles virtually all the radioactivity
and has five roof-top stacks, serving 140 hoods and hot cells. Three stacks are on one
building; two stacks are on a second building. For dose calculations, these were modeled as
two stacks (18 m and 24 m high), one for each building. This study used the emission data
obtained from Dupont Boston for 1989. The 1987 release data are shown for comparison in
Table 3-8.
Table 3-8. Dupont Boston emission data.1
[
NttcIMe- '
H-3
C-142
C-143
S-35
Release JRata
3-17
-------
Table 3-9. Dupont Billerica emission data.
NucIidV
Xe-133
P-32
S-35
1-125
1-131
Kr-85
:* **dOT fete «OHt'
198?' '
2.84
1.6E-02
1.6E-02
2.0E-02
2.5E-02
9.5E-01
1,989
n/a
n/a
n/a
1.9E-02
n/a
n/a
At DuPont Billerica, four radiological stacks serve many hoods, glove boxes, hot
cells, and reaction vessels. For dose calculations, they were modeled conservatively as a
single 15 m stack.
For dose calculations, the Mallincrodt facility was modeled with two roof-top stacks.
Stack #1 (19 m high) models all stacks at the northwest end of the site; stack #2 (13 m high)
models those at the southeast end. ,
Effluents for the Mallincrodt facility were provided for the 12 months ending August
31, 1989, based on measured data. Effluent values based on a calendar year were not
available; however, the radiation safety officer (RSO) indicated that the values provided were
representative of a typical year. Release data are provided in Table 3-10.
Table 3-10. Mallincrodt emission data.
Niidida \.,
1-131
1-125
1-123
Tc-99m
Mo-99
La-111
Ga-67
; Keleasa Sate (Ci/yt1)
13-ia Stadc :
1.5E-02
-
1.6E-03
-
'
-
-
19-ffl Slack
2.2E-01
7.0E-04
1.5E-03
7.7E-02
6.3E-03
l.OE-03
6.0E-04
3-18
-------
Meteorologic. Demographic, and Agricultural Data
Dose calculations for DuPont Boston were performed using wind rose data for Logan
Airport which were obtained from DuPont. Doses for DuPont Billerica were calculated
using COMPLY's default mean wind speed of 2 m/sec. Dose calculations for Mallincrodt
were performed using wind rose data for the St. Louis, Missouri, Airport. Mallincrodt
supplied the meteorological data.
Several residences are located across the street from the Dupont Boston facility. The
distance between stack #1 and one of these residences is 60 m. The distance between stack
#2 and another residence is 50 m. Although not the same residence, COMPLY treats them
as such. The nearest farm is assumed to be 1,000 m away. Meat, milk, and vegetable
production was assumed to take place at this distance. The closest receptor to the Dupont
Billerica facility is a residence located 165 m from the stack. A vegetable farm is located
about 500 m from this stack. A milk and meat farm is located about 1,400 m away.
An office was identified as the closest receptor during a site visit to Mallincrodt.
This office, located within the same industrial park as the licensee, is 215 m from stack #1
and 130 m from stack #2. Aerial photographs made available for inspection by Mallincrodt
and onsite inspections were used to locate a vegetable garden 261 m from stack #1 (410 m
from stack #2). No.milk or meat farms were found within 800 m. Thus, a default distance
of 800 ni was used for these receptors.
3.3.3 Results of the Designated Survey for Radiopharmaceutical and Radiolabeled
Compound Manufacturers
The calculations resulted in a receptor ede of 5 mrem/yr for the DuPont Boston
facility. The total ede for the DuPont Billerica and Mallincrodt facilities, respectively, were
0.2 and 0.09 mrem/yr. For Mallincrodt, the dose is to a nonresident in an office; therefore,
an occupancy factor of 0,3 was applied. The value of 0.3 is based upon 10 hours per day,
5 days per week, 52 weeks per year (10x5x52/8760=0.3). Refer to Section 3.9 for a
summary of dose estimates.
3-19
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'3.4 HOSPITALS AND MEDICAL RESEARCH FACnJTIES
Licensees engaged in medical diagnosis, treatment, and biomedical research constitute
the largest subgroup of NRC-licensed facilities using radioactive materials in unsealed forms.
The facilities within this subgroup range from individual medical practices to large medical
centers. An individual physician may perform an occasional diagnostic procedure using
radiopharmaceuticals, while the large medical centers may engage in extensive biomedical
research using radioactive materials as well as perform diagnostic and therapeutic procedures
involving radiopharmaceuticals on a daily basis.
3.4.1 Previous Evaluations
In its previous assessments of NRC-licensed facilities using radionuclides for medical
purposes (EPA89), EPA focused on large hospitals and medical research facilities.. Due to
the quantity of radioactive materials used and the proximity of potential receptors, such
facilities provide an upper-bound of the dose for this large segment of the NRC-licensed
source category. •
In EPA's previous assessments, data on airborne emissions from such facilities were
limited. Limitations were also inherent in the near-field estimates of air concentrations
provided by the Gaussian plume dispersion model incorporated in the assessment code
AIRDOS-EPA. When EPA first proposed a NESHAP for NRC-licensed facilities in 1983, it
attempted to identify whether the proposed standard would have an impact on medical
facilities (SCA84). Based on discussions with personnel involved in nuclear medicine, EPA
identified approximately 15 facilities with extensive programs. Information on these facilities
was gathered to determine the concentration of radioiodines in their effluent and the location
of the nearest receptors. Based on assessments of the dispersion factors needed to reduce the
effluent concentrations to a level consistent with the proposed standard, it was concluded that
the facilities could comply with the NESHAP without having to install additional effluent
controls.
During the 1988-1989 radionuclide NESHAPs rulemaMng, EPA sought to overcome
the limitations in the emissions data by evaluating the doses that could result from the largest
releases from medical licensees, as reported in the database maintained by the Conference of
Radiation Control Program Directors (CRCPD). Calculations performed to evaluate the
3-20
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"large hospital" category in EPA89, indicated that the maximum estimated dose to nearby
individuals would be approximately 0.2 mrem/yr ede. However, the evaluation cautioned
that "the absence of reported radioiodine releases is common, due to the lack of effluent
monitoring at hospitals." When coupled with the limitations of the assessment code in
evaluating near-field concentrations, considerable uncertainty remained as to whether the
releases evaluated for the "large hospital" actually bound the doses and risks caused by this
class of licensees. -
3.4.2 Evaluations of Specific Facilities Made During the Reconsideration Period
When the NESHAP for NRC-licensed facilities was promulgated on December 15,
1989, the Administrator announced that he was treating the concerns relating to duplicative
regulation and possible adverse impacts on the availability of medical treatment raised by
NRC and the National Institutes of Health (NIH) during the public comment period as a
petition to reconsider the NESHAP. The Administrator granted this reconsideration, and the
effective date of the NESHAP was stayed during the reconsideration.
Inasmuch as the concerns raised by the NEH and other commentators on the NESHAP
focused on the stringency of the 3 mrem/yr ede limit for doses from radioiodines, EPA again
attempted to identify medical facilities using large quantities of radioiodines. Beginning with
information supplied by the medical facilities, EPA determined that the following medical
centers have therapeutic and biomedical research programs that are among the largest in the
country: the National Institutes of Health, Johns Hopkins Medical Center, the University of
California at Los Angeles (UCLA), Washington University Medical Center, M.D. Anderson
Medical Center, the University of Wisconsin, the University of California at San Francisco
(UC San Francisco), and the University of California at Irvine (UC Irvine).
Cognizant personnel at each facility, usually the Radiation Safety Officer, were
contacted, and voluntary cooperation in assisting EPA was requested. Information on
quantities of radioactive materials used, effluent concentrations, effluent controls employed,
and locations of nearby individuals was obtained for each facility (SCA91). In several
instances, site visits were arranged.
It was determined that the doses caused by releases from the M.D. Anderson Medical
Center and the Washington University Medical Center, both of which employ multi-curie
3-21
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quantities of radioiodines but with double or single charcoal filtration, would be bounded by
the estimates for Johns Hopkins and the NIH which handle large quantities of radioiodines
and do not have filtration systems. Therefore, formal COMPLY evaluations of M.D.
Anderson and Washington State University Medical Centers were not performed.
Source Term Determination
Data obtained from NIH, Johns Hopkins, the University of Wisconsin, UCLA, UC
San Francisco, and UC Irvine (SCA91) were evaluated using the EPA computer code
COMPLY. Where measured effluent data were unavailable, source terms were estimated by
multiplying the amount of each radionuclide used during a one-year period by the appropriate
release fraction, as established in EPA89a. However, two facilities, UCLA and Johns
Hopkins, the EPA-approved release fraction of 1 was not used for materials heated to above
100° C. Instead, the evaluation relied on release fractions determined from measurements of
actual releases of the radionuclides of interest. The source terms used in the COMPLY runs
are given in Table 3-11.
Meteorological. Demographic, and Agricultural Data
Johns Hopkins CSCA91^: Because this facility is in an urban setting, the receptors are close
to the release points. Multi-story buildings, containing both commercial stores and
residences, are directly across the street from the licensee. One such building is located
approximately 30 m north of the Biophysics (P-B) and Wood Basic Sciences (WBS)
buildings. Another is located 30 m north of the Traylor (T) building. Analysis showed the
maximum receptor to be located 30 m north of the P-B and WBS buildings, and 153 m from
the T building. Given the urban siting of the facility, it was assumed that no food production
occurs within 4,500 m. Dose calculations were performed using an average wind speed of
3.17 m/s. This wind speed was based on 5-year meteorological information collected from
the Baltimore-Washington International Airport, approximately 10 km from the site.
University of Wisconsin (SCAgi"): Demographic data obtained for this study show that the
nearest receptor is a campus heat plant located 105 m to the west of the incinerator stack.
Although there is an agricultural program on campus, no commercial farming is done. The
nearest farms are estimated to be 1,500 m from the incinerator stack. Doses were calculated
using the COMPLY default wind speed, of 2 m/s.
3-22
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Table 3-11. Hospital and medical research facilities effluent release rates.
Facility Hams/Location
Johns Hopkins
Baltimore, MD
University of Wisconsin
Madison, WI
r
Release Itoint/Staok
Height ' '
WBS Building
51m
P-B Building
13 m
T Building (incinerator)
51 m
'
Incinerator
10m
JSfscJide
••
H-3
C-14
Mo-99
Tc-99m
P-32
S-35
Xe-133
1-125
1-131
Cr-51
Ce-141
Gd-153
1-125
Ih-114
Nb-95
Ru-103
Sc-46
Sn-113
H-3
C-14
P-32
S-35
Ca-45
1-125
1-131
Sr-85
Na-22
Sc-46
Cl-36
Cr-51
Co-57
In-Ill
Sn-113
Ce-141
Se-75
Cl/yr
5.0E+00
5.0E-01
2.1E-10
1.7E-03
2.1E-05
2.5E-05
15.6
1.4E-02
9.5E-04
4.0E-03
7.3E-03
8.5E-02
2.0E-03
1.3E-01
7.2E-02
6.3E-O2
3.1E-02
7.7E-02
3.5E-02
7.9E-02
4.7E-02
3.9E-01
2.0E-02
3.3E-02
8.5E-04
2.0E-03
1.7E-03
1.4E-03
l.OE-03
1.7E-03
1.6E-03
2.4E-03
1.2E-03 .
2.0E-03
7.0E-05
3-23
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Table 3-11 (Continued)
Facility Name/Lofcatiofl,
UCLA Hospital
Los Angeles, CA
UC San Francisco
San Francisco, CA
UC Irvine
Irvine, CA
NIH
Bethesda, MD
fe*^3**tfSta*
Height:
Hospital
5 m
MS Building
56 m
Nuclear Med. Building
5 m
.
NIH Complex
42m
Nuclids
H-3
C-ll
C-14
F-18
P-32
S-35
Ca-45
Cr-51
1-125
1-131
Mo-99
Tc-99m
Xe-133
Tl-201
1-125
1-131
P-32
Cr-51
Mo-99
Tc-99m
1-125
1-131
Xe-133
Co-57
C-14
Cr-51
Ga-67
H-3
1-123
1-125
1-131
Mo-99
P-32
S-35
Te-99m
Ci/yr
v
2.2E-03
7.0E-03
1,OE-04
32.8
6.7E-03
3.2E-03
5.0E-04
1.3E-03
1.5E-O3
3.0E-03
1.6E-04
7.0E-01
6.2E+00
6.5E-03
2.5E-03
2.0E-03
LOE-04
l.OE-05
1.5E-07
6.8E-02
l.OE-04
2.0E-03
10.4
5.0E-05
3.8E-04
6.5E-03
2.6E-03
2.2E-02
3.2E-05
6.7E-03
1.3E-02
3.4E-04
2.2E-02
1.9E-02
2.1E-02
3-24
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UCLA (SCA91): Due to the lack of specific demographic data, the distance to the closest
receptor was estimated to be 100 m. It was assumed that this receptor grows vegetables.
Meat and milk farms were estimated to be at a distance of 1,000 m. Doses were calculated
using the COMPLY default wind speed of 2 m/s.
! - " '
UC San Francisco (SCA91^: The nearest receptor is a commercial office across the street
from the top of the MS building. The height of this building is 56 m. The nearest receptor
is a commercial office approximately 30 m from the MS building. The location of the
nearest farms was not known. It was estimated that a vegetable garden could be found
500 m away and that the distance to the nearest farms is 1,600 m. Doses were calculated
using the COMPLY default wind speed of 2 m/s.
UC Irvine (SCA91): The nearest receptor was determined to be a commercial building
across the street from the hospital at an estimated distance of 50 m. Estimated distances to
the nearest vegetable garden and farm are 800 m and 16,000 m, respectively. The receptor
is a nonresident in an office; therefore, an occupancy factor of 0.3 was applied. The value
of 6.3 is based upon 10 hours per day, 5 days per week, 52 weeks per year
(10x5x52/8760=0.3). Doses were calculated using the COMPLY default wind speed of
2 m/s.
MHS: The nearest receptor was determined to be a resident located at a distance of 200 m.
The resident is assumed to grow vegetables at home. Meat and milk farms are placed at
2,000 m. Doses were calculated using the COMPLY default wind speed of 2 m/s.
*. -. ~
3.4.3 Results of the Designated Survey for Hospitals and Medical Research Facilities
The highest estimated dose from any of these facilities is 8 mrem/yr ede to a receptor
located directly across the street from the incinerator at Johns Hopkins. Radioiodines
contributed 0.4 mrem/yr ede to this total. The highest estimated ede from iodines is
1 mrem/yr. This dose was calculated for Nffi. The total ede calculated for N3H was
2.0 mrem/yr; therefore, the dose from iodines constitutes 50 percent of the total. The
remainder of the ede from the hospitals and research facilities included in the Designated
Survey ranges from 0.03 mrem/yr to 3 mrem/yr. Refer to Section 3.9 for a summary of
dose estimates.
5 Personal correspondence between R. Zoon (Nffl) and A. Colli (EPA), November 1989.
3-25
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3.5 MANUFACTURERS OF SEALED SOURCES
/
Sealed source manufacturers take radionuclides in an unsealed form and put them into
a permanently sealed container. Two categories of sealed source manufacturers contribute to
airborne emissions. The first category consists of manufacturers (eight are known) that
produce sealed radiation sources other than tritium. An additional six manufacturers of this
type (e.g., The Nucleus, Oak Ridge, Tennessee) use only exempt quantities of radionuclides
and produce negligible emissions.
The other category of sealed source manufacturer seals tritium gas into self-luminous
lights. Currently, two firms are known to perform this type of work. They are Safety Light
Corporation, in Bloomsburg, Pennsylvania, and NRD, Incorporated, in Grand Island, New
York. Both facilities are located in industrial areas. They rely heavily on engineered
safeguards to prevent releases of radionuclides.
3.5.1 Previous Evaluations
Three tritium light sealed source manufacturers, Safety Light Corporation in
Bloomsburg, Pennsylvania, NRD in Grand Island, New York, and GE Lighting Group in
Cleveland, Ohio, were originally evaluated in EPA89. One manufacturer, GE Lighting
Group, has since gone out of production. The evaluations reported in EPA89 estimated the
highest dose to nearby individuals from non-tritium sealed source manufacturers to be
l.OE-04 mrem/yr ede, and from tritium sealed source manufacturers to be 6.0 mrem/yr ede.
3.5.2 Evaluations of Specific Facilities Made During the Reconsideration Period
In EPA89, a model facility was used to represent manufacturers that produce non-
tritium sealed radiation sources. Since EPA89 was prepared, an actual facility, Neutron
Products (Dickerson, Maryland) which is a major producer of cobalt-60 sealed sources, has
been identified as a large manufacturer of non-tritium sealed sources. This facility was
evaluated based on emissions data and demography information supplied by the licensee.
The findings are incorporated in this study.
3-26
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Source Term ,
Sealed Sources/Non-Tritium: Neutron Products is a major producer of cobalt-60 sealed
sources. All operations with possible airborne emissions are conducted in the hot cell. All
site releases are exhausted from a single vent, which is located approximately 7 m above the
ground. The exhaust rate is 23 m3 per minute (800 cfm).
The effluent exhaust from the hot cell passes through a roughing filter and two HEPA
filters mounted in series. The exhaust vent was recently equipped with a continuous
monitoring system. The sampling is isoMnetic, drawing 0.03 m3 per minute (1 cfm) through
a fiber filter. The filter is changed at least weekly and counted using single-channel gamma
spectrometry.
-«* - '
The sampling system described above is reported to have a minimum detection limit
(MDL) of 1E-12 jtiCi/ml, approximately 0.3 percent of the MFC for insoluble forms of
cobalt-60. All measurements with this new system show activity below the MDL, The 1989
source term for the facility is estimated to be 1.2E-05 Ci/yr (see Table 3-12), assuming the
MDL for the concentration in the effluent and a continuous flow rate of 23 m3 per minute
(800 cfm). , : ' .
Sealed Sources/Tritium: Because effluent data for 1984 were available for each tritium
lighting producer when the EPA89 analysis was being done, no model facility was needed.
The emissions data used in, the analysis are also shown in Table 3-12.
Table 3-12. Effluent release rates (Ci/yr) for sealed source manufacturers.
B&fipaaeSjfe
H-3
Co^-60
Ni-63
Po-210
Am-241
Nea^tjal'radactg
,
1.2E-05
-
-
'
MRD, lac,
3.4E+02
-
8.0E-06
1.4E-04
6.1E-05
Safety Light Corp.
2.2E+03
_
_
'
" " . -
3-27
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Meteorological. Demographic, and Agricultural Data
The meteorological data used in this study for NRD and Safety Light were originally
collected at Buffalo, New York, and Harrisburg, Pennsylvania, respectively. The stability
array data for these locations, that had been used in EPA89, were converted to wind roses
for use with the COMPLY code. Neutron Products was evaluated using the default values of
25 percent frequency of wind towards the receptor and a wind speed of 2 m/s.
Demographic data obtained for Neutron Products show several farms in the area.
The nearest residence is a farm approximately 120 m from the vent. All meat, milk, and
vegetable production was assumed to occur at that location.
Detailed demographic data were obtained for NRD and for Safety Light Corporation.
Based on the COMPLY runs, the receptor near NRD who is exposed to the highest
concentration is a resident located approximately 170 m from the stack. At Safety Light, this
individual is a resident located 190 m from the release point.
Agricultural information supplied by the NRD and Safety Light facilities indicates that
there are no farms located within 2,000 m of either site. Doses for both facilities were
calculated assuming that the residents produce all their own vegetables and that meat and
milk production occurs on farms located at 2,000 m. These assumptions were made to
maintain consistency with the Random Survey portion of this study.
3.5.3 Results of the Designated Survey for Manufacturers of Sealed Sources
The results from the COMPLY model for the non-tritium sealed source manufacturer
(Neutron Products), using the source term of 1.2E-05 Ci/yr, actual vent and building
f dimensions, and a default wind speed of 2 m/s, indicate that the receptor ede would be
0.007 mrem/yr. The dominant pathway is exposure to contaminated ground.
The maximum ede fqr a tritium light sealed source manufacturer is calculated for
Safety Light Corporation. Using the source term described above, a release height of 10 m,
and meteorological data from Harrisburg, Pennsylvania, the ede calculated by COMPLY for
the maximum individual is 3.5 mrem/yr. Most of the dose results from the inhalation and
ingestion pathways.
3-28
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At NRD, the receptor for whom the highest dose is calculated is a resident.
Assuming the source term described above, a release height of 10 m, and meteorological data
from Buffalo, New York, the ede calculated by COMPLY for the maximum individual is
0.05 inrem/yr. Inhalation is the primary pathway of exposure.
3.6 TESTING OF DEPLETED URANIUM MUNITIONS
The processing of natural uranium to obtain uranium enriched in the uranium-235
isotope results in abundant tails referred to as depleted uranium. The density and low
specific activity of depleted uranium make it useful for several applications, including
radiological shielding, counterweights in aircraft, and in military munitions. This latter
activity has the greatest potential to release radioactive material to the air.
The military uses depleted uranium in munitions designed to pierce armor plating.
The design of these munitions is developed and refined by the Army based on "soft" and
"hard" testing. Soft testing is conducted to define and refine the accuracy of the munitions.
The tests are done on outdoor firing ranges where the depleted uranium round is fired at the
"target" located in a sand-filled testing enclosure several kilometers from the gun. After
impact, the depleted uranium "rod," which is generally intact, is simply left in the ground as
the risk from unexploded munitions makes retrieval too dangerous. Hard testing is
conducted to evaluate and refine the destructive capability of the munitions. In hard testing,
either actual munitions or scale mockups are fired at an armor-plated target. By license
conditions, all hard testing of depleted uranium munitions is conducted in indoor test
enclosures; the ventilation stacks of which are equipped with roughing and HEPA filters; the
exhaust is monitored during testing.
The Department of Defense tests depleted uranium munitions at a number of proving
grounds around the country. The Army's Ballistic Research Laboratory and Combat Systems
Test Activity facilities at the Aberdeen Proving Ground in Aberdeen, Maryland, conduct both
hard and soft testing. The Army also conducts soft testing at the Yuma Proving Ground near
Yuma, Arizona, and at the Jefferson Proving Ground near Madison, Indiana; the Navy
conducts soft test firings at the Naval Weapons Center at China Lake, California. Once
every several years, the Army conducts an open-air hard test firing at the Nevada Test Site.
3-29
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The Aberdeen Proving Ground conducts the greatest number of test firings. Because
it is also very close to many residences, EPA considers Aberdeen to be the bounding case for
this category.
3.6.1 Previous Evaluations
This source category of airborne radionuclide emissions was not previously evaluated
because it was believed unlikely that munitions testing could create emissions in the
respkable range. However, to remove any uncertainty, this source category was evaluated in
this study.
» . • \
3.6.2 Evaluations of Specific Facilities Made During the Reconsideration Period
A site visit to the Aberdeen facility was conducted during the course of this
reconsideration. The releases from the test firing of depleted uranium munitions include
stack releases from the indoor test enclosures used for hard firings and releases to the .
ambient air from the soft testing target enclosures, which may occur when the depleted
uranium rods land. Given the size of the rods left in the enclosures (on the order of 1 to
8 kilograms), releases due to resuspension are not a problem, as confirmed by ambient air
monitoring conducted by the Army. Monitoring data on the stack releases from the indoor
testing enclosure, along with stack parameters and distances to the nearest receptors, were
obtained directly from the Army (DA92).
Source Term
. t •
The emissions used in the analysis are shown in Table 3-13. These emissions
represent monitored stack release data from indoor testing enclosures as provided by the
Army.
Meteorological. Demographic, and Agricultural Data
The meteorological information is stability array data from Aberdeen, Maryland. The
distances to the nearest residences or office, school, or business for each of the operations
listed in Table 3-13 are provided in Table 3-14.
3-30
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Table 3-13. Source term used for Aberdeen Proving Ground.
- Qpeiatiott
Range 9
Range 14
Range 14A
Range HOE
Abrasive Blaster
BTD Enclosure
Superbox
Cut Box
t!»238 Release Hate, Ci/yjr1 -
6.6E-07
1.2E-07
1.2E-07
4.5E-08
8.9E-07
1.8E-06
5.7E-05
1.6E-05
1. It was assumed that Th-234 and Pa-234m were also released at the
same rate as the U-238 as they are in secular equilibrium.
Table 3-14. Distances to receptors at Aberdeen Proving Ground.
Operation
Range 9 .
Range 14
Range 14A
Range HOE
Abrasive Blaster
BTD Enclosure
Superbox
Cut Box
Instance, jn
5000 (R)1
7000
-------
3.7 RARE EARTH AND THORIUM PROCESSORS (SOURCE MATERIAL)
Approximately 10 licensed rare earth processors are engaged in the recovery of
metals from source materials. Of the 10 facilities licensed to process rare earths, only three
are operating. These three form the basis for this study: Cabot-Boyerton, Molycorp-York,
and Shieldalloy-Newfield. The doses resulting from the operations of rare earth processors
were assessed using the actual emissions and site characteristics for the three facilities.
Rare-earth elements are metals possessing distinct individual properties which make
them potentially valuable as alloying agents. The name rare earths is deceiving, however,
because they are neither rare nor earths. Rare earth minerals exist in many parts of the
world, and the overall potential supply is essentially unlimited.
Rare earth facilities possessing NRC Source Material Licenses process natural and
synthetic ores which contain at least 0.05 percent, by weight, of. naturally occurring uranium
and thorium. The principal environmental impacts of rare earth facility operations include
the potential release of radioactive particles and radon from the storage, handling, and
processing of the ores. The operation of a rare earth facility involves grinding, dissolving,
and processing the natural and synthetic ores. These are relatively closed processes, and it is
generally believed that very limited amounts of radioactivity escape. These facilities utilize
various methods to store the radioactive wastes. The wastes are often stored on-site in
barrels or slag piles.
3.7.1 Previous Evaluations
EPA conducted a screening assessment in 1983 and concluded that rare earth and
thorium processors did not pose a public health risk (EPA83). However, EPA decided to
reduce the uncertainty associated with the 1983 evaluation.
NRC conducted an evaluation of Cabot-Boyerton (NRC88). The rate of release of
the materials had not been previously determined, so conservative assumptions were made.
Doses were estimated using AIRDOS. For the nearest individual (350 in), the total-body
dose of 0.046 mrem resulted primarily from the inhalation (54 percent) and ingestion
(30 percent) pathways. The highest dose was to the lungs (0.48 mrem).
3-32
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In 1985, the Oak Ridge Associated Universities conducted a radiological study of
Molycorp-York (ORAU85). The summary noted that air monitoring at two process stacks
indicated that radioactive emissions from plant operations were within licensed limits. The
ORAU study also noted that residues from plant processes are stored in onsite low-level
waste drums and a residue pile located in the southeast corner of the site.
3.7.2 Evaluations of Specific Facilities Made During the Reconsideration Period
Cabot-Boyerton: This facility is located in a rural setting in southeastern
Pennsylvania, 2.4 km northeast of Boyertown. Ores are processed in order to extract
tantalum and niobium. Typical concentrations of uranium and thorium range from
0.04 percent to 0.5 percent by weight. Surface radiation dose rates typically range from
0.1 mrem/hr to 2 mrem/hr.
Raw ores are ground into a flour-like consistency and then transferred into digester
tanks which selectively dissolve the tantalum and niobium. The unwanted uranium and
thorium react with the acid to form insoluble uranium and thorium fluorides. Particles less
than 10 pm in .diameter are exhausted through the 90 percent efficient dust-collection system.
Up to 100 g/d of respirable particles might enter the atmosphere. After dissolution, the
mixture passes through filters where the insoluble material (containing the uranium and
thorium) is removed from the solution and collected for disposal.
The sludge is temporarily stored in open portable carts until a truckload of filled
containers is collected and transported to above-ground concrete storage buildings. Each
building is open-air vented where the roof meets the side walls to prevent radon gas from
accumulating inside the building.
Cabot does not have a formal environmental monitoring program, and routine outside
air monitoring has not been conducted. It is thought that the operating procedures and
emission controls combine to limit radiological airborne releases to low levels. However, no
monitoring data are available to confirm this. NRC does not require any offsite
environmental monitoring program due to the limited effects expected.
Molycorp-York: This facility, active since the mid-1960s, is located in an urban
area. .The Molycorp plant carries on three basic processes, all of which involve low
3-33
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concentrations of source material. All three processes operate under the same basic theory,
although only one is now operating. The main working process at Molycorp converts code
5300 cerium mineral concentrate into a line of 95 percent pure cerium products. The cerium
concentrate process feed material is a dry powder. Thorium and uranium are present at
about 0.225 percent and 20 ppm, respectively. A typical cerium reaction charge is 1,800 kg
per digest tank, containing about 0.4 kg each of thorium and uranium. All chemical
processing after the initial feed dissolution is wet processing, thereby reducing airborne
particulates.
After the dissolution process, thorium and uranium remain as insoluble byproducts.
These byproduct materials, containing about 50 percent moisture, are shoveled into 208-liter
(55-gallon) plastic drums for storage. Approximately 145 barrels (52,200 kg) are produced
per month.
In order to reduce airborne particles, a 0.8 m diameter, 4.3 m high, wet scrubber is
used at the cerium feed point to capture any dust and recycle it back into the system. The
scrubber is equipped with an 85 m3 per minute (3,000 cfm) blower and circulates 170 1pm
(45 gpm) of scrubbing solution over the packed bed. Employees are periodically monitored
at the points of greatest exposure to radioactive dust. Results show that the radiation dose to
plant personnel is low; therefore, Molycorp expects that the dose to the surrounding
population is minimal. There is no routine monitoring program for effluents into the
atmosphere. The dust collectors and scrubbers are inspected periodically, but the inspections
are usually only visual, without monitoring of the effluents.
SbJeldaJlby-Newfleld: This active facility is located in a rural area. Shieldalloy
manufactures a variety of specialty ferro-alloys, using the raw material ferro-niobium
(Fe-Nb). Waste slag is separated from the nonradioactive slag and stored in two separate
piles. A large quantity of material has accumulated since operations began in 1955.
Processing activities generate airborne dusts, containing low concentrations of
radionuclides from the thorium and uranium decay series. Exhaust air from the processing
area passes through 10,000 mVmin baghouse dust collectors before its release to-the
environment. The maximum amount processed per day would be about 400 /tCi of thorium
and about 3.6E+08 g of natural uranium. The bags are 98 percent efficient. Shieldalloy
uses an air sampler to monitor releases.
3-34
-------
There is no indication that the waste slag piles are stabilized or have any sort of cover
on them. The most likely pathway and source of contamination appears to be overland
runoff from the pile. Sample analysis was underway as of August 14, 1991. Shieldalloy
will also perform a risk assessment of offsite contamination, and remediation of both the
radiological and chemical contamination will be evaluated. Following cleanup, the source
material will be stabilized. No measures have been taken so far to keep additional low levels
of radiological contamination from being transported off site. NRC has also requested
Shieldalloy to provide a plan that would demonstrate compliance with the stricter limits
proposed in 10 CFR Part 20 (effective January 1994), and also with NESHAPs. Shieldalloy
considers perimeter air sampling sufficient to demonstrate compliance.
Source Term Determination
The three operating rare earth processors were surveyed by EPA., Molycorp and
Shieldalloy supplied process source term data in response to the survey. Comparable
information was not available for Cabot Corporation. Instead, site meteorology was used in
conjunction with the methods in Regulatory Guide 3.59 to derive the airborne source term
for sludge that is stored in open-air vented mausoleums. Table 3-i5 presents the source
terms used in this study.
Table 3-15. Rare earth processors' annual release rates.
Facility
Cabot Corp.
Molycorp, Inc.
Shieldalloy
Bslease PoM
Mausoleums
Tank Room
Waste Treatment
Moly Building
Department 111
Stack Height
(»)
1
10
5
2
12.2
Natural
Thorium*
-------
Meteorologic. Demographic, and Agricultural Data
The meteorological data used in this study for Cabot Corporation, Molycorp, and
Shieldalloy were originally collected at Reading, Pennsylvania; Harrisburg, Pennsylvania;
and Millvalle, New Jersey; respectively. These data, which were in the form of stability
arrays (EPA89), were converted to wind roses for use with the COMPLY code.
Demographic data obtained for Cabot Corporation indicated that the nearest resident
is approximately 270 m from the mausoleums used to stpre sludge. Vegetables were
assumed to be grown at this location. There are no milk- or meat-producing farms within
2,000 m. Therefore, to maintain consistency with the assumptions used for the Random
Survey, milk- and meat-producing farms were placed at 2,000 m.
Demographic data obtained for Molycorp showed that the individual closest to the
tank room is a resident located at a distance of 100 m. The individual closest to the waste
treatment building is a non-resident located at a distance of 200 m, and the individual closest
to the Moly building is a resident located at a distance of 100 m. Residents were assumed to
produce their vegetables at home. There are no milk- or meat-producing farms within
2,000 m of the facility. Therefore, to be consistent with the assumptions used for the
Random Survey, milk- and meat-producing farms were placed at 2,000 m.
Demographic data for Shieldalloy indicated that the closest individual is a resident
located 225 m from the facility. Residents were assumed to produce their vegetables at
home. There are no milk- or meat-producing farms within 2,000 m of the facility.
Therefore, to maintain consistency with the assumptions used for the Random Survey, milk-
and meat-producing farms were placed at 2,000 m.
ted Survey for Rare Earth and Thorium Processors
The receptor exposed to the highest offsite concentration for Shieldalloy and for
Cabot Corporation is a resident. At Molycorp, this individual is a non-resident; therefore, a
0.3 factor was applied to the dose calculated by COMPLY.6 The maximum ede is
calculated for Shieldalloy Metallurgical Corporation; the dose received by this individual is
* Tie value of 0.3 is based upon 10 hours per day, 5 days per week, 52 weeks per year
(10x5x52/8760=0.3).
3-36
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1.6 mrem/yr. The edes calculated for Molycorp and Cabot are 0.56 and 0.01 mrem/yr,
respectively. Inhalation is the dominant exposure pathway for all three facilities. Refer to
Section 3.9 for a summary of dose estimates.
3.8 COMMERCIAL LOW-LEVEL RADIOACTIVE WASTE DISPOSAL AND
INCINERATION ^
Many users of unsealed radioactive materials will generate solid, low-level radioactive
wastes (LLW) that require disposal. Such wastes may be incinerated on site or packaged and
shipped off site to a licensed low-level waste disposal facility.
LLW is generated from a variety of commercial sources: research, power plants,
diagnostic and therapeutic medicine, manufacturing, and others. When contaminated through
contact with radioactive material, items such as paper, clothing, plastics, power reactor
liquids, and medical fluids are classified as LLW.
Waste Brokers
Waste receivers and shippers (sometimes called "waste brokers") are primarily
collection and shipping agents for facilities generating LLW. Most such receiving-shipping
facilities simply collect the wastes from a number of waste-generating facilities in shipping
containers approved by the Department of Transportation, monitor the packages for
contamination, and hold the wastes at a warehouse until they arrange a shipment to a licensed
disposal site. The licenses of most such receiving and shipping facilities do not allow the
facility to repack or even open the waste packages. However, several such facilities are
licensed to open, compact, and repackage Waste materials before shipment.
Incinerators
Most airborne effluents from handling LLW come from incinerators; The practice of
evaporating disposal site liquids has ceased, so this is no longer a source of releases to air.
Although incineration is done primarily by hospitals and large research laboratories (about
100 such medical incinerators are operating - EPA89), this section deals exclusively with
incinerators licensed specifically for commercial use.
3-37
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Disposal Facilities
Some radionuclides may also be emitted from LLW disposal sites. Currently, only
three sites (Bamwell, South Carolina; Beatty, Nevada; and Richland, Washington) are
operating. Disposal of LLW is controlled by the Low-Level Radioactive Waste Policy Act
of 1980 and its 1985 amendment. Any state that wishes to dispose of its LLW may join an
interstate compact or design its own facility in accordance with 10 CFR 61, among other
options.
LLW disposal sites do not accept/spent reactor fuels, but may accept special nuclear
materials and transuranics meeting the classification requirements in 10 CFR Part 61. The
majority of LLW .comes from power reactor operations, laboratory research, and medical
facilities. "
Currently operating disposal sites typically consist of a large fenced burial area with
buildings for decontamination, maintenance, and waste preparation in one location. Wastes
are usually buried in the transport containers in which they arrive, which minimizes releases
to the atmosphere. The buried wastes are covered by overburden. New facility designs
being proposed typically use a liner and a clay and/or concrete cap in addition to engineered
barriers.
3.8.1 Previous Evaluations
Both incinerators (EPA89) and disposal facilities (EPA79, EPA84) have been
previously investigated. Airborne emissions from waste brokers are judged to be bounded by
the operation of burial and incineration facilities.
Previously, EPA's evaluation of incinerators was limited to those which were part of
hospital and medical research facilities because no commercial LLW incinerators existed.
Since EPA89, a commercial LLW waste incinerator has been licensed and is included in this
study. .
The potential public health impacts of the release of radioactive materials into ambient
air from LLW burial sites have been evaluated previously (EPA79, EPA84). The doses
received by the most exposed members of the public were found to be below the limits
established in the NESHAP. ;
3-38,
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3.8.2
ations of Specific Facilities Made During the Reconsideration
EPA investigated the LLW disposal process by evaluating the quarterly emissions
reports of licensees for operating facilities and reviewing the license application of newly
proposed LLW waste compact facilities.
For operating LLW disposal sites (Hanford, Barnwell, and Beatty), EPA confirmed
its prior analyses (EPA79, EPA84) through conversations with state radiation control
officers.7
For compact sites, EPA reviewed the NESHAP applications submitted by U.S.
Ecology for the disposal site proposed for Needles, California, and by Bechtel for the site
proposed for Butte, Nebraska. The applications were prepared following conservative EPA
guidelines (EPA89a). For example, for the Needles application, it was assumed that the
nearest receptor produces his own vegetables, meat, and milk at his home.
For incineration, EPA investigated the SEG incinerator located in Oak Ridge,
Tennessee. Quarterly data include radionuch'de content in waste incinerated, stack effluent,
scrubber effluent, and ash generated. EPA's review was based on data reported during the
12 months of 1990. Independent analyses were not performed.
3.8.3 Results of the Designated Survey for Waste Disposal and Incineration
Environmental monitoring results for the operating LLW disposal sites indicate that
releases above background have not been detected. As a result, no COMPLY calculations
were made for these sites. i
For the proposed compact LLW disposal sites, the dose to the nearest receptor is
estimated to be 7E-01, mrem/yr ede and 6E-01 mrem/yr ede from radioiodines for the Butte,
Nebraska, site (USE91). For the Needles, California, site, the dose is estimated to be
7E-01 mrem/yr ede and 7E-01 mrem/yr ede due to iodine (USE89).
7 Mr. L. T. Skoblar of SC&A held conversations with the following persons during February 1992: Mr.
Virgil Autry, South Carolina; Mr. John Vaden, Nevada; and Mr. Gary Robertson, Washington.
3-39
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For incineration operations, the dose to the maximally exposed individual is
established at less than 7E-03 mrem/yr ede, with 3E:04 mrem/yf ede from radioiodines
(SEG91).
3.9 SUMMARY OF RESULTS
Table 3-16 summarizes all doses estimated as part of the Designated Survey. As in
previous EPA assessments of actual facilities, the NRC licensees studied are found to be
currently meeting the dose limits of Subpart I.
3-40
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Table 3-16. Summary of Designated Survey doses.
Type of Facii%-
Uranium Tailings
Company & Facility Name
American Nuclear Corp
Federal American Mill
Eiverton, WY
Anaconda ;
Anaconda Mill
Bluewater, NM
Atlas
MoabMill
Moab, UT
Exxon
Highland Mill
Douglas, WY -
Homestake
Homestake Mill
Grants, NM
KerrMcGee
Kerr McGee Mill
Ambrosia Lake, NM
Minerals Exploration ^
Sweetwater Mill
Rawlins, WY
Pathfinder
Lucky Me Mill
Riverton, WY
Pathfinder
Shirley Basin Mill
Casper, WY
Petrotomics
Petrotomics Mill -
Medicine Bow, WY
Rio Algom
Rio Algom Mill
La Sal, UT
Umetco Minerals
Gas Hills Mill
Riverton, WY
Umetco Minerals
White Mesa Mill
Blanding, UT
Umetco Minerals
UravanMill . _.• •
Uravan, CO
Total
ede1 {BBi£em/yr>
3E-01
6E-01
1E-01
1E-01
2E-01
4E-02
4E-O2
6E-O2
2E-01
2E+00
4E-02
3E-01
2E-01
8E-03
Iodine
«de {jmrem/yr}
N/A
N/A
N/A
. N/A
N/A
N/A
N/A
N/A
N/A
N/A
N/A
N/A
N/A
N/A
3-41
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Table 3-16. (Continued)
Type of Facility
UFfi Plants
Fuel Fabrication Facility
Test & Research Reactors
Radiopharmaceutical
Manufacturers
Hospital & Medical
Research Facilities
" Company & Facility Name
United Nuclear ,
Church Rock Mill
Church Rock, MM
Western Nuclear
Sherwood Mill
.Wellpinit, WA
Western Nuclear
Split Rock Mill
Jeffrey City, WY
- Wet Process
Allied-Signal Inc.
Metropolis, IL
- Dry Process
Sequoyah Fuels Corp.
Gore, OK
Westinghouse CNFD
Columbia, SC
National Institute of
Standards and Technology
Gaithersburg, MD
University of Missouri
Columbia, MO
MTT
Cambridge, MA
DuPont Boston
Boston, MA
DuPont Bilierica
Billerica, MA
Mallincrodt
Maryland Heights, MO
NIH
Bethesda, MD
UCLA
Los Angeles, CA
UC Irvine
Irvine, CA
Johns Hopkins
Baltimore, MD
University of Wisconsin
Madison, WI
Total
ede1 (mrera/yr)
3E-01
2E-01
4E-01
7E+00
3E+00
6E-02
8E-01
2E+00
4E+002
5E+00
2E-01
9E-022
2E+00
3E+00
3E-022
8E+00
6E-012
JodJttfe
ede {narem/yr)
N/A
N/A
N/A
N/A
N/A
• N/A
N/A
N/A
N/A
N/A
< 2E-01
.9E-02
1E+00
1E-01
< 3E-02
4E-01
6E-02
3-42
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Table 3-16. (Continued)
Tyjpe of Facility
Manufacturers of Sealed
Sources •
Testing of Depleted
Uranium Munitions ,
Rare Earth & Thorium
Processors
Commercial Low-Level
Radioactive
Waste Disposal &
Incineration
Cojtap. any ijjfc Facility JSfeme
UC San Francisco
San Francisco, CA
Safety Light Corp
Bloomsburg, PA
NRD, Inc.
Grand Island, NY
Neutron Products
Dickerson, MD
Aberdeen Proving Grounds
U.S. Army
Aberdeen, MD
Molycorp, Inc.
York, PA
Cabot Corporation
Boyertown, PA
Shieldalloy Metallurgical .Corp
Newfield, NJ
US Ecology (USE89)
Needles Site, CA
US Ecology (USE91)
Butte, NE
SEG (SEG91)
Oak Ridge, TN
Barnwell Site . . -
Aiken, SC
Beatty Site
Beatty, NV
HanfordSite
Richland, WA
Total ,
ede1 {jateaat/yt)
3E-022
4E+00
5E-02
7E-03
6E-04
6E-012
1E-02
2E+00.
7E-01
7E-01
<7E-03
Iodine
ede fatoeai/yar}
< 3E-02
N/A
N/A
N/A '
N/A
N/A
N/A
N/A
7E-01
6E-01
<3E-04
Emissions not measurable above
background
Emissions not measurable above
background
Emissions not measurable above
background
•1. Results are for residents unless otherwise stated. All values are rounded to the nearest whole number.
2. Nonresident: COMPLY result multiplied by a factor of 0.3 for people in businesses or offices.
3-43
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4. Results of Random Survey of Licensees
4.1 PURPOSE OF THE RANDOM SURVEY
In the previous radionuclide NESHAPs rulemaking for the source category of the
NRC-licensed facilities other than nuclear power reactors, the Administrator found that
current levels of emissions were acceptable. The dose and risk assessments the
Administrator used in making his decision were based on evaluations of facilities believed to
have the greatest potential emissions, i.e., they were the worst case facilities (see Chapter 3).
However, limitations in EPA's knowledge about the thousands of facilities included in this
source category led to some uncertainty as to whether the facilities evaluated bounded the
maximum doses and risks.
The purpose of the random survey is to provide additional confidence that the
facilities evaluated previously by EPA, and presumed to represent the "worst cases" in terms
of MIR, actually do represent the upper-bound of the doses caused by the NRC-licensed
facilities. Given the number of facilities in the sample (approximately 350), the probability
statement that the highest estimated dose observed in the sample is greater than or equal to
the 99th percentile dose for the entire population can be made at the 95 percent confidence
level.
This chapter evaluates the radiological impacts of NRC's programs, using actual or
estimated data reported by all sampled operating facilities. It presents a current "snapshot"
in time of the doses caused by the normal operation of the NRC-licensed facilities.
In making its evaluation, EPA chose the computer code COMPLY to estimate doses.
COMPLY was chosen because, for many of the situations being assessed, COMPLY's
dispersion model is more appropriate than available alternatives, including the CAP-88
codes. In making dose evaluations, the procedures set forth in EPA89a were followed with
two adjustments. First, the default release fraction of 1 was not used to estimate Xe-133
emissions from radiopharmaceutical manufacturers and nuclear pharmacies. Second, for sites
where the location of the receptor was a school or office rather than a residence, an
occupancy factor was applied. These adjustments are discussed more fully in Section 4.2.
4-1
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4.2 METHODS FOR SELECTING THE RANDOM SAMPLE AND DATA
REQUIREMENTS
4.2.1 Selection Criteria
Because the facilities of interest number in the .thousands, it was not feasible to
evaluate all emissions and doses. Accordingly, the only way to increase the certainty that the
maximum doses observed in the Designated Survey of facilities actually represent the upper
bound is by using statistically significant data obtained from a sample of all facilities. The
statistical approach is based on a random sample of facilities selected from lists of licensed
facilities provided to EPA by NRC and the Agreement States. Faculties with no potential for
airborne emissions during routine operations, i.e., those using radioactive sources only in a
sealed form (sealed sources), such as well-logging, were excluded from the Random Survey.
The only other facilities excluded from the survey were fuel cycle facilities licensed by NRC.
4.2.2 Data Requirements
In order to make the dose estimates, site-specific data were required from users of
unsealed sources of radioactivity. Questionnaires were sent to a random sample of facilities
using radioactive materials to obtain the release rates and other necessary parameters from
those using unsealed sources. The selected assurance level of 95 percent requires a sample
of dose estimates for approximately 300 facilities to infer the dose below which 99 percent of
all licensed facilities lie.
Table 4-1 summarizes the sample selection process and responses. The database
available for sampling, compiled from NRC- and state-supplied data, included approximately
12,000 facilities. State-supplied data were generated in response to an EPA request for
information, sent to each Agreement State. The input obtained for the database includes
licensed facilities using both sealed and unsealed sources.
The database distinguished between those facilities licensed directly by NRC (strata
one) and those licensed through Agreement States (strata two). The relative frequency of
facilities using only unsealed sources differs in these two strata (i.e.,-population of facilities)
due to the differing sources of information on licensees in these strata. Initial sampling of
these strata permitted estimation of the relative proportion of unsealed source sites in each
4-2
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Table 4-1. Summary of Random Survey responses.
,.- .- -• ..- .
1 . Number of Facilities in Database
2. Number of Facilities Surveyed
(percent of item 1 above)
3. Number of Unsealed Source Sites
(percent of item 2 above)
4. Estimated Number of Unsealed
Source Sites in Population
(percent of total)
5. Number of Sites Submitting
Questionnaire Data for COMPLY
(percent of total) . "
6. Estimated Sampling Frequency
of Unsealed Source Sites8
(item 5 as percent of item 4)
NUC
6,600
360
5.5%
170
47%
2,800
45%
169
46%
6.2%
Agreement -
States
5,700
310
5.4%
200
65%
3,400
55%
198
54%
5.9%
1 ' fbtal •
12,300
670
5.4%
370
55%
6,200
367
6.0%
a. The agreement of the three percentages in item 6 indicates a nearly proportional sample of
NRC and Agreement State unsealed source sites.
strata. Sampling frequencies for selection from .the two strata were adjusted slightly to yield
a targeted number of unsealed source sites in each strata. From the entire sample, it is
estimated that 47 percent of the NRC-licensed facilities and 65 percent of the facilities
licensed by Agreement States used unsealed sources. Selected facilities using other than just
sealed sources were asked to complete the questionnaire. The questionnaire is presented in
Appendix G.
Based on the sample proportions of unsealed source sites in each strata, it is estimated
that there are approximately 6,200 facilities using only unsealed sources.
The final result of the sample selection procedure was a nearly proportional
representation (approximately 6 percent) of the estimated number of unsealed source sites in
each strata. Information was obtained from 367 sites, with 169 from the NRC strata and 198
from the Agreement State strata. These sample sizes result in approximately equal sampling
weights for sample facilities in each strata. Due to the equal weighting of the selected
sample facilities, the sample is considered to be "self-weighting" in the statistical analysis
below.
4-3
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Some or all of the following information was obtained through EPA's questionnaires:
• The emission rate or annual usage of each radionuclide to calculate the annual
amount released;
• The size of the building (maximum length, width, and height), which
influences the dispersion pattern;
• The distance and direction to the receptor (both a resident and the closest
office, school, business, or classroom) and the distances to the locations where
vegetables, milk, and meat are produced (farms, not restaurants or stores).
These factors influence the dose received through various pathways; and
• Information regarding the height, diameter, and flow rate of the stacks or
vents from which the radioactivity is released.
In addition, data regarding the frequency the wind blows from a given direction and
its average speed for each of 16 sectors (e.g., N, NNE, NE, ...) were obtained from the
National Oceanic and Atmospheric Administration (NOAA). This is called a wind rose and,
together with the dimensions of the building from which the radionuclides are released, is
used to determine the radionuclide concentrations in air at the receptor locations.
The cases studied were based upon specific data and assumptions. The emission rates
were either the measured values supplied by the facility on the survey form or were based
upon the actual amount of radioactive material used at the facility as indicated on the survey
form. The product of the actual amount of each radionuclide used during a one-year period
and a release fraction gives the estimated release rate. The release fractions used were those
given in "A Guide for Determining Compliance with the Clean Air Act Standards for
Radionuclide Emissions from NRC-Licensed and Non-DOE Federal Facilities" (EPA89a). If
the respondent indicated that effluent controls (HEPA filters,, charcoal filters, etc.) were
used, then the emissions estimated using the release fractions were reduced by the factors
given in EPA89b for the various effluent controls. The only exception to this was the use of
a release fraction of 0.01 (1.0 percent) for xenon at nuclear pharmacies. The Food and Drug
Administration limits the leakage of xenon to 0.5 percent per day (Mu91).
The meteorological information from the closest location having terrain similar to the
site was used. Data from 453 weather stations in the United States were available to
generate wind roses to use in COMPLY (NOAA90).
4-4
-------
The closest receptor was located at the distance and direction indicated by the survey
form. If the closest receptor was a resident not living in the building releasing the
radionuclides, the receptor's source of vegetables was taken to be at the location of the
receptor's home. The receptor's source of milk and meat was located at the closer of either
the distance indicated by the survey form or a default value of 2,000 m. If the closest
receptor was in an office, school, or business, or if the receptor lived in the building where
the release occurred, the sources of vegetables, milk, and meat were taken to be the closer of
either the distance given by the survey form or 2,000 m.
The building dimensions and stack parameters used were those supplied by the survey
form. If there were no offices or residences in the building from which the release occurs,
then COMPLY does not need stack information unless there is a tall stack (greater than
2.5 times the building height). If the stack is less than 2.5 times the building height,
COMPLY treats the release as a ground-level release and applies modified Gaussian plume
or empirical models to estimate dispersion.
If the closest receptor was in an office, school, or business (as opposed to a
residence), an occupancy factor of 0.3 was applied. The value of 0.3 is based upon 10 hours
per day, 5 days per week, 52 weeks per year (10x5x52/8760=0.3). If the closest receptor
was in a classroom at a college or university, an occupancy factor of ,0.1 was applied. The
value of 0.1 is based upon 20 class hours per week, 45 weeks per year (20x45/8760=0.1).
The reported dose is the larger of the calculated dose to the closest resident, 0.3 times
the calculated dose to someone in the nearest office, school, or business, or 0.1 times the
calculated dose received in a college classroom.
4.3 METHODS FOR EVALUATING DATA
Radioactive releases from a facility may contribute to radiation exposure through
several external and internal exposure pathways. External exposures may result from direct
cloud immersion or from radionuclides deposited on the ground. Internal exposure may
result from inhalation of airborne radioactivity or from ingestion of contaminated food
products. The magnitude of public exposure from a facility is largely determined by the
quantity of specific radionuclides contained in the airborne emissions and by the atmospheric
dispersion and deposition processes.
4-5
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Computer codes are commonly used to model dispersion and deposition processes that
determine human exposure. EPA has developed the COMPLY computer program to
estimate doses from radionuclide emissions to the air. The following documents provide
more information about COMPLY:
• EPA 520/1-89-001, "BID Procedures Approved for Demonstrating Compliance
with' 40 CFR Part 61, Subpart I"
• EPA 520/1-89-002, "A Guide for Determining Compliance with the Clean Air
Act Standards for Radionuclide Emissions from NRC-Licensed and Non-DOE
Federal Facilities"
• EPA 520/1-89-003,. "User's Guide for the COMPLY Code"
COMPLY is an air-dispersion code. That is, it takes estimated or measured airborne
effluent release rates, calculates the amount by which the radioactivity is diluted as it is
carried by the wind, and estimates air, ground, plant, and animal radionuclide concentrations
at various distances from the release point. From these concentrations, COMPLY calculates
the radiation dose resulting from immersion, ingestion, inhalation, and exposure to ground
contaminated by deposition of airborne radioactivity.
COMPLY has several levels of complexity. As the complexity increases, the
estimates become more realistic, and more information is required to run the code. All cases
in the random survey of licensees were run using the most realistic level (Level 4).
4.4 RAW RESULTS OF THE SURVEY
4.4.1 Results <
NRC's programs have been evaluated based on the calculated maximum individual
doses resulting from the operation of licensed facilities. Maximum individual doses were
calculated using COMPLY with input from EPA's questionnaires. The highest estimated
dose is 8 mrem/yr for all nuclides and 0.67 mrem/yr for iodine. Table 4-2 presents the
number of facilities having doses in various ranges. Seven facilities have doses above
1 mrem/yr, and none has doses above 10 mrem/yr. These doses are below the limits
established by the NESHAP. Section 4.5 contains a statistical analysis of these results.
4-6
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Table 4-2. Number of facilities having doses in various ranges.
Maximum Individual Effective
Dose Ecjaivaieflt (mreta/yr)
1E-13 to 1E-12:
1E-12 to 1E-11
1E-11 to 1E-10 .
1E-10 to 1E-09
1E-09 to IE-OS
1E-08 to 1E-07
1E-07 to 1E-06
1E-06 to IE-OS
lE-05to 1E-04
1E-04 to 1E-03
IE-OS to 1E-02
1E-02 to 1E-01
1E-01 to 1.0
1.0 to 10
> 10
Total
From AS
Nttdides
0
0
3
2 .
4
7
16
29 i
56
79
82
66
16
7
0
367
fsom
Radioiodiae
1
2
,7
24
18
23
30
47
33
36
32,
28
9
0
P
290
4.4.2 Translation from Dose to Risk
The EPA standard for NRC licensees under Subpart I is in terms of effective dose
equivalent, a system of dose estimation recommended by the International Commission on
Radiation Protection (ICRP). EPA adopted this system because it is simple, related to risk,
and widely accepted by leading national and international advisory bodies.
EPA's past risk models differ slightly from those underlying the ICRP
recommendations, primarily due to advances in the field of radiation risk since the ICRP
recommendations were published. As a result, the risks calculated by EPA are not strictly
proportional to the ede derived using the ICRP quality factors and organ weighting factors.
4-7
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While the risk methodology underlying the ICRP ede differs from that used by EPA in the
past, EPA believes that 3 mrem/yr ede is approximately equal to a lifetime individual risk of
1 in 10,000.
4.4.3 Assumptions
Because assumptions can have a significant effect on the outcome of a study, those
made in running the COMPLY code are discussed in greater depth in Appendix H.
4.4.4 Population Dose Estimates
As discussed in Section 1.2, the multi-factor approach adopted by EPA for
determining whether the emissions from a given source category are safe with an ample
margin considers both the total incidence of health effects and the distribution of the risk
across all individuals in the exposed populations in conjunction with the risk to the maximally
exposed individuals. In the BID supporting the 1989 rulemaking (EPA89), the fatal cancer
incidence for the NRC-licensed source category was given as 0.2 deaths/yr, and 99 percent
of the exposed population (the entire U.S. population, assumed to be 240 million) was
estimated to be at a risk level of less than 1E-06. The data obtained from the Random
Survey have been examined to determine whether they are consistent with these population
risk estimates. ' , • •
As indicated in Table 4-2,, no facilities were estimated to produce doses in excess of
10 mrem/yr. This would indicate that very few, if any, individuals have been exposed to a
lifetime risk substantially in excess of 1E-04, as discussed below.
Table 4-2 also reveals that several facilities have produced doses in excess of
approximately 0.01 to 0.1 mrem/yr, which is associated with a lifetime risk of cancer on the
order of 1E-06. Using 0.03 mrem/yr as the dose associated with 1E-06 lifetime risk of
cancer, 52 of the 367 facilities evaluated may have emissions associated with risks in excess
oflE-06.
The population dose was estimated for each of the facilities in the sample where a
person received a dose greater than 0.03 mrem/yr. The calculation was carried out by
finding the distances at which the sector-averaged doses fell in the ranges of 0.03 to 0.3,
4-8
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0.3 to 3.0, and 3.0 to 10 mrem/yr. The numbers of people in the annul! defined by these
i
distances were estimated using Census Bureau data (CB88). The number of people at the
various levels of exposure is shown in Table 4-3.
Table 4-3. Population dose estimates.
Dose{3SB^ni/yr>
0.03 to 0.3
0.3 to 3
3 to 10
Ap^timteXMt
1E-06 to IE-OS
IE-OS to. 1E-04
1E-04 to 1E-03
EstinsM Namber
of People ia
Sample of 30?
2,000
89
1
Estimated. Jfember
of People &r 6,200
BKalUies *
34,000
1,500
17
The estimated number of people at each level of exposure for the 6,200 facilities is
6,200/367 (= 17) times the number in the sample of 367. The estimated number of cancer
deaths is about 0.3 per year, and more than 99 percent of the population is at. a risk level of
less than 1E-06. These estimates are consistent with the estimates in the 1989 BID (EPA89).
4.5 STATISTICAL INTERPRETATION OF THE RESULTS
The principal objective of the Random Survey design was to answer the following
question: "What is the value of X such that, with at least 95 percent assurance, the 99th
percentile of the distribution of doses from these facilities does not exceed X mrem/yr, where
X mrem/yr is the highest dose estimated for all the facilities in the sample?" A second
objective was to estimate other percentiles of dose based on the statistics derived from a
fitted dose distribution. Finally, models fitted to the sample distribution of exposures permit
extrapolation of the fitted curves out to 10 mrem/yr and beyond.
Previous analyses of the maximum dose to the public from NRC-licensed facilities
other than nuclear power reactors relied on the analyst's judgment in selecting the facilities
most likely to have high exposures. This current analysis was designed to reduce the
uncertainty inherent in these judgments by using random sampling methods to provide
additional information about the population distribution of doses to maximally exposed
individuals at these sites.
4-9
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Extrapolating the results of this study to the entire population of the NRC-licensed
facilities other than nuclear power reactors involves three assumptions:
1. All facility estimates are based on running the COMPLY code and thus depend
both on collecting appropriate data on emissions and nearby individuals from
each facility and on the ability of the code and its user to model the maximum
individual exposure based on these data. All the data must be submitted,
interpreted, and used in a similar manner.
2. Non-parametric estimates of population parameters, such as using the sample
maximum to provide an upper bound on the 99th percentile of the population
distribution or extrapolating sample percentiles to the population, depend on
the representativeness of the selected sample.
3. Parametric estimates of the population parameters, such as the arithmetic or
geometric mean, depend on assumptions concerning the specific mathematical
form of the population distribution. Parametric estimates for the upper
percentiles of the population distribution are least robust to departures from the
assumed probability distribution.
The sources of uncertainty associated with these three assumptions are difficult to
quantify.
The maximum individual dose estimates cited in Table 4-2 from all nuclides and from
radioiodine only are summarized in Tables 4-4 and 4-5, respectively. Table 4-4 provides the
range, median, arithmetic and geometric mean, as well as distribution percentiles of the
effective dose equivalents for the 367 facilities in the Random Survey that use unsealed
sources. Table 4-4 summarizes results for doses from all radionuclide sources, and
Table 4-5 summarizes radioiodine doses for the 290 sample facilities using radioiodines.
Sample doses in Part A of Table 4-4 range from 2.3E-11 mrem/yr up to 8 mrem/yr,
with a median dose of 6.9E-04 mrem/yr. The geometric mean is below the median, at
4.4E-04 mrem/yr, while the arithmetic mean is significantly higher {han the median and
geometric mean, at 9.1E-02 mrem/yr.
Examination of the estimated percentiles of the dose distribution in Part B of
Table 4-4 supports the following conclusions, based on the use of the sample distribution
percentiles to provide unbiased point estimates of the population percentiles:
4-10
-------
Table 4-4. Estimated distribution of maximum individual doses.
A. Selected cfeaeacteristics of the rrandom sample ,
i ' ' Sample Charaeteristie
Minimum Dose
Geometric Mean
Median Dose
Arithmetic Mean
Maximum Dose
Sample Size
Vahi&
2.3E-11 mrem/yr
4.4E-04 mrem/yr
6.9E-04 mrem/yr
9.1E-02 mrem/yr
8.0E+00 mrem/yr
367
B, Selected petcentiles of the estimated, ctose distribufiott
Sample or Population '
Pereaafile '.
10
20
30
40
50
60
70
80
90
95
99.0
99.7
Dose
{ffireja/yr}
1.6E-06
1.7&05
9.2E-05
2.7E-04
6.9E-04
2.0E-03
5.2E-03
1.6E-02
5.9E-02
2.0E-01
3.9E+00
8.0E+00
JEsfimated Nnmter of Faeflities at &e Exeeediag
Tins Dose;
In Sample
331
294
' 257
220
184 ""
147
110
73
36
18
3
... 1
la Populatioa
5,538
4,922
4,307
3,692
3,077
2,461
1,846
1,231 .
615
308
62
18
4-ii
-------
Table 4-5. Estimated distribution of maximum individual doses for radioiodine.
A» Selected characteristics of the radioiodine dose random sample
Sample Characteristic.
Minimum Dose
Geometric Mean
Median Dose
Arithmetic Mean
Maximum Dose
Sample Size
Value
1.9E-13
6.1E-06
8.1E-06
1.3E-02
6.7E-01
290
',• f :: S$J* : J f
' B. Selected percentiles of the estimated radioiodine dose distribution
Sample or Population
Percentile ;
10
20
30
40
50
60
70
80
90
95
99.0
99.7
% Dose
{jroensfyr): ;
8.0E-10
3.9E-08
2.4E-07
1.8E-06
8.1E-06
4.2E-05
3.4E-O4
2.4E-03
2.0E-02
6.0E-02
3.9E-01
6.7E-01
Estimated Number of Facilities at or Exceeding
- - ThisDoaei
- In Sample
261
232
203 ,
174
145
116
87
58
29
14
3
1
-1& Population
4,376
3,890
3,403
2,917
2,431
1,945
1,459
972
486
243
49
19 ,
4-12
-------
1. Doses from all sources at over half of the facilities in the population are below
0.001 mrem/yr.
2. The 95th percentile of the dose due to all sources is estimated to be
0.20 mrem/yr. This dose is exceeded by 18 (approximately 5 percent) of the
sample facilities. We estimate that there are approximately 310 facilities in the
population exceeding this level of dose.
3. The 99th percentile of the dose due to all sources is estimated to be
3.9 mrem/yr. This dose is exceeded by three (approximately 1 percent) of the
sample facilities. We estimate that there are approximately 60 facilities in the
population exceeding this dose level.
Each of these point estimates has an associated uncertainty region. The maximum
dose at any of the 367 sample facilities is 8 mrem/yr, indicating that there is more than
95 percent assurance that the 99th percentile of the dose distribution for the entire population
of facilities is below 8 mrem/yr, regardless of the form of the population distribution (G178).
As noted in (3) above, the expected value of the 99th percentile is 3.9 mrem/yr. There is
over 95 percent assurance that the true 99th percentile of the population is less than a factor
of 2.1 greater than this point estimate.
Sample radioiodine doses in Part A of Table 4-5 range from 1.9E-13 mrem/yr up to
0.67 mrem/yr, with a median dose of 8.1E-06 mrem/yr. The geometric mean is slightly
below the median, at 6.1E-06 mrem/yr, while the arithmetic mean is significantly higher than
the median and geometric mean, at1.3E-02 mrem/yr. . . . .
Examination of the estimated percentiles of the iodine dose distribution in Part B of
Table 4-5 yields the following conclusions based on the use of the sample distribution to
provide unbiased point estimates of the population percentiles:
1. Doses at over half of the population of facilities using iodine sources are below
l.OE-05 mrem/yr.
2. The 95th percentile of the dose due to iodine sources is estimated to be
0.06 mrem/yr. This dose is exceeded by 14 (approximately 5 percent) of the
sample facilities using iodine sources. We estimate that there may be
approximately 250 facilities in the population exceeding this level of dose.
3. The 99th percentile of the dose due to iodine sources is estimated to be
approximately 0.4 mrem/yr. This dose is exceeded by three (approximately
1 percent) of the sample facilities using iodine. We estimate that there may be
approximately 50 facilities in the population exceeding this iodine dose level.
4-13
-------
These point estimates of the iodine dose distribution have an associated uncertainty
region. The maximum dose at any of the 290 sample facilities using iodine sources is
0.67 mrem/yr, indicating that there is more than 95 percent assurance that the 99th percentile
of the iodine dose distribution for the population of facilities using iodine sources is below
0.67 mrem/yr, regardless of the form of the population distribution (G178). As noted in (3)
above, the expected value of the 99th percentile is 0.4 mrem/yr. There is over 95 percent
assurance that the true 99th percentile of the population is less than a factor of 1.7 greater
than this point estimate.
In the following discussions, additional information concerning the distribution of
maximum individual dose from all sources at all facilities, and for the distribution of
maximum individual iodine dose at all facilities using iodine sources, is provided by
graphical analysis of the sample distributions. The empirical sample distributions are
compared to fitted models from the lognormal distribution and the hybrid lognormal (HLN)
distribution.
4.5.1 Frequency Distribution Analysis
The frequency distribution of sample doses for all sources is graphed in Figure 4-1,
which also shows a lognormal distribution fitted to the data. The vertical bars on this figure
show the histogram (bar graph) of base 10 logarithms of the dose estimates at each site.
This histogram of the logarithms of the estimated dose would have the standard normal
"bell-curve" shape of the fitted distribution if the underlying population distribution were
lognormal. Some depletion in the right tail of the sample distribution is evident above
0.1 mrem/yr; otherwise, the data appear to be approximately lognormally distributed from
this perspective. There are no obvious extreme outliers in the sample data.
A similar graph showing the distribution of iodine doses and a fitted lognormal model
is presented in Figure 4-2. The lognormal model appears less appropriate in this case. The
large "shoulder" in the sample distribution near 0.1 mrem/yr gives way to a sudden depletion
in the right tail above 0.1 mrem/yr. As a result, the lognorinal curve underestimates the
sample distribution in the shoulder region and overestimates the sample distribution in the
upper tail. . - .
4-14
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4-16
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Distributions of the type shown in Figure 4-2 are often encountered in the analysis of
dose distributions (EPA84a). The HLN model (EPA84a, Ku81, Ne82) was developed to
better fit the depleted upper tail in these distributions. One argument for the HLN
distribution is that in the absence of regulations that restrict maximum exposures, the
observed distribution of doses would probably be lognormal. Due to the existence of
dose-limiting regulations, exposures in the upper tail are "moved down" through active
control measures to below the legal threshold, thus depleting the upper tail without changing
the general shape of the lower tail. This may lead to a "shoulder" of the type shown in
Figure 4-2. In the HLN model, the upper tail is modeled as a normal distribution, and the
lower portion of the distribution is modeled as a lognormal. The mixing parameter is
defined as "rho" (rho > 0). If the random variable is X, the HLN distribution is
predominately a normal distribution above rho»X = 1 and predominately a lognormal
distribution below rho»X = 1; i.e., X = 1/rho is the boundary.
Figure 4-3 compares the fitted lognormal and HLN density functions to the sample
iodine dose distribution. The rho parameter was estimated to be 7.7 (see below), indicating
that the normal model becomes predominant at approximately 0.1 mirem/yr. The HLN
model appears to fit better in the shoulder and upper tail regions; however, there is equal
lack of fit in the middle and lower tail regions for both models.
4.5.2 Cumulative Distribution Analysis '
Figure 4-4 shows the cumulative sample distribution function and the cumulative
distribution of the fitted lognormal distribution for the dose from all sources. The
cumulative distribution function plots the percentage of facilities with dose less than or equal
to level X. At this scale, the lognormal model appears to fit well. However, in the enlarged
view of the upper tail provided by the graph in Figure 4-5, the lognormal model appears to
overestimate at the nine highest dose values observed in the sample. Note that the graph in
Figure 4-5 does not use a logarithmic scale, which tends to obscure the upper tail region.
Also, the vertical axis is defined as the percentage of facilities exceeding a given dose on the
X-axis.
Figure 4-6 shows the cumulative sample distribution function, and the cumulative
distribution of the fitted lognormal distribution, for the dose from iodine sources. At this
scale, the lognormal model appears to fit fairly well, except in the uppermost tail region. In
4-17
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4-21
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the linear scale graph of Figure 4-7, the lognormal model appears seriously to over-estimate
the 19 highest iodine dose values observed in the sample. The graphs in Figures 4-4 through
4-7 show that the density function and the cumulative distribution function graphed on a
logarithmic scale may obscure the lack-of-fit of the lognormal distribution, particularly in the
upper tail. One approach to this problem is the use of a normal probability scale for the
vertical axis. The advantage of such transformations is that the data should appear as a
straight line when normal probability scales are used, if the selected model is appropriate.
Deviation from a straight line, an indication of lack of fit, is easy to observe in graphs of this
type.
If a lognormal model is to be fitted, then, a plot could be made with the horizontal
axis transformed to a natural logarithmic scale (hi X). Alternatively, when fitting the HLN
model with mixing parameter rho, the appropriate transformation for the horizontal axis is
rho»X + ln(rho»X). Figure 4-8 shows a plot of this type, termed an HLN-probability plot,
with the X and Y axes transformed so that data from an HLN distribution would be a straight
line. With this transformation, the HLN-fitted line is straight, but not the lognormal line.
Note that the upper tail of the sample distribution fits the HLN (straight-line) model slightly
better at the highest few data points. The rho parameter for the HLN was estimated to be
0.14, indicating that the normal model is appropriate above approximately 7 mrem/yr, near
the highest observed data values. This accounts for the similarity of appearance of the two
models over most of the range of observed' sample values.
A similar plot for the radioiodine dose distribution is shown in Figure 4-9. In this
figure, the poor fit of the lognormal model in the upper tail is very evident. Again the data
appear to form a straight line, indicating that the HLN model is appropriate. The rho
parameter for the HLN was estimated to be 7.7, indicating that the normal model
isappropriate above 0.1 mrem/yr. In this case, the transition to normality occurs well within
the range of the observed sample data. .
Figure 4-10 compares the fitted lognormal and HLN cumulative distribution functions
to the sample dose distribution for dose from all nuclides. In this graph, the distribution for
sample doses from all nuclides appears to fit equally well to the two models, and the two
models are barely distinguishable. Figure 4-11 shows an enlargement of the extreme right
tail of this distribution, with the data for the 18 highest values denoted by "+'s." The fitted
HLN model passes nearer the highest 10 data points, which demonstrates that the fit of the
4-22
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HLN distribution is better in the extreme right tail of the sample distribution. The HLN
model departs from the lognormal model at approximately 1 mrem/yr and then approaches
zero more quickly. Alternatively, the lognormal model decreases slowly out to 10 mrem/yf
and beyond. The cumulative distribution of radioiodine doses in the sample and the fitted
lognormal and HLN models are shown in Figure 4-12. The distribution of sample
radioiodine doses appears to fit somewhat better to the HLN line. Figure 4-13 shows an
enlargement of the extreme right tail, with the data for the 20 highest values denoted by
"+'s." The fitted HLN model line is close to the highest four values, demonstrating that the
fit of the HLN distribution is much better in the extreme right tail of the sample radioiodine
distribution. The HLN model departs from the lognormal model at approximately
«
0.1 mrem/yr, and then approaches zero more quickly. Alternatively, the lognormal model
decreases slowly out to 1 mrem/yr and beyond.
'•.
The graphs in Figures 4-11 and 4-13 show the fitted lognormal and HLN models for
dose from all nuclides and for radioiodine doses, respectively. These models, fitted to the
sample distribution of exposures, permit extrapolation of the fitted curves out to 10 mrem/yr
and beyond. These estimates derived from the fitted models are presented in Table 4-6,
which contains estimates of the percentage and number of facilities exceeding 10 mrem/yr
from all nuclides and exceeding 3 mrem/yr from radioiodine nuclides. In part A of the
table, the lognormal and HLN estimates of the percentage and number of facilities with
maximum individual dose exceeding 10 mrem/yr are quite different. The lognormal model
estimates are 0.54 percent or approximately 33 facilities. Based on the analysis above, these
estimates are high, since the lognormal model appears to overestimate the size of the upper
tail. A more realistic estimate is given by the HLN model: 0.22 percent or 14 facilities may
exceed 10 mrem/yr.
Radioiodine doses are analyzed in Part B of Table 4-6. The lognormal estimates are
highlighted, since this model fits the upper tail of the sample distribution very poorly. The
HLN model estimates that less than one facility will exceed 3 mrem/yr of radioiodine dose to
the maximum individual.
Short of obtaining information and determining the doses to the public from every one
of the roughly 6,000 facilities licensed by NRC or an Agreement State, some questions are
likely always to remain. Although the HLN and lognormal models allow for the possibility
that a relatively small number of facilities may exist that exceed the NESHAP limits, no
facility studied was found to do so.
4-28
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Table 4-6. Estimated percentage and number of facilities exceeding specified
dose using the lognormal and hybrid-lognormal models.
A, Mcaetrbased estimates fat dose from aB auelides
Percent of FaeJEties Bxeeedkg 10
mtem/yr
Exjgaonnal
0.54%
HLN
0.22%
Kwjjfaer of FaciMes BjEeeedbg 1O
naeittfyf
(Owl of M53 Pacifie.es)
Logaormal
33
HOJ
14
B. Modal-based estimates fbf dose from radioiodaie
Percent M PadEties Exceedbg 3
mrsaa/yf
Lognoiujal1
1.26%
mN
< 0.0000001%
Naiaber of ^acfiSaes Bxceediog 3
mrem/yr
-------
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5. Quality Control
Several planned and systematic actions were taken to provide confidence in the quality
of the BID's results. These actions were designed to control activities affecting the quality of
the dose calculations and the quality of the technical background information document.
First, EPA prepared a sample questionnaire to assess the licensees' ability to interpret
EPA's needs and to respond with useful information. The samples were sent to a test group
of NRC licensees and the responses analyzed. Based on these samples, EPA's questionnaires
were modified to improve clarity for the formal mailings,
Second, all questionnaires received from licensees in response to the formal mailings
were logged in to provide a traceable record. Technical analysts then reviewed the
questionnaires to assure that the data submitted reflected a proper interpretation of the
questionnaire's requirements. In several instances, the review suggested that licensees may
have erred in filling out their forms. In all such cases,, the respondents were contacted to
discuss the items in question. Where appropriate, questionnaires were resubmitted with
corrected data.
» •>« \ .-•''•'. • ' ' ' .
Third, a single analyst performed the initial set of calculations to assure a consistent
approach in interpreting the respondents' questionnaires. To preclude the possibility that the
single analyst was himself misinterpreting respondents' data, two independent analysts were
asked to (a) verify the initial calculations by interpreting the data from the questionnaires and
calculating the doses from the 50 facilities yielding the highest doses, and (b) review all
assumptions made by the original analyst. ^
Finally, the entire manuscript was submitted for multi-disciplinary peer review.
5-1
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R-l
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R-2
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and Toxic Airborne Source Terms for Uranium Milling Operations,"
Regulatory Guide 3.59, March 1987, -
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Renewal of License No. SMB-920," NUREG-1027, November 1988.
R-3
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r
NRC88a U.S. Nuclear Regulatory Commission, "Final Generic Environmental Impact
Statement on Decommissioning of Nuclear Facilities," NUREG-0586, August
1988.
ORAU85 Boermer, A.J., "Radiological Measurements of Molycorp of America Plant,
York, PA," prepared for the U.S. Nuclear Regulatory Commission, Oak
Ridge Associated Universities, November 1985.
ORAU88 Berger, J.D., "Survey of Shieldalloy Corp.," prepared for the U.S. Nuclear
Regulatory Commission, Oak Ridge Associated Universities^ ORAU 88/G-79,
November 1988.
SCA84 S. Cohen & Associates, Inc., "Impact of Proposed Clean Air Act Standards
for Radionuclides on Users of Radiopharmaceuticals," prepared for the U.S.
Environmental Protection Agency, Office of Radiation Programs, under Work
Assignment #5, Contract #68-02-3853, with Jack Faucett & Associates,
October 1984.
SCA91 S. Cohen & Associates, Inc., "Additional Information on the Reconsideration
of 40 CFR 61, Subpart I: Impact on Selected NRC Licensees," draft report
prepared for the U.S. Environmental Protection Agency, Office of Radiation
Programs, under Work Assignment 1-12, March 1991.
SEG91 SEG April 2, 1991, amendment to 1990 Fourth Quarter NESHAPs Report,
letter RTS-91-001L, dated January 29, 1991.
USE89 US Ecology, Ward Valley LLRW Disposal Facility, Needles, CA, "Report on
Compliance with the Clean Air Act Limits for Radionuclide Emissions from
the COMPLY Code, Version 1.2," September 1989.
USE91 US Ecology, "A Report to Support the Application for a NESHAPs Permit for
the Emission of Radionuclides for the Central States Compact Low-Level
Radioactive Waste Disposal Facility Near Butte, Nebraska," January 1991,
Lincoln, Nebraska. '
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APPENDIX A
NRC's ORGANIZATION, REGULATIONS, AND CONTROLS
This appendix explains the origins and need for NRC and its predecessor, the
Atomic Energy Commission, dating back to the Atomic Energy Act. It
describes how NRC's organization promotes the discharge of its responsi-
bilities and its ability to fulfill its legislative charter. Regulations and effluent
controls for NRC-licensed facilities other than nuclear power reactors are
described.
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Contents
A.1 Organization and Responsibilities of NRC . . . . - A-3
A.I.I Basic Functions A-3
A.1.2 Organization A-4
A.2' Controls Applicable to Licensees - General A-9
A.2.1 Establishing Airborne Emission Controls A-9
A.2.2 Licensing Program . A-ll
A.2.3 Airborne Emissions Monitoring A-17
A.2.4 Inspection Programs A-22
A.2.5 Enforcement Programs A-23
A.3 Controls Applicable to Airborne Emissions ; A-23
A.4 References A-24
A-2
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APPENDIX A
NRC's ORGANIZATION, REGULATIONS, AND CONTROLS
A.1 ORGANIZATION AND RESPONSIBILITIES OF NRC
NRC regulates the civilian uses of source, byproduct, and special nuclear materials
and nuclear reactors in the United States. This mission is accomplished through the develop-
ment and implementation of controls (i.e., rules, regulations, guidance, etc.) governing
licensed activities; licensing of nuclear facilities (i.e., issuance of permits and licenses) and
the possession, use, and disposal of nuclear materials; and inspection and enforcement
activities to ensure compliance with these controls and the conditions imposed through
permits and licenses. , ,
A.I.I Basic Functions
NRC's responsibilities include protecting public health and safety; protecting the
environment; protecting and safeguarding materials and plants in the interest of national
security; and ensuring conformity with antitrust laws. During fiscal year 1990, NRC had
approximately 3,200 employees and a budget of over $400 million to carry out three basic
functions: regulatory research and standards development, licensing, and inspection and
enforcement.
As part of its regulatory research and standards development function, NRC is
mandated by law to conduct an extensive confirmatory research program in the areas of
safety, safeguards, and environmental assessment. The Commission establishes regulations,
standards, and guidelines governing the various licensed uses of nuclear facilities and
materials. ,
In its licensing function, the agency reviews and issues licenses for the construction
and operation of nuclear power plants and other nuclear facilities, and it licenses the
possession and use of nuclear materials for medical, industrial, educational, research, and
other purposes. Regulatory authority for certain nuclear materials licensing has been
transferred to certain States under the Agreement States Program authorized by the AEA.
However, NRC retains authority for licensing and regulating nuclear reactors.
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NRC's inspection and enforcement activities include various kinds of inspections and
investigations designed to ensure that licensed activities are conducted in compliance with its
regulations and other requirements. NRC enforces compliance as necessary.
A.1.2 Organization
A. 1.2.1 The Commission. The Commission is composed of five members, appointed by the
President and confirmed by the Senate, one of whom the President designates as Chairman.
The Chairman is the principal executive officer of, and the official spokesman for NRC, as
mandated by the Reorganization Plan No. 1 of 1980 (NRC90). The Advisory Committee on
Reactor Safeguards (ACRS), which was assigned a statutory role by Congress, independently
reviews and reports on safety studies and applications for construction permits and operating
licenses. The ACRS advises the Commission with regard to hazards at proposed or existing
reactor facilities and the adequacy of proposed reactor safety studies. On its own initiative,
the ACRS may review specific generic matters or nuclear facility safety issues.
A. 1.2.2 NRC Offices. NRC reorganized in 1987 to reflect progressively less involvement
with the construction of large, complex nuclear facilities and increased involvement with the
operation and maintenance of these facilities.
Office of Nuclear Reactor Regulation (NRR). The primary responsibilities of this
Office are to conduct the inspection and licensing activities associated with operating power
reactors, including contractors and suppliers for such facilities. The Office also is
responsible for evaluating applications to build and operate new power reactors,'for
inspection and licensing activities related to the construction and operation of research and
test reactors, and for licensing reactor operators. In addition, the Office is responsible for
inspecting NRC-licensed activities under its jurisdiction to ensure that they comply with all
NRC regulations and requirements.
Except for research and test reactors, this Office has no responsibilities for NRC-
licensed facilities other than nuclear power reactors.
Office of Nuclear Material Safety and Safeguards flSMSSX All non-reactor NRC
licenses are regulated by the Office of Nuclear Material Safety and Safeguards (NMSS).
NMSS's responsibilities fall into six principal areas: (1) licensing of nuclear fuel cycle
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facilities, (2) licensing of nuclear materials for uses other than in reactors, (3) regulation of
the transportation of nuclear materials, (4) safeguarding of nuclear materials from sabotage
and diversion to unauthorized uses, (5) regulation of radioactive waste disposal facilities, and
(6) regulation of the decommissioning of previously licensed nuclear facilities that are no
longer in use. Some of these functions are carried out by the five NRC Regional Offices.
The various processing operations required to produce fuel for nuclear reactors are
conducted in NRC-licensed fuel cycle facilities. Activities at these facilities include: certain
types of uranium mining activities, milling and refining uranium ore to produce uranium
concentrations, production of uranium hexafluoride from uranium concentrates to provide
feed material for isotopic enrichment of U-235 to levels needed for a nuclear reaction,
isotopic enrichment processing of uranium hexafluoride to produce fuel with a higher
percentage of U-235 than in natural uranium, fabrication of nuclear reactor fuel, and
reprocessing spentefuel for recycle.1
Most of the manufacturing operations that make up the nuclear fuel cycle are licensed
by NRC. Exceptions are uranium mining., uranium milling in Agreement States, and
enrichment by the U.S. Department of Energy. "NMSS reviews operational safety, radiation
protection, and criticality safety programs as part of the licensing process for fuel cycle
facilities. NMSS also provides policy guidance and technical support to Agreement States on
their licensing and inspection activities and on emergency responses. At present, NRC fuel
cycle licenses number about 30.
NRC regulates approximately 8,200 licenses for the possession and use of radioactive
materials for purposes other than the generation of electricity or operation of a research
reactor. The 28 Agreement States regulate about 15,000 radioactive materials licenses.
These totals include licensees authorized to possess and use radioactive materials only in the
form of sealed sources. Most of NRC's licenses are administered by NRC's Regional
Offices.
Office of Nuclear Regulatory Research (RES). This Office has three primary
responsibilities: (1) to plan, recommend, and implement programs of nuclear regulatory
research, standards development, and resolution of safety issues of facilities regulated by
1 The latter step is not being performed in the United States.
: . A-5
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NRC; (2) to develop and promulgate all technical regulations; and (3) to coordinate research
activities within and outside the agency including appointment of staff to committees and
conferences.
With respect to air emissions from NRC-licensed facilities other than nuclear power
reactors, this Office is responsible for the promulgation and revision of regulations affecting
emissions, such as 10 CFR Part 20. Additionally, the Office manages the development of
regulatory guides.
Office for Analysis and Evaluation of Operational Data f AEODX This Office
independently analyzes and evaluates operational safety data associated with NRC-licensed
activities to identify issues that require action by NRC or the industry. Its other
responsibilities include the reactor performance indicators program and the management and
direction of programs for diagnosing evaluations and investigations of significant operational
events.
With respect to air emissions from NRC-licensed facilities other than nuclear power
reactors, this Office evaluates semiannual plant airborne emissions data and unusual events
that contribute to airborne emissions.
\
Office of Enforcement. This Office develops policies and programs for enforcement
of NRC's requirements. It manages major enforcement actions and assesses the effectiveness
and uniformity of enforcement actions taken by the Regional Offices. Enforcement powers
include notices of violation, fines, and orders for license modification, suspension, or
revocation.
. / .
Regional Offices. NRC's five Regional Offices execute the established NRC policies
and assigned programs relating to inspection, enforcement, licensing, State agreements, State
liaison, and emergency response within each region. Each regional division inspects and
evaluates assigned NRC programs. For Part 70 licensees, NRC's Resident Inspector
Program is applicable for assigned facilities. The Division of. Radiation Safety and
Safeguards performs inspections and evaluations in radiological safety and environmental
monitoring. . -
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A.l.2.3 State Programs. Prior to enactment of the Atomic Energy Act of 1954, nuclear
energy activities in the United States were largely confined to the federal government. The
Act made it possible for private commercial firms to enter the field for the first time.
Because of the hazards associated with nuclear materials, Congress determined that these
activities should be regulated under a federal licensing system to protect the health and safety
of workers in the nuclear industry and the public. NRC is the federal agency charged with
this responsibility.
Although protection of the public health and safety has traditionally been a State
responsibility, the Atomic Energy Act of 1954 did not specify such a role for the States in
nuclear matters. This policy was changed in 1959 when Congress enacted Section 274 of the
Atomic Energy Act. Section 274, spells out a State role and provided a statutory basis under
which the federal government can relinquish to the States portions of its regulatory authority.
The 1959 amendment made it possible for the States to license and regulate byproduct
material (radioisotopes), source material (the raw materials of atomic energy), and small
quantities of special nuclear material.2 The Commission is required, however, to retain
regulatory authority over the regulation of nuclear facilities vital to the national common
defense and security and certain types of radioactive wastes. The Atomic Energy Act was
amended in 1978 by the passage of the Uranium Mill Tailings Radiation Control Act
(UMTRCA) of 1978 which requires NRC Agreement States regulating uranium and thorium
tailings or wastes resulting from recovery operations to adopt certain technical and
procedural requirements. The 1978 amendment also requires NRC to review periodically
Agreement State programs for adequacy and compatibility.
Section 274j of the Atomic Energy Act allows NRC to terminate its agreement with a
State if the Commission finds that such termination is necessary to protect the public health
and safety. In 1980, Section 274j was amended to authorize the Commission to suspend
temporarily all or part of an agreement with a State in the case of an emergency situation
where the State failed to take necessary action. Such suspensions may remain in effect only
for the duration of the emergency.
2 In 1981, the Commission amended its Policy Statement, "Criteria for Guidance of States and NRC in
Discontinuance of NRC Authority and Assumption Thereof by States Through Agreement" to allow a State to
seek an'amendment for the regulation of low-level radioactive waste as a separate category.
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The mechanism for the transfer of NRC authority to a State to regulate the
radiological health and safety aspects of nuclear materials is a formal agreement between the
Governor of the State and the Commission. Criteria for such agreements have been
published by NRC as a Policy Statement in the Federal Register. Before actually signing the
document, the Commission, by statute, must determine that the State's radiation control
program is compatible with the Commission's, meets the applicable parts of Section 274, and
is adequate to protect the public health and safety. For its part, the State establishes its
authority to enter such an agreement by passing enabling legislation.
At present, 28 States have entered into such agreements with NRC.3 These States
now regulate over 65 percent of the 24,000 licensees for byproduct, source material, and
special nuclear material in the United States. In 1981, the Commission determined that
qualified States may also enter into limited agreements for regulation of low-level waste in
permanent disposal faculties.
Each agreement provides that the State will use its best efforts to maintain continuing
compatibility with the NRC's program. NRC maintains a continuing relationship with each
Agreement State to assure continued compatibility of the State's regulatory program and its
adequacy to protect health and safety. This relationship includes: exchange of current
information covering regulations, licensing, inspection and enforcement data; consultation on
special licensing, inspection, enforcement, and other regulatory problems; and an annual
meeting of all Agreement States to consider regulatory matters of common interest. Special
technical assistance is routinely provided to the States upon request.
As mandated by the Atomic Energy Act, NRC conducts onsite, in-depth program
reviews periodically in each Agreement State. This review covers organizational,
administrative, personnel, regulatory, licensing, compliance, and enforcement program areas.
Selected Agreement State licensing and compliance casework is reviewed in detail. State
inspectors are accompanied by NRC staff on selected inspections of State licensees. A copy
of the guidelines that NRC uses in conducting such reviews have been published in the
Federal Register as a Commission Policy Statement. t
3 Alabama, Arizona, Arkansas, California, Colorado, Florida, Georgia, Illinois, Iowa, Kansas, Kentucky,
Louisiana, Maryland, Mississippi, Nebraska, Nevada, New Hampshire, New Mexico, New York, North
Carolina, North Dakota, Oregon, Rhode Island, South Carolina, Tennessee, Texas, Utah, and Washington.
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NRC provides training for Agreement State personnel. Examples are short-term
courses in health physics, radiography radiation safety, nuclear medicine, licensing,
inspection procedures, radiological engineering, well-logging, transportation of nuclear
materials, and project management for the licensing of low-level waste disposal facilities.
The NRC State Agreements Program is administered by the Office of State Programs.
NRC Regional Offices participate in implementation of the State Agreements Program.
As a rule of thumb, one to one-and-a-half staff-years per 100 licenses is needed for
effective administration of the program assumed from NRC. This is a general index, and
actual staffing needs will vary according to the particular circumstances in any given State.
i
The Agreement State experience since 1962, the year of the first State agreement, has
been that the States generally conduct effective radiation control programs. When NRC
notes major program deficiencies, NRC (with its' resources) offers technical advice,""
assistance, and training. The main area of concern is maintaining adequate staffing levels, a
reflection of State salary structures and funding. On the other hand, Agreement States
typically excel in having highly trained staff and in conducting more frequent inspections
than NRC.
A.2 CONTROLS APPLICABLE TO LICENSEES - GENERAL
A.2.1 Establishing Airborne Emission Controls
This section describes NRC's procedures for setting facility controls to protect
the health and safety of the public. These controls may take several forms: rules and
regulations; regulatory guides; generic letters, bulletins, and information notices; and NRC
reports. The first two categories of controls for facilities are administered by the Office of
Nuclear Regulatory Research (RES); the others are administered by the Office of Nuclear
Reactor Regulation (NRR).
A.2.1.1 RulemaMng and Regulatory Guides. The term mlemaking actually covers the
establishment of two kinds of regulatory documents - the regulations-of NRC contained in
Title 10 of the Code of Federal Regulations (10 CFR) and regulatory guides. The decision
to move forward with either a rule or a regulatory guide is based upon the results of a
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regulatory analysis (itself based upon a Technical Findings Document [e.g., NUREG]).
Thereafter, both types of documents, rules and guides, are developed in a process that
provides for internal and external (public) review and comment. The entire process is
repeated again for the final rule or guide, developed in light of comments received from the
public.
Both types of documents are prepared in a two-step process. In the first step, a draft
is produced for public comment. RES usually develops such drafts in consultation with and
on behalf of NRR, NMSS, or both. The drafts are developed at a technical staff level,
coordinated through parallel management chains of the affected offices, reviewed by the
appropriate advisory committee (usually the ACRS except for waste management matters
which now have their own advisory committee), reviewed by a senior management review
group called the Committee for the Review of Generic Requirements (CRGR), and then
presented to the appropriate decision maker(s) for action.
.1
When the development of a rule or a guide reaches the point where it is presented to
the decision makers, the process diverges. Substantive rules can be issued for public
comment only by a majority vote of the five NRC Commissioners. Therefore, proposed
rulemakings are recommended for action by RES, with the concurrence of the affected
program office, through the NRC's Executive Director for Operations, to the Commission.
The Commission requests input from the appropriate advisory committees and the CRGR to
assist in its decision.
Once the Commission has decided to issue a proposed rule for public comment, a
notice of the proposed action is issued in the Federal Register, the notice also identifies the
time allowed for comments and may specify particular questions on which the Commission
desires input. These particular questions often involve the matters treated in the regulatory
analysis performed for the proposed rule; e.g., the anticipated costs and other impacts of ,
imposing the new rule.
The RES staff, in consultation with the affected program office, evaluates public
comments received on a proposed rule. The Commission has used both rulemaMng hearings,
which are formal adjudicatory proceedings, and public meetings, which are less formal, to
further discussion and obtain additional information concerning a proposed rule. Once the
additional information has been received and evaluated, the staff modifies the rule as
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necessary, repeats the entire review process followed for the proposed rule, and returns the
rulemaMng package to the Commission for final action. When the Commission makes its
final decision on the rule, it is issued as "effective" with a notice in the Federal Register.
The rule then becomes a part of Title 10 of the Code of Federal Regulations.
The process followed by the RES in developing a draft and then a final regulatory
guide is essentially the same as that for a rule, except that the Executive Director for
Operations and the Commission are not involved. Rather, the Director of the Office of RES
is the final decision authority for issuing regulatory guides, either in draft form for public
comment or in final form.
A.2.1.2 Generic Letters. Bulletins and Information Notices. Generic letters, bulletins, and
information notices are written NRC notifications sent to groups of licensees that identify
specific problems, developments, or other matters of interest to the licensees. In some cases,
NRC is calling for or recommending that the licensees take specific steps.
A.2.1.3 NRC Reports. NRC reports (usually referred to genetically as NUREGs) are
prepared by NRC's staff, contractors, or national laboratories and provide the technical basis
for decision making. Special categories of such reports include Safety Evaluation Reports
(SERs), Environmental Impact Statements (EISs), and Standard Review Plans (SRPs). NRC
issues the first two categories of reports to establish the conditions under which the license to
construct or operate will be issued. The SRPs are issued to disseminate information about
the regulatory licensing process and to improve the general public's and the nuclear
industry's understanding of the staffs review process.
Standard Review Plans address the responsibilities of the persons performing the
review, the matters that are reviewed, the Commission's regulations and acceptance criteria
necessary for the review, how the review is accomplished, the appropriate conclusions, and
the implementation requirements.
A.2.2 licensing Program
Licensing programs utilize a system of controls, compliance guidance, and
independent review to establish (with reasonable assurance) the ability of a facility to meet
performance requirements. Of particular relevance is NRC's ability to establish and maintain
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an acceptable level of performance through (1) independent review to verify that regulatory
criteria were correctly translated into design, construction, and operations documents and (2)
monitoring of operating data.
NRC has delegated to the five Regional Administrators licensing authority for selected
parts of its decentralized licensing program for nuclear materials. The delegated licensing
program includes authority to issue, renew, amend, cancel, modify, suspend, or revoke
licenses for nuclear materials issued pursuant to 10 CFR Parts 30 through 35\ 39, 40, and 70
to all persons for academic, medical, and industrial uses, with the exceptions of activities in
the fuel cycle and special nuclear material, sealed sources and devices design review, and
processing of source material for extracting of metallic compounds.
A.2.2.1. Part 30 Licenses. The regulations in 10 CFR Parts 30, 32, 33, 35, and 39 provide
for licensing facility categories listed in Table D-l. A license applicant is required to file an
application in duplicate on NRC Form 313, "Application for Material License," in
accordance with the instructions in 10 CFR 30.6 and 30.32. Form 313 asks a wide range of
information including: the name and mailing address of the applicant; the location of use; a
person who can be contacted about the application; the materials requested; the purpose of
use; the training and experience of the authorized users and Radiation Safety Officer; the
worker radiation safety training program; facilities and equipment; the radiation safety
program; and waste management program. The information will be transformed into license
conditions upon approval. The applicant mails the license application, with application fee,
to the NRC office identified on the form.
Because of the potential radiation hazard to workers and the public, NRC's specific
license program for regulating byproduct material use incorporates three regulatory features:
case-by-case review of applications, onsite inspections, and periodic license renewals. NRC
staff will review the application to determine whether the applicant's radiation safety program
complies with the regulations. After completing the review, if the applicant's program
appears incomplete or inadequate, NRC will issue a deficiency letter that describes the
apparent shortcomings in the applicant's program and requests clarification or correction. If
the applicant's response to the deficiency letter is satisfactory, or if no deficiency letter was
needed, NRC will issue a specific license authorizing the possession and use of byproduct
material on NRC Form 374, "Byproduct Material License."
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To help license applicants prepare the application and design their radiation safety
programs, NRC has published the following guidance documents:
Regulatory Guide 8.18
Regulatory Guide 8.21
Regulatory Guide 8.23
Regulatory Guide 10.2
Regulatory Guide 10.5
Regulatory Guide 10.7
Regulatory Guide 10.8
Draft Guide DG-8001
Draft Guide OP 212-4
NUREG-0267
Information Relevant to Ensuring That Occupational
Radiation Exposures at Medical Institutions Will Be As
Low As Reasonably Achievable
Health Physics Surveys for Byproduct Material at NRC-
Licensed Processing and Manufacturing Plants
Radiation Safety Surveys at Medical Institutions
Guidance to Academic Institutions Applying for Specific
Byproduct Material Licenses of Limited Scope
Applications for Type A Licenses of Broad Scope
Guide for the Preparation of Applications for Licenses
for Laboratory and Industrial Use of Small Quantities of
Byproduct Material
Guide for the Preparation of Applications for Medical
Programs
Basic Quality Assurance Program for Medical Use
Radiation Protection Training for Personnel Employed in
Medical Facilities
Principles and Practices for Keeping Occupational
Radiation Exposure at Medical Institutions As Low As
Reasonably Achievable
A.2.2.2. Part 40 Licenses. The regulations in 10 CFR Part 40, "Domestic Licensing of
Source Material," provide for licensing facility categories listed in Table D-l. A license
applicant is required to provide detailed information on the facilities, equipment, and
procedures to be used and an environmental report discussing the operation's impact on the
health and safety of the public and on the environment. The Commission uses this
information to determine whether the applicant's proposed activities will, among other tilings,
result in undue risk to the health and safety of the public or adversely affect the environment.
General guidance for filing an application and an environmental report is provided in Section
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40.31, "Application for Specific Licenses," of 10 CFR Part 40, and in 10 CFR Part 51,
, "Licensing and Regulatory Policy and Procedures for Environmental Protection,"
respectively.
The application must contain information specified in NRC Form 313, "Application
for Material License," which primarily addresses processing, in-plant radiation safety, and
environmental considerations. In essence, the applicant is required to submit, as part of the
license application, a Safety Analysis Report (SAR) pursuant to 40 CFR Part 190 and an
Environmental Report (ER) pursuant to 10 CFR Part 51. Based on the information provided
in these reports, NRC will in turn develop a Safety Evaluation Report (SER) and an
Environmental Impact Statement (FJS). Under 10 CFR 51.22, "Criterion for Categorical
Exclusion; Identification of Licensing and Regulatory Actions Eligible for Categorical
Exclusion or Otherwise Not Requiring Environmental Review," some licensees are not
required to prepare an Environmental Report if NRC's first finding is that the applicant's
proposed actions do not individually or cumulatively have a significant effect on the human
environment.
These licenses are generally issued for 10-year periods and are renewable over the life
of the project. License renewal applications are processed in a manner similar to that used
for new applications. Operational experience, site-specific data, and proposed continuing
activities are the primary factors considered by the NRC staff in processing renewal
applications.
To help licensees develop the application, NRC has published the following guidance
documents (a comprehensive list is provided in Appendix B):
• Regulatory Guide 3.5
• Regulatory Guide 3.8
• Regulatory Guide 3.46
• Regulatory Guide 3.51
Standard Format and Content of License Applications for
Uranium Mills
Preparation of Environmental Reports for Uranium Mills
Standard Format and Content of License Applications,
Including Environmental Reports, for In Situ Uranium
Solution Mining
Calculational Models for Estimating Radiation Doses to
Man from Airborne Radioactive Materials Resulting from
Uranium Milling Operations
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Regulatory Guide 3.55
Regulatory Guide 3.56
Regulatory Guide 3.59
Regulatory Guide 4.4
Regulatory Guide 4.15
Regulatory Guide 8.30
Regulatory Guide 8.31
NUREG/CR-2011
Standard Format and Content for the Health and Safety
Sections of License Renewal Applications for Uranium
Hexafluoride Production
General Guidance for Designing, Testing^ Operating, and
Maintaining Emission Control Devices at Uranium Mills
Methods for Estimating Radioactive and Toxic Airborne
Source Terms for Uranium Milling Operations
Radiological Effluent and Environmental Monitoring at
Uranium Mills
Quality Assurance for Radiological Monitoring Programs
(Normal Operations) - Effluent Streams and the
Environment
Health Physics Surveys in Uranium Mills
Information Relevant to Ensuring that Occupational
Radiation Exposures at Uranium Mills Will Be As Low
As Is Reasonably Achievable
MILDOS - A Computer Program for Calculating
Environmental Radiation Doses from Uranium Recovery
Operations
A.2.2.3. Part 50 (Type 104) Licenses. The licensing process begins with the filing of a
license application, consisting of general information, an Environmental Report, and a Safety
Analysis Report (SAR). The general content requirements of the SAR for a reactor are
contained in 10 CFR 50.34.
NRC initiates a comprehensive technical review of the license application and any
supporting documents after initial acceptance review and docketing. During this period,
NRC's staff and the Advisory Committee on Reactor Safeguards (ACRS) conduct
independent technical reviews of the license application, resulting in the issuance of a
Safety Evaluation Report (SER) by NRC's staff and a formal letter of recommendation from
the ACRS to the Chairman of NRC. -
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In determining whether to grant a construction permit, NRC holds an adjudicatory
public proceeding conducted by the Atomic Safety and Licensing Board (ASLB). At the end
of the adjudicatory proceeding, the ASLB renders a decision supported by a written opinion.
A decision of the ASLB could be appealed to an Atomic Safety and Licensing Appeal Board
(ASLAB). The Commissioners may also consider the matter upon a petition requesting such
review. After all avenues of administrative appeal have been exhausted and if the ASLB's
initial decision prevails, the Director of Nuclear Reactor Regulation issues a letter authoriz-
ing construction to begin. .. '. '
Prior to anticipated completion of construction, the applicant submits an updated
license application to NRC in support of obtaining a license to operate. NRC's staff and the
ACRS again conduct technical reviews which, if favorable, result in the issuance of a Safety
Evaluation Report by NRC's staff and a formal letter of recommendation from the ACRS to
the Chairman of NRC.
\ , •
A.2.2.4. Part 70 Licenses. The regulations in 10 CFR Part 70, "Domestic Licensing of
Source Material," provide for licensing facility categories listed in Table D-l. A license
applicant is required to provide detailed information on the facilities, equipment, and
procedures to be used and an environmental report that discusses the operation's impact on
the health and safety of the public and on the environment. The Commission uses this •
information to determine whether the applicant's proposed activities will, alnong other things.,
result in undue risk to the health and safety of the public or adversely affect the environment.
The license application can be filed in letter form and provides the information specified in
section 70.22, "Contents of Applications."
General guidance for filing an application and an environmental report is provided in
Section 70.21, "Filing," of 10 CFR Part 70 and in 10 CFR Part 51, "Licensing and
Regulatory Policy and Procedures for Environmental Protection," respectively. Basically,
the applicant is required to submit, as part of the license application, a Safety Analysis
Report (SAR) pursuant to 40 CFR Part 190 and an Environmental Report (ER) pursuant to
10 CFR Part 51. Based on the information provided in these reports, NRC will in turn
develop a Safety Evaluation Report (SER) and an Environmental Impact Statement (EIS).
Under 10 CFR 51.22, "Criterion for Categorical Exclusion; Identification of Licensing and
Regulatory Actions Eligible for Categorical Exclusion or Otherwise Not Requiring
Environmental Review," some licensees are not required to prepare an Environmental Report
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if NRC first finds that the applicant's proposed actions do not individually or cumulatively
have a significant effect on the human environment.
These licenses are generally issued for 10-year periods and are renewable over the life
of the project. License renewal applications are processed in a manner similar to that used
for new applications. Operational experience, site-specific data, and proposed continuing
activities are the primary factors considered by the NRC staff in processing renewal
applications.
To help licensees prepare the application, NRC has published the following guidance
documents (a comprehensive list is provided in Appendix B):
• Regulatory Guide 3.6
Regulatory Guide 3.12
• Regulatory Guide 3.25
• Regulatory Guide 4.9
Regulatory Guide 8.10
Regulatory Guide 10.3
Content of Technical Specifications for Fuel
Reprocessing Plants
General Design Guide for Ventilation Systems of -
Plutonium Processing and Fuel Fabrication Plants
Standard Format and Content of Safety Analysis Reports
for Uranium Enrichment Facilities
Preparation of Environmental Reports for Commercial
Uranium Enrichment Facilities
Operating Philosophy for Maintaining Occupational
Radiation Exposures As Low As Is Reasonably
Achievable
it
Guide for the Preparation of Applications for Special
Nuclear Material Licenses of Less Than Critical Mass
Quantities
A.2.3 Airborne Emissions Monitoring
N - ' .
During the period of operation, the licensee is subject to various terms and conditions
to ensure that activities are conducted in accordance with the design bases and performance
objectives agreed to in the license. Airborne effluent monitoring programs and inspections
are means by which NRC monitors facility operations.
A-17
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NRC regulations limiting routine radionuclide airborne emissions are contained in
10 CFR Part 20, Standards for Protection Against Radiation, which applies to all licensees.
In addition, the recent amendments to Part 20 require licensees by January 1994 to keep
exposures as low as is reasonably achievable (ALARA). ALARA means making every
reasonable effort to maintain exposures to radiation as far below the dose limits in
10 CFR Part 20 as is practical, consistent with the purpose for which the licensed activity is
undertaken. The requirement takes into account the state of technology, the economics of
improvements in relation to the state of technology, the economics of improvements in
relation to benefits to the public health and safety, and other societal and socioeconomic
considerations, and the value of utilizing nuclear energy and licensed materials in the public
interest.
A.2.3.1 Part 30 Licenses. The possible airborne radionuclide emissions are from unsealed
byproduct material on foils or plated sources or radioactive aerosols or gases in a
manufacturing facility, laboratory, or radiopharmaceutical. In general, facility design or
engineered safety features in the facility and operating restrictions or procedures would
reduce airborne radionuclide release. Use of charcoal traps or fume hoods with charcoal
filtration system or HEPA filter can significantly reduce air contamination during operations.
Incineration operations (e.g., at hospitals) must be conducted in a way that all
airborne effluent releases are reduced to levels as low as reasonably achievable (ALARA).
The primary means of accomplishing this objective is emission controls including filtration,
scrubbing, and air dilution. Discharge stacks, types and estimated composition and flow
rates of atmospheric effluents, and emissions control methods must be designed and analyzed
to limit potential releases to ALARA levels.
For medical use of byproduct materials, according to 10 CFR 35.205:
(a) A licensee that administers radioactive aerosols or gases is required to do so in a
room with a system that will keep airborne concentrations within the limits prescribed
by 10 CFR 20.106. The system must either be directly vented to the atmosphere
through an air exhaust or provide for collection and decay or disposal of the aerosol
or gas in a shielded container.
(b) A licensee is required to administer radioactive gases only in rooms that are at
negative pressure compared to surrounding rooms.
A-18
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(c) Before receiving, using, or storing a radioactive gas, the licensee is required to
calculate the amount of time needed after a spill to reduce the concentration in the
room to the occupational limits listed in Appendix B to 10 CFR Part 20. The
calculation must be based on the highest activity of a gas handled in a single
container, the air volume of the room, and the measured available air exhaust rate.
(d) A licensee is required to make a record of the calculations required in (c) that
includes the assumptions, measurements, and calculations made and shall retain the
record for the duration of use of the area. A licensee is also required to post the
calculated time and safety measures to be instituted in case of a spill at the areas of
use.
(e) A licensee is required to check the operation of reusable collection systems each
month and measure the ventilation rate available in areas of radioactive gas use each
6 months. In addition, according to 10 CFR 35.90, a licensee is required to store
volatile radiopharmaceuticals and radioactive gases in the shipper's radiation shield
and container. A licensee is also required to store a multi-dose container in a fume
hood after drawing the first dosage from it.
Airborne effluent concentration at the release point must be calculated and compared
to the appropriate value of Table n of Appendix B to 10 CFR Part 20. NRC or the
Agreement State often recommends that the license applicant use a "10 percent at the stack"
rule for the calculation. Except for medical institutions, this calculation is required to be
submitted as part of the license application under Item 10.13.3 of Form NRC-313. Medical
license applicants do not have to submit the calculations with the application, but they are
required to keep them on record for NRC (or Agreement State) review during onsite
inspections.
1 • •'
If aerosols and gases are not directly vented to the atmosphere, the license applicant
may respond with a statement that it will not directly vent spent aerosols and gases to the
atmosphere and therefore no effluent estimation is necessary. If aerosols or gases are
directly vented to the atmosphere, airborne effluent concentrations must be calculated. For
medical institutions, NRC recommends the following estimation procedure, described in
Regulatory Guide 10.8, for use in the license application:
' / ' .
(a) . Divide the total activity released to an unrestricted area (activity used each week that
is released in an exhaust system) by the total volume of air exhausted over the week
("on time" multiplied by measured airflow rate). The quotient must be less than the
applicable maximum permissible value for an unrestricted area.
A-19
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(b) If this is not the case, plan for fewer studies and do the calculation again.
Alternatively, consider collection and decay-in-storage for waste, or restriction of
access to the release point and calculation of concentration at the boundary of the
restricted area.
A.2.3.2 Part 40 Licenses. To achieve airborne emission control, facility operations must be
conducted in a way that reduces all airborne effluent releases to levels that are ALARA. The
primary means of accomplishing this objective is by means of emission controls including
ventilation, filtration, and confinement systems. Discharge stacks, types and estimated
composition and flow rates of atmospheric effluents, and emission control methods are
required to be designed and analyzed to limit potential releases to ALARA levels.
Calculations must be supplemented by stack monitoring appropriate for the planned and
potential releases. Minimum performance specifications, such as filtration or scrubber
efficiency and airflow for operating the ventilation, filtration, and confinement systems
throughout the facility, are normally determined.
Institutional controls, such as extending the site boundary and exclusion area, are also
employed to ensure that offsite exposure limits are met, but only after all practical measures
have been taken to control emissions at the source. Notwithstanding the existence of
individual dose standards, strict control of emissions is necessary to assure that population
exposures are reduced to the maximum extent reasonably achievable and to avoid site
contamination.
Effluent and environmental monitoring programs, including methods and procedures
for measuring concentrations and quantities of both radioactive and nonradioactive materials
released to and in the environs, must comply with the technical basis specified in Sections
20.1301 and 20.1302 of 10 CFR Part 20. For both effluent and environmental monitoring,
the frequency of sampling and analysis, the types and sensitivity of analysis, action levels
and corrective action requirements, and the minimum number and criteria for locating
effluent and environmental monitoring stations also must be determined. A survey program
is essential to monitor the adequacy of containment and effluent control.
From release rates of airborne radioactivity, meteorological data, and locations of
release points (e.g. stack, roof vent), total annual body and significant organ doses can be
estimated for (1) individuals exposed at the point of maximum ground-level concentrations
off site, (2) individuals exposed at the site boundary in the direction of the prevailing wind,
A-20
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(3) individuals exposed at the site boundary nearest to the sources of emission, and
(4) individuals exposed at the nearest residence in the direction of the prevailing wind. The
license applicant must also estimate deposition of radioactive materials on food crops and
pasture grass, and total annual body doses and significant annual doses received by other
organs via such potential pathways to the public. The licensee is required to demonstrate
compliance with the exposure limits specified in 10 CFR Part 20 and 40 CFR Part 190 and
also effluent concentrations set forth in Table 2 of Appendix B of 10 CFR Part 20.
Each licensee is required to submit a semiannual effluent monitoring report to the
appropriate NRC Regional Office, specifying the quantity of each of the principal
radionuclides released to unrestricted areas in gaseous (and in liquid) effluents during the
previous 6 months of operations. The licensee must also submit such other information that
' NRC may require to estimate maximum potential annual radiation doses to the public
resulting from effluent releases. If quantities of radioactive materials released during the
reporting period are significantly above the licensee's design objectives previously reviewed
as part of the licensing action, the report shall cover this specifically.
A.2.3.3 Part 50 (Type 104") Licenses: Changes. Tests, and Experiments (10 CFR 50.591
Once a license to operate has been issued, NRC allows changes in facility design, operational
procedures, and activities unless the proposed change involves a modification to the technical
specifications or an unreviewed safety question. The licensee is required to maintain records
and to report all changes in facility descriptions or procedures contained in the FSAR.
Records and Reports (10 CFR 50.71). Each licensee and each holder of a
construction permit is required to maintain records and make reports in accordance with the
conditions established in the license or permit, or by the rules, regulations, and orders of the
Commission.
Backfitting (10 CFR 50.109). The Commission may require backfitting of a facility if
it finds that such action is necessary to protect public health and safety or that it will provide
substantial, additional protection at a justifiable cost
i
A.2.3.4 Part 70 Licenses. The requirements for airborne emissions -monitoring are
essentially the same as for Part 40 licensees.
A-21
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A.2.4 Inspection Programs
The fuel cycle facility inspection program is described in Chapter 2600 (NRC90a).
The materials licenses inspection program is described in detail in NRC Manual Chapter
2800 (NRC90b).
Initial inspections of licensees are generally conducted within 6 months to 1 year after
material is received and operations under the license have begun.
In conjunction with the licensee's required semiannual effluent monitoring reports to
NRC, the inspections determine the degree to which each plant is complying with its license
and technical specifications. If problems are identified, follow-up inspections are scheduled
in order to ensure that deficiencies are corrected. If a facility has persistent problems in
particular areas, inspections are performed more frequently.
A.2.4. 1 Part 30 Licenses. The inspection frequency for the various procedures at these
facilities is:
•
•
•
Medical Institution Broad & Medical Institution Other -various, every 1 to
5 years, average 18 months
Medical Private Practice - various, 1 to 5 years
Well-Logging - every 3 years
Manufacturing and Distribution Licenses - various, every 1 to 5 years
Incineration Licenses - yearly - ~
A.2.4.2 Part 40 Licenses. Tnitfol inspection of licenses are generally conducted within
6 months to 1 year after material is received and operations under the license have begun.
The frequency of subsequent inspections is shown below:
• Mills - at least once every year
• Military Munitions Testing - every 3 years
• Uranium Hexafluoride Production - at least once every year
• Rare Earth Extraction and Processing - every 3 years
A.2.4.3 Part 50 (Type 104") Licenses. Inspections (10 CFR 50.70). Each licensee (and
holder of a construction permit) must permit NRC to inspect its records, premises, and
activities. The licensee is required to provide office space onsite for a full-time NRC
resident inspector.
A-22
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A.2.4.4 Part 70 Licenses. The inspection program described in Section A.2.4.2 for Part
40 licenses applies, except for the following frequencies:
• Uranium Fuel Fabrication - at least once every year
• Interim Spent Fuel Storage - at least once every year
A.2.5 Enforcement Programs
The objective of the NRC's enforcement programs is to protect the public health and
safety by ensuring that Ucensees comply with regulatory requirements. The NRC's
enforcement policy, contained in 10 CFR Part 2, Appendix C, calls for strong enforcement
measures to ensure full compliance and is designed to prohibit operations by any licensees
who fail to achieve adequate levels of protection.
NRC's enforcement action has several levels of severity. The level of severity used
in a given situation varies with the seriousness of the matter and the licensee's previous
compliance record. The levels include:
• Written Notices of Violation — used in all instances of noncoinpliance with
NRC's requirements.
• Civil penalties — considered for licensees who evidence significant or
repetitive instances of noncompliance, especially if a previous Notice of
Violation has not been effective in achieving the expected corrective action.
Civil penalties may also be imposed in the case of a particularly significant
first-of-a-Mnd violation.
• Orders to "cease and desist" operations, or for modification, suspension, or
revocation of licenses — used in situations where licensees have not responded
to civil penalties or where violations pose a significant threat to public health
and safety or the common defense and security.
A.3 CONTROLS APPLICABLE TO AIRBORNE EMISSIONS
.\
' \
Current regulations limiting routine radionuclide airborne emissions from NRC-
licensed facilities are forth in 10 CFR 20 and 40 CFR 190. Part 20 establishes "Standards
for Protection Against Radiation." The recent revisions to Part 20 establish a new limit of
100 mrem/yr for members of the public. The 100 mrem/yr limit covers doses from both
A-23
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gaseous and liquid effluents and considers exposures from all sources. Part 20 also imposes
the requirement that exposures be as low as reasonably achievable (AIARA). Licensees may
demonstrate compliance with this limit using the effluent concentrations set forth in Table 2
of Appendix B of 10 CFR Part 20. The values in Table 2 for air are based on 50 mrem/yr.
EPA's environmental radiation standards for fuel cycle facilities are set forth in 40
CFR Part 190. 40 CFR 190 requires, in part, that the radiation doses to real individuals
from all uranium fuel cycle sources, including all gaseous and liquid effluent pathways and
direct radiation, should not exceed 25 mrem/yr to the whole body or any organ, except the
thyroid. The dose limit to the thyroid is established at 75 mrem/yr.
A.4 REFERENCES
* '
NRC90 U.S. Nuclear Regulatory Commission, "U.S. Nuclear Regulatory Commission
Functional Organizational Charts," NUREG-0325, Revision 14, August 1990.
NRC90a U.S. Nuclear Regulatory Commission, "NRC Inspection Manual, Chapter
2600, Fuel Cycle Facility Operational Safety Inspection Program," March
1990. -
NRC90b U.S. Nuclear Regulatory Commission, "NRC Inspection Manual, Chapter
2800, Materials Inspection Program," April 1990.
A-24
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APPENDIX B
NRC REGULATORY GUIDES
This appendix provides a partial list of the regulatory guides published by
NRC that are relevant to airborne effluents from nonreactor NRC-licensed
facilities.
B-l
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NO.
TITLE
REV. DATE
DIVISION 2 - RESEARCH AND TEST REACTORS
23.
Development of Technical Specifications for Experiments in Research Reactors
-
11/73
DIVISION 3 - FUELS AND MATERIALS FACILITIES
3.2
33
3.5
3.6
3.7
3.S
3.12
3.25
3.26
332
333
334
335
339
3.42
3.44
3.46
3.48
3.49
3.51
3.52
3.55
3.56
Efficiency Testing of Air-Cleaning Systems Containing Devices for Removal of Particles
Quality Assurance Program Requirements for Fuel Reprocessing Plants and for Plutonium Processing
and Fuel Fabrication Plants
Standard Format and Content of License Applications for Uranium 'Mills
Content of Technical Specifications for Fuel Reprocessing Plants
Monitoring of Combustible Gases and Vapors in Plutonium Processing and Fuel Fabrication Plants
Preparation of Environmental Reports for Uranium Mills
General Design Guide for Ventilation Systems of Plutonium Processing and Fuel Fabrication Plants
Standard Format and Content of Safety Analysis Reports for Uranium Enrichment Plants
Standard Format and Content of Safety Analysis Reports for Fuel Reprocessing Plants '
General Design Guide for Ventilation Systems for Fuel Reprocessing Plants
Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear
Criticality in a Fuel Reprocessing Plants
Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear
Criticality in a Uranium Fuel Fabrication Plant
Assumptions Used for Evaluating the Potential Radiological Consequences of Accidental Nuclear
Criticality in a Plutonium Processing and Fuel Fabrication Plants
Standard Format and Content of License Applications for Plutonium Processing and Fuel Fabrication
Plants
Emergency Planning for Fuel Cycle Facilities and Plants Licensed Under 10 CFR Parts 50 and 70
Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage
Installation (Water-Basia Type)
Standard Format and Content of License Applications, Including Environmental Reports, for In Situ
Uranium Solution Mining
Standard Format and Content for the Safety Analysis Report for an Independent Spent Fuel Storage
Installation or Monitored Retrievable Storage Installation (Dry Storage) '
Design of an Independent Spent Fuel Storage Installations (Water-Basin Type)
Calculational Models for Estimating Radiation Doses to Man from Airborne Radioactive Materials
Resulting from Uranium Milling Operations
Standard Format and Content for the Health and Safety Sections of License Renewal Applications for
Uranium Processing and Fuel Fabrication
Standard Format and Content for the Health and Safety Sections of License Renewal Applications for
Uranium Hexaflouride Production
General Guidance for Designing, Testing, and Maintaining Emission Control Devices at Uranium
Mills
-
1
1
-
-
1
2
-
-
-
'
-
1
1
-
1
1
'2
-
1
-
1
-
-
01/73
01/73
03/74
02/73
11/77
04/73
03/73
04/73
09/78
10/82
08/73
12/74
02/75
04/77
09/75
04/77
07/79
05/77
07/79
01/76
08/77
09/79
12/78
11/80
01/89
06/82
08/89
12/81
03/82
07/82
11/86
04/85
05/86
B-2
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3.59
3.60
3.61
3.62
3.63
3.64
3.65
Methods for Estimating Radioactive and Toxic Airborne Source Terms for Uranium Milling
Operations
Design of an Independent Spent Fuel Storage Installation (Dry Storage) •
Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry Storage Cask
Standard Format and Content for the Safety Analysis Report for Onsite Storage of Spent Fuel Storage
Casks
Onsite Meteorological Measurement Program for Uranium Recovery Facilities - Data Acquisition and
Reporting
Calculation of Radon Flux Attenuation by Earthen Uranium Mill Tailings Covers
[
Standard Format and Content of Decommissioning Plans for Licenses Under 10 CFR Parts 30, 40, and
70 -.-..'
- '
„
_
-
-
-
-
03/87
03/87
02/89
02/89
03/88
06/89
08/89,
DIVISION 4 r- ENVIRONMENTAL AND SITING
4.1
4.5
4.6
4.9
4.13
4.14
4.15
4.16
4.17
4.18
Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants
Measurements of Radionuclides in the Environment - Sampling and Analysis of Plutonium in Soil ,
Measurements of Radionuclides in the Environment - Strontium-89 and Strontium-90 Analyses
Preparation of Environmental Reports for Commercial Uranium Enrichment Facilities
Performance, Testing, and Procedural Specifications for Thermoluminescence Dosimetry:
Environmental Applications
Radiological Effluent and Monitoring at Uranium Mills
•Quality Assurance for Radiological Monitoring Programs (Normal Operations) - Effluent Streams and
the Environment
Monitoring and Reporting Radioactivity in Releases of Radioactive Materials in Liquid and Gaseous
Effluents from Nuclear Fuel Reprocessing and Fabrication Plants and Uranium Hexaflouride
Production Plants
Standard Format and Content Guide of Site Characterization Plans, for High-Level-Waste Geologic
Repositories ,
Standard Format -and Content of Environmental Reports for Near-Surface Disposal of Radioactive
Waste
1
_
_
1
1
1
1
1
1
. - .
01/73
04/75
05/74
05/74
12/74
10/75
11/76
07/77
06/77
04/80
12/77
02/79
03/78
12/85
07/82
03/87
06/83
DIVISION 5 -MATERIALS AND PLANT PROTECTION
5.4
5.5
5.13
5.18
5.24
5.33
5.42
5.45
Standard Analytical Methods for the Measurement of Uranium Tetraflouride (UF«) and Uranium
Hexaflouride (UFs)
Standard Methods for Chemical, Mass Spectrometric, and Spectrochemical Analysis of Nuclear-Grade
Uranium Dioxide Powders and Pellets
Conduct of Nuclear Material Physical Inventories
Limit of Error Concepts and Principles of Calculation in Nuclear Materials Control
Analysis and Use of Process Data for the Protection of Special Nuclear Material
Statistical Evaluation of Material Unaccounted For
Design Considerations for Minimizing Residual Holdup of Special Nuclear Material in Equipment for
Dry Process Operations - .
Standard Format and Content for the Special Nuclear Material Control and Accounting Section of a
Special Nuclear Material License Application (Including That for a Uranium Enrichment Facility)
-
-
. •
_
_
„
-
-
02/73
02/73
11/73
01/74
06/74
06/74
01/75
12/74
B-3
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5.51
5.58
5.62
Management Review of Nuclear Material Control and Accounting Systems
Considerations for Establishing Traceability of Special Nuclear Material Accounting Measurements
Reporting of Safeguards Events -
-
1
1 '
06/75
11/78
02/80
02/81
11/87
DIVISION 8 - OCCUPATIONAL HEALTH
8.2
8.10
8.18
8.21
8.23
8.24
8.25
8.30
8.31
Guide for Administrative Practices in Radiation Monitoring
Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Reasonably
Achievable
Information Relevant to Ensuring that Occupational Radiation Exposures at Medical Institutions 'Will
Be As Low As Reasonably Achievable i
Health Physics Surveys for Byproduct Material at NRC-Licensed Processing and Manufacturing Plants
Radiation Safety Surveys at Medical Institutions
Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication
Calibration and Error Limits of Air Sampling Instruments for Total Volume of Air Sampled
Health Physics Surveys in Uranium Mills
Information Relevant to Ensuring that Occupational Radiation Exposures at Uranium Mills Will Be As
Low As Reasonably Achievable
-
1
1-R
1
1
1
1
-
•
-
02/73 ,
04/74
09/75
05/77
12/77
10/82
05/78
10/79
02/79
01/81
11/78
10/79
08/80
06/83
05/83
DIVISION 10 - GENERAL
10.1
10.2
103
10.4
10.5
10.7
10.8
10.10
Compilation of Reporting Requirements for Persons Subject to NRC Regulations
Guidance to Academic Institutions Applying for Special Nuclear Material Licenses of Limited Scope
Guide for the Preparation of Applications for Special Nuclear Material Licenses of Less Than Critical
Mass Quantities
Guide for the Preparation of Applications for Licensees to Process Source Material
Applications for Type A Licenses of Broad Scope
Guide for the Preparation of Applications for Licenses for Laboratory and Industrial Use of Small
Quantities of Byproduct Material
Guide for the Preparation of Applications for Medical Use Programs
Guide for the Preparation of Applications for Radiation Safety Evaluations and Registration of Devices
Containing Byproduct Material
1
2
3
4
1
1
1
2
1
1
1
2
-
01/75
07/75
08/75
05/77
10/81
03/76 •
12/76
07/76
04/77
07/76
03/77
12/87
09/76
12/80
02/77
08/79
01/79
10/80
08/87
03/87
B-4
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APPENDIX C
DESCRIPTION OF NRC AND AGREEMENT STATE
LICENSED ACTTVniES
This appendix describes the activities for which an NRG or Agreement State
license is required (NRC91).
C-l
-------
Contents
> " • - .> • •
C.I General C-3
C.2 Byproduct Material Program (10 CFR 30) C-5
C.3 Source Material Program (10 CFR 40) . . . C-8
C.4 Research and Test Reactor Program (10 CFR 50, Type 104) C-10
C.5 Special Nuclear Material Program (10 CFR 70) C-10
C.6 References . '. C-14
C-2
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APPENDIX C
DESCRIPTION OF NRC AND AGREEMENT STATE
.-..'. LICENSED ACTIVITIES
C.I GENERAL
NRC assigns a five-digit program code number to each license to designate the major
activity or principal use provided for in the license.1 The regulations applicable to the
various activities and uses of byproduct, source, and special nuclear materials are contained
in Parts 30, 40 and 70, respectively, of Title 10 of the Code of Federal Regulations (CFR).
A basic understanding of these regulations is a necessary prerequisite to the proper
assignment of a program code to a particular activity or use. NRC uses about 100 program
codes to classify the approximately 8,200 active licenses under its direct control. Some of
these program codes narrowly define an activity, such as radiography, while other program
codes have a wider scope. More than one code may apply to a given license. However, the
primary code indicates the licensee's principal use of material. A secondary code may be
used to indicate other significant uses.
"Broad" licenses are issued to large facilities having a more comprehensive
radiological protection program. These licenses authorize possession of a wide variety of
radioactive materials without having each radionuclide and authorization listed on the license.
There are three types of broad licenses—Type A, Type B, arid Type C. Most broad licenses
are Type A. (For a clear understanding of these three types, see 10 CFR Part 33.)
Broad Type A licenses are issued pursuant to 10 CFR 33.13 and typically authorize
possession of any byproduct material with an atomic number between 1 and 83, in
any chemical or physical form. The maximum possession limit is usually specified
both for the individual radionuclide and for the total activity of all radionuclides.
These licensees must have a radiological safety officer and a committee that acts in
the place of NRC to make day-to-day decisions about the program.
i
Broad Type B licenses are issued pursuant to 10 CFR 33.14 and authorize possession
of a variety of radionuclides. The maximum possession limit is specified in 10 CFR
33.100, Schedule A, Column I. Broad Type B licensees must have a radiological
safety officer and adequate administrative controls
1 The program codes referred to are designated by NRC and may or may not be used by Agreement States.
C-3
-------
Broad Type C licenses are issued pursuant to 10 CFR 33.15 and authorize possession
of a variety of radionuclides. The maximum possession limit is specified in 10 CFR
33.100, Schedule A, Column H. Broad Type C licensees must have training and
experience as specified in the regulations, and the licensee must have adequate
administrative controls.
' * ' '
"Other" licenses are usually issued to smaller organizations requiring a more
restrictive license. These licenses are usually more specific in identifying each radionuclide,
the chemical and physical form, and the authorized activities and users.
The program codes are also Used to indicate the inspection category and priority and
fee categories. Materials licensing and inspection fee categories are described in 10 CFR
Part 170.31. The fuel cycle facility inspection program is described in NRC Manual Chapter
2600 (MC 2600)(NRC90). The inspection frequency for the various procedures at these
facilities is described in Table 1 of MC 2600. Inspection program categories and priorities
for materials licenses are described in detail in NRC Manual Chapter 2800 (MC 2800)
(NRC90a).
Initial inspection of licenses in categories with priorities 1 through 5 are conducted
within 6 months after material is received and operations under the license have begun.
Initial inspections of licenses in categories with priorities 6 and 7 are conducted within
1 year.
Routine, periodic inspections are normally conducted at intervals in years
corresponding to the inspection priority for that category:
• Priority 1 - yearly
• Priority 2 - every two years
• Priority 3 - every three years
• Priority 4 - every four years
• Priority 5 - every five years
• Priority 6 or 7 - inspected initially and thereafter normally only for resolution
of problems.
C-4
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C.2 BYPRODUCT MATERIAL PROGRAM (10 CFR 30)
Byproduct materials are man-made radioactive materials (except special nuclear
material - refer to Section C.5) produced or made radioactive by exposure to the radiation
incident to the process of producing or utilizing special nuclear materials such as in a nuclear
reactor. Byproduct material does include activation products from nuclear reactors and from
Plutonium-beryllium (Pu-Be) neutron sources, but does not include activation products from
other neutron sources such as Cf-252 or accelerators.
Byproduct Material Licenses (10 CFR 30. 32. 33. &
Byproduct Material Licenses are issued to educational institutions, medical facilities,
industrial facilities, and individuals for the possession and use of byproduct materials and
radionuclides for teaching, training, research and development, manufacturing, equipment
calibration, medical research and development, medical diagnosis and/or therapy. There are
many Byproduct Material Licenses categories, including Medical Private Practice Licenses,
Well-Logging Licenses, Measuring Systems Licenses, Waste Disposal Services Licenses,
General License Distribution licenses, Exempt Distribution Licenses, Industrial Radiography
Licenses, Irradiators Licenses, and Low Level Waste Storage Licenses, some of which do
not have air emission concerns. Listed below are those licenses that are required to comply
with regulatory limits on airborne radionuclide emission.
• Academic Broad and Academic Other - These licenses are issued to educational
institutions for the possession and use of radionuclides for teaching, training and
some research purposes, such as C-14 dating, equipment calibration, tracer studies,
and the identification of substances in compounds.
• Medical Institution Broad & Medical Institution Other - A medical institution is
defined in 10 CFR 35.2 to be an organization in which several medical disciplines are
practiced. It typically provides 24-hour-per-day medical, surgical, or psychiatric
treatment, nursing, food, and lodging to ill or injured patients. Medical Institution
Broad and Medical Institution Limited licenses are issued to organizations for the
application of byproduct material, or its radiation, to humansr Separate licenses are
issued to authorize teletherapy. Radioactive material administered to patients is an in-
vivo procedure.
C-5
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Medical Private Practice - These licenses are issued, pursuant to 10 CFR 35.12, to
physician for the possession and use of radionuclides in well established diagnostic
and therapeutic procedures usually in their offices outside a medical institution.
Well Logging - WeE-logging licenses are issued, pursuant to 10 CFR 39, to firms for
the possession and use of radionuclides for subsurface surveying to obtain geological
information. These testing procedures are primarily used in oil, gas, and mineral
exploration to identify subsurface geologic formations.
Measuring Systems - Measuring system licenses are issued for the possession and use
of measuring devices such as gauges and gas chromatographs containing
radionuclides. Frequently, the equipment is serviced and leak tested by the
manufacturer or lessor of the equipment.
Manufacturing and Distribution - These licenses are issued for the manufacture and
distribution of products containing byproduct material in various forms for a number
of diverse purposes. Licensees include medical suppliers that process, package and
distribute products such as diagnostic test kits, radioactive surgical implants, and
tagged radiochemicals for use in medical, academic and industrial research, and for
diagnosis and therapy. Licensees are also suppliers who, after purchasing bulk
quantities of byproduct material, process, encapsulate, package, and distribute these
sealed sources for use in gamma radiography, cobalt irradiation, and well-logging.
Firms are also involved with the manufacture, assembly, and distribution of various
other products that contain radionuclides. The broad licenses are issued to the larger
facilities having more comprehensive radiological protection programs.
Waste Disposal Services - Waste disposal licenses authorize the collection,
transportation, and storage of radioactive wastes. These licenses authorize firms to
collect packaged waste material, transport the waste, and temporarily store it before
transporting the waste to an authorized burial ground. Some licenses authorize the
opening of packages and treatment of the waste to reduce the volume, e.g.,
compaction.
C-6
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General License Distribution - General license distribution licenses are issued for the
distribution of byproduct material, usually sealed sources in devices, to general
licensees. Examples of such devices are: gauges, luminous aircraft safety devices,
calibration and reference sources, ice detection devices, and in vitro test kits. The
requirements for a license for distribution to general licensees are specified in various
sections of 10 CFR 32. A general licensee does not need to submit a formal
application and does not receive a formal license. The conditions of a general license
are described in 10 CFR 31.
Exempt Distribution - Exempt distribution licenses are issued for the commercial
distribution of byproduct material to persons who are exempt from the licensing
requirements. These exemptions and their limitations, if any, are defined in 10 CFR
30.14-30.20. Examples of exempt items are: watches, balances, locks, compasses,
electron tubes, synthetic plastic resin for sand consolidation, and smoke detectors.
The requirements for a license to distribute byproduct material to persons exempt
from licensing are presented in 10 CFR 32.
Industrial Radiography - Industrial radiography licenses are issued for the possession
and use of sealed radioactive materials, usually in exposure devices or "cameras," that
emit gamma rays for nondestructive examination of pipelines, weld joints, steel
structures, boilers, aircraft and ship parts, and other high-stress alloy parts. The
radioisotopes most commonly used are Co-60 and Lr-192. Radiography can be
conducted either in a permanent facility or at a temporary job site.
Irradiators - Irradiator licenses are issued for the possession and use of high-activity
sealed sources of radioactive material in an irradiator constructed so that the sealed
sources and the material being irradiated are contained in a shielded volume. Primary
uses include non-human medical and nonmedical research, conducted chiefly by
universities, and industrial uses, such as the sterilization of medical products and
drugs and treatment of hardwoods, plastics, and semi-conductor materials. The
radioisotopes most commonly used in these irradiators are Co-60 and Cs-137. Self-
shielded units are designed so that the operator cannot inadvertently place any part of
his/her body in the path of the beam. Units other than self-shielded units may rely on
facility alarms and interlocks to prevent accidental exposure to radiation. The
"Irradiators Other" category includes units where the source is stored and/or used
under water.
C-7
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• Research and Development Licenses - These licenses are issued to private
organizations, universities, and government agencies for the possession and use of
radionuclides in research. Typical uses include: irradiation of materials, tracers and
catalysts in chemical reactions, measurement using industrial gauges, and the
identification of substances in compounds. In private industry, uses are primarily in
product development. In academic institutions, research and development includes
training of students in the use of radioactive materials. Broad licenses are issued to
larger facilities having a more comprehensive radiation protection program where the
types of research being conducted may change rapidly. Typical activities include
environmental analysis, food quality studies, aerospace and engineering applications,
and product development.
• Civil Defense - Civil defense licenses are issued for the possession and use of sealed
sources for training individuals in civil defense activities, such as calibrating and
' demonstrating the use of radiation survey and monitoring equipment.
• Low-Level Waste Storage - Other - Licenses are issued to allow additional onsite
storage of low-level radioactive waste generated on site.
C.3 SOURCE MATERIAL PROGRAM (10 CFR 40)
f '
Source materials are materials essential to the production of special nuclear materials
(refer to Section C.5). Source material includes: (1) uranium (and depleted uranium
produced as enrichment tails) or thorium, or any combination thereof, in any physical or
chemical form, or (2) ores that contain by weight one twentieth of one percent (0.05%) or
more of uranium, thorium, or any combination thereof. Source material does not include
special nuclear material.
Source Material Licenses
Source Material Licenses are issued for the possession and use of refined uranium
and/or thorium for fabrication, research, and manufacture of consumer products such as
ceramics and glassware, manufacture of refractors, uranium shielding, analytical standards,
and other uses not specifically classified. A smaller number of these licenses are issued to
allow the possession of uranium and/or thorium for uses other than processing or fabrication
C-8
-------
of any kind, such as distribution and storage. An even smaller number of these licenses are
issued for the use of uranium in subcritical assemblies. The Source Material Licenses are
divided into the following categories:
• Mills - These licenses are issued for the extraction of uranium from uranium ore. In
mining operations, the ore is crushed, ground to fine mesh, and chemically treated to
. extract the uranium and convert it to a form called yellowcake.
• Source Material. Other. Less Than 150 Kilograms - These licenses are issued for the
possession and use of source material for fabrication, research, or manufacture of
consumer .products. These licenses do not allow the possession of more than 150
kilograms of material.
• Source Material. Shielding - These licenses are issued for the possession and use of
source material in shielding for protection against radiation.
• Source Material, Military Munitions Testing - These licenses are issued for the
possession, use and testing of depleted uranium products designed for the military.
• Source Material. General License Distribution - These licenses are issued to authorize
the initial transfer of industrial products and devices containing depleted uranium, or
to allow the initial transfer of such products or devices to persons issued a general
license under Part 40.25.
• Source Material. Other. Greater Than 150 Kilograms - These licenses are issued for
the possession and use of source material for fabrication, research, or manufacture of
consumer products. These licenses authorize the possession of more than 150
kilograms of material.
Hexafluoride Production Plants - These licenses are issued for the possession
and use of uranium 19 allow the conversion of yellowcake and/or ore concentrates to
uranium hexafluoride
Solution Mining - These licenses are issued for the extraction of uranium from
uranium ores. The only mining operation licensed by NRC is solution mining, which
is leaching of ore by injection of liquid chemicals into the geologic formation.
C-9
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• Heap Leach. Ore Buying Stations and Byproduct Recovery - These licenses are issued
for the recovery of source material from low-grade uranium ores, from old tailings
piles, or from a small ore body at a location distant from a mill complex. The heap
leach process consists of spraying or trickling an acid dilution over sections of the
heap pile. Pipes or covered drains in the base of the pile collect the uranium-enriched
solution after it percolates through the heap.
• Rare Earth Extraction and Processing - These licenses are issued for the possession
and use of source material for processing activities not directly related to the nuclear
fuel cycle. This category includes licenses for extraction of metals, heavy metals, and
rare earths.
• Source Material licenses - These licenses are issued for the possession and use of
source material for miscellaneous activities including licenses for sites that once
processed source material but are now being decommissioned.»Some sites include
disposal areas, such as tailings or slag piles. Licenses for these sites are issued for
possession and storage only.
C.4 RESEARCH AND TEST REACTOR PROGRAM (10 CFR 50, TYPE 104)
Research and test reactors include those used in medical therapy and research and
development facilities. The latter means (1) theoretical analysis, exploration, or
experimentation; or (2) the extension of investigative findings and theories of a scientific or
technical nature into practical application for experimental and demonstration purposes,
including the experimental production and testing of models, devices, equipment, materials,
and processes. .
C.5 SPECIAL NUCLEAR MATERIAL PROGRAM (10 CFR 70)
Special nuclear materials include plutonium, U-233, uranium enriched in the isotopes
of U-233 or U-235, and any material artificially enriched in any of these materials.
C-10
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Special Nuclear Material Licenses
Special Nuclear Material licenses are issued to licensees to receive, own, acquire,
deliver, possess, use, and initially transfer special nuclear material. These licenses are
divided into the following categories:
• Hot Cell Operations - These licenses are issued for the processing and fabrication of .
reactor fuels containing uranium and/or plutonium for experimental purposes. Some
facilities also perform chemical operations to recover the uranium and plutonium from
scrap'and other off-specifications materials.
• Decommissioning of Advanced Fuel R&D and Pilot Plants - These licenses are issued
to facilities which has notified NRC of their intent to terminate a portion or all of
their activities involving special nuclear material and/or have submitted to NRC a plan
and schedule for the facilities, property, and equipment so that they may be released
for unrestricted use.
•' Uranium Enrichment Plants - Uranium enrichment plant licenses are issued for the
possession and use of source and special nuclear material for the purpose of enriching
natural uranium in the U-235 isotope. Existing and planned plants enrich uranium in
the form of uranium hexafluoride, either by gaseous diffusion or gas centrifuge
methods. Future plants may use other forms of uranium and methods of enrichment.
Plants whose product is for eventual use in commercial power reactors enrich uranium
up to about 5 percent U-235, while plants whose product is for naval reactor
propulsion enrich uranium to greater than 90 percent U-235.
• Uranium Fuel Fabrication Plants - These licenses are issued for the possession and
use of special nuclear material for the purpose of fabricating uranium fuel elements.
In most uranium facilities where light water reactor fuels are processed, low-enriched
uranium hexafluoride is converted to uranium dioxide pellets and inserted into
zirconium tubes. The tubes are fabricated into fuel assemblies which are shipped to
commercial nuclear power plants. In other facilities, high-enriched uranium is
processed into naval reactor fuel and fabricated into naval reactor cores or core
components. Licenses are for possession and use of 5 kilograms or more of U-235
that has been enriched to less than 20 percent.
C-ll
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Decommissioning of Uranium Fuel Fabrication Plants - These licenses are issued to
facilities that have notified NRC of their intent to terminate a portion or all of their
activities involving special nuclear material and/or has submitted to NRC a plan and
schedule for the facilities, property, and equipment so that they may be released for
unrestricted use.
Uranium Fuel Research and Development and Pilot Plants - These licenses are issued
for the possession and use of enriched uranium for purposes such as academic training
and in research and development activities associated with nuclear fuel other than fuel
processing. Licenses authorize possession and use of 5 kilograms or more of
enriched U-235 in unsealed form, or 2 kilograms or more of U-233 in unsealed form.
j
Critical Mass Material - These licenses are issued for the possession and use of
special nuclear material in quantities sufficient to form a critical mass, specifically,
more than 350 grams of enriched U-235, more than 200 grams of U-233, more than
200 grams of plutonium, or any combination thereof.
Decommissinninp nf Critical Mass - Other Than Universities - These licenses are
issued to facilities that have notified NRC of their intent to terminate a portion or all
their its activities involving special nuclear material and/or has submitted to NRC a
plan and schedule for the facilities, property, and equipment so that they may be
released for unrestricted use.
Special Nuclear Material. Plutonium-Unsealed. Less Than a Critical Mass -. These
licenses are issued for the possession and use of small quantities of plutonium (less
than 200 grams total) in unsealed form for purposes such as biological and chemical
testing and for calibration of instruments.
Special Nuclear Material. U-235 and/or U-233 Unsealed. Less Than a Critical Mass -
These licenses are issued for the possession and use of small quantities of uranium
(less than 350 grams,of U-235 and/or less than 200 grams of U-233) in unsealed form
for purposes such as biological and chemical testing and for calibration of
instalments. "
C-12
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Special Nuclear Material. Plutonium Neutron Sources. Less Than 200 Grams - These
licenses are issued for the possession, and use of small quantities of plutonium (less
than 200 grams total) usually combined with beryllium as the source of neutrons for
instrument calibration, teaching and demonstration purposes, and industrial
applications.
Power Source with Byproduct and/or Special Nuclear Material- These licenses are •
issued for the possession and use of byproduct and/or special nuclear material to
generate heat or power that will be used in remote weather stations, space satellites,
and other special applications.
Special Nuclear Material Plutonium - Sealed Source in Devices - These licenses are
issued for the possession and use of sealed sources containing special nuclear material
installed in devices such as gauges.
'
Special Nuclear Material Plutonium - Sealed Source Less Than a Critical Mass -
These licenses are issued for the possession and use of small quantities of plutonium
(less than 200 grams total) in sealed sources for purposes such as biological and
chemical testing and for calibration of instruments, etc.
Special Nuclear Material. U-235 and/or U-233 - Sealed Source Less Than a Critical
Mass - These licenses are issued for the possession and use of small quantities of
uranium (less than 350 grams of U-235 and/or less than 200 grams of U-233) in
sealed sources for purposes such as biological and chemical testing and for calibration
of instruments, etc.
^ ' •'
Pacemaker - Byproduct Material and/or Special Nuclear Material - These licenses are
issued to: (1) medical facilities for the surgical implantation of pacemakers that are
powered by a device containing byproduct or special nuclear material;
(2) manufacturers and distributors for the distribution of these pacemakers; and
(3) individuals, most often Canadian citizens on holiday, with implanted nuclear
pacemakers who are visiting in the United States.
C-13
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• Special Nuclear Material. General License Distribution - These licenses are issued to
individuals for the initial distribution of calibration or reference sources containing
plutonium to persons who have been issued a general license under Part 70.19.
General licenses under Part 70.19 authorize the possession and use of plutonium in
calibration or reference sources. A person may be a general licensee only if the
person is already a specific licensee.
• Fresh Fuel Storage at Reactor Sites - These licenses are issued to commercial nuclear
power reactors that have been granted a Construction Permit (CP) but not an
Operating License (OL). These licenses authorize the storage of new unirradiated
reactor fuel elements containing special nuclear material. Once a reactor has been
granted an OL, this Part 70 materials license is terminated. (The OL includes
authorization for the possession of the fuel.)
• Interim Spent Fuel Storage - These licenses are issued under 10 CFR Part 72 for
possession of power reactor spent fuel and other radioactive materials associated with
spent fuel storage, in an independent spent fuel storage installation. (These licenses
are issued for up to 20 years.)
• Transport - Private Carriage - Transport-Private Carriage licenses are issued for the
possession of byproduct, source, and special nuclear materials in packages authorized
under Part 71, and in private carriage from a carrier's terminal to the licensee's
facility, all within the United States.
C.6 REFERENCES
NRC90 U.S. Nuclear Regulatory Commission, "NRC Inspection Manual, Chapter
2600, Fuel Cycle Facility Operational Safety Inspection Program," March
1990.
NRC90a U.S. Nuclear Regulatory Commission, "NRC Inspection Manual, Chapter
2800, Materials Inspection Program," April 1990.
NRC91 U.S. Nuclear Regulatory Commission, "Program Code Descriptions Used In
NRC Licensing and Inspection Programs," January 1991.
C-14
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APPENDIX D
DESCRIPTION OF FACILITIES EVALUATED
This appendix describes the types of facilities other than nuclear power
reactors, licensed by NRC and Agreement States, whose radioactive effluents
were evaluated for the purpose of conducting the NESHAPs rulemaking.
D-l
-------
Contents
D.I Byproduct Material Licensees (10 CFR 30) D-3
D.I.I Users and Producers of Radionuclides for Medical Purposes . . . D-3
D.1.2 Sealed Source Manufacturers , D-ll
D.I.3 Waste Receivers-Shippers and Disposal Facilities . D-12
D.2 Non-Power Reactor Licensees (10 CFR 50, Type 104) . D-13
D.2.1 Test and Research Reactors D-13
D.3 Uranium Fuel Cycle Facilities (10 CFR 40 and 70) . . D-14
D.3.1 Source Material Licensees (10 CFR 40) . D-14
D.3.2 Special Nuclear Material Licensees (10 CFR 70) D-18
Table D-l. NRC licensees other than power reactors . ,-. .... D-4
D-2
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APPENDIX D
DESCRIPTION OF FACILITIES EVALUATED
The NESHAP applies to approximately 8,000 NRC-licensed and non-DOE federal
facilities other than nuclear power reactors that possess unsealed sources of radioactive
materials. NRC-licensed facilities other than nuclear power reactors include material
licensees, non-power reactor licensees, and facilities engaged in the uranium fuel cycle.
NRC-licensed facilities other than nuclear power reactors also include facilities licensed by
the Agreement States but exclude low-energy accelerators and facilities regulated under 40
CFR Part 191, Subpart B. Pertinent information regarding the facility types considered for
evaluation, including those where further study was not warranted, is listed in Table D-l.
The major types of facilities covered by the standard are described in the following
sections. The discussion focuses on the physical forms of the radionuclides used and the
handling and processing that the materials undergo. These factors are major determinants of
the quantities of materials handled that become airborne.
The descriptions provided below were obtained from the Nuclear Regulatory
Commission's public document room(s), supplemented as necessary by "EPA's
Environmental Impact Statement, NESHAPs for Radionuclides, Background Information
Document - Volume 2," dated September 1989, and "Background Information Document -
, Procedures Approved for Demonstrating Compliance with 40 CFR Part 61, Subpart I," dated
October 1989.
D.I BYPRODUCT MATERIAL LICENSEES (10 CFR 30)
D.I.I Users and Producers of Radionuclides for Medical Purposes
The users and producers of radioactive materials for medical purposes constitute by
far the largest category of facilities handling unsealed radioactive sources. Approximately
two-thirds of the 8,000 facilities covered by the NESHAP are engaged in some aspect of the
production and distribution of radiopharmaceuticals or in the medical application of these
materials. Medical uses of radiopharmaceuticals include biomedical research and patient
administration of radiopharmaceuticals for both diagnostic and therapeutic purposes.
D-3
-------
Table D-l. NRC licensees other than power reactors.
PROGRAM
CODE
01100
OHIO
01120
01200
02110
02120
02121
02200
-02201
02209
02220
02400
02410
02500
02511
02512
02513
03211
03212
03213
03214
03218
03232
03234
03610
03611
03612
03613
03620
11200
11210
11230
11300
11500
11600
11800
21130
21135
21215
21240
A. ^<>I4C£NSEES COVERED BY THE RANDOM SURVEY1
•, •:•••• '& < •• v
, (,,, ,„ ^ ' ~
Academic Type A
Academic Type B
Academic Type C '
Academic Other
Medical Institution Broad
Medical Institution limited
Medical Institution Custom
Medical Private Practice Limited
Medical Private Practice Custom
Grandfathered In-Vivo General Medical Use
Mobile Nuclear Medicine Service
Vet, Non-Human
In-Vitro Tesflab
Nuclear Pharmacies
Medical Product Distribution - 32.72
Medical Product Distribution - 32.73
Medical Product Distribution - 32.74
Manufacturing/Distribution Broad Type A
Manufacturing/Distribution Broad Type B
Manufacturing/Distribution Broad Type C
Manufacturing/Distribution Other
Nuclear Laundry
Waste Disposal Service Prepackaged Only
Waste Disposal Service Processing/Repackaging
Research and Development Broad Type A
Research and Development Broad Type B
Research and Development Broad Type C
Research and Development Broad
-Multisite-Multiregional
Research and Development Other
Source Material Other < 150kg
Source Material Shielding
Source Material General License Distribution
Source Material Other > 150 k
Solution Mining (R&D and Commercial Facilities)
Heap Leach, Ore Buying Stations & Byproduct Recovery -
Source Material
Hot Cell Operations
Decommissioning Uranium Fuel R&D & Pilot Plants
Decommissioning Uranium Fuel Processing Plants
Uranium Fuel R&D and Pilot Plants
JSIHMBERQt?'
•ACTIVE , '
UCEKfSEES
,44 •
14
19
0
121
1384
14
306
165
69
22
4
124
50
3
7
6
18
17
3
134 •
5
7
7
130
13
21
3
561
26 •
44
0
84
9
3
4
5
2
3
1
D-4
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Table D-l. NRC licensees other than power reactors (continued).
, PROGRAM ^
CODE
21310
21320
21325
22110
22111
22170
25110
A> 1SQRC-IJCE1SSEES CQ\^^
PROGRAM CODE DESCRIPTION
Critical Mass Material for Universities
Critical Mass Material Except Universities
Decommissioning Critical Mass Except Universities
Special Nuclear Material, Unsealed Plutonium < 200g
Special Nuclear Material, Unsealed TJ-235 < 350g,
U-233 < 200g
Special Nuclear, General License Distribution
Transport - Private Carriage •
Miscellaneous
SUBTOTAL
, NUMBER OF
ACTIVE
10
4
0
.16
12
0
2
13
3,509
1 10 CER 50 Kcensees (reactors and test/research reactors) were not part of the data base used to select
the random sample. Other source categories were deleted from the data base whenever (1) EPA had a
specific interest in studying that source category (e.g., rare earth processors), or (2) due to their .small
numbers, there was no guarantee that the source category would show up in the random selection (e.g.,
low-level radioactive waste disposal facilities). -
% B. mC-LICENSEES JNCLT3DED W THE DESIGNATED SUMVBY - PMOK EVALUATIONS
UPDATED AND NEW SOURCE CATEGORIES ADDED2
PROGRAM
CODB
03231
03233
03235
06100
11100
11220
11400
11700
21210
PROGRAM CODE DESCRIPTION
Test and Research Reactors
Waste Disposal - Burial
Waste Disposal Service - Incineration
Incineration, Non-Commercial
Low Level Waste Storage
Mills,
Source Material Military Munitions Testing
Uranium Hexaflouride Production Plants
Rare Earth Extraction and Processing
Uranium Fuel Processing Plants
SUBTOTAL
A NUMBER OF
ACTIVE
! ' LICENSEES
70
2
1
0
0
20
9
2
11
11
126
2 The designated facility data base consists of (1) specific facilities (e.g., large hospitals) EPA has
evaluated in prior studies and in need of updating (e.g., site-specific demographics), and (2) specific
facilities in which EPA has developed an interest (e.g. , rare earth processors).
D-5
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Table D-l. NRC licensees other than power reactors (continued).
C. NRC-ZICENSEES EXCO3DED £ROM THE RANDOM SORVBY 'AND DESIGNATED St^RYEY :
SEALED SOURCES3
PROGRAM
CODE
02210
02300
03110
03111
03112
03113
03120
03121
03122
03123
03124
03220
03221
03222
03223
03224
03225
03240
03241
03242
03243
03244
03250
03251
03252
03253
03254
03255
03310
03320
03510
03511
03520
03521
03710
22120
PROGRAM CODE DESCRIPTION
Eye Applicators Sr-90
Teletherapy
Well-Logging Byproduct and/or SNM Tracer and Sealed
Sources
Well-Logging Byproduct and/or SNM Sealed Sources
Well Logging Byproduct Only - Tracers Only
Field Flooding Studies
Measuring Systems Fixed Gauges
Measuring Systems Portable Gauges
Measuring Systems Analytical Instruments
Measuring Systems Gas Chromatographs
Measuring Systems Other
Leak Test Service Only ,
Instrument Calibration Service Only, Source < 100 Curies
Instrument Calibration Service Only, Source > 100 Curies
Leak Test & Instrument Calibration Service, Source < 100 Curies
Leak Test & Instrument Calibration Service, Source > 100 Curies
Other Services
General License Distribution - 32.51
General License Distribution - 32.53
General License Distribution - 32.57
General License Distribution - 32.61
General License Distribution - 32,71
Exempt Distribution - Exempt Concentrations and Items
Exempt Distribution - Certain Items
Exempt Distribution - Resins
Exempt Distribution - Small Quantities
.Exempt Distribution - Self Luminous Products
Exempt Distribution. - Smoke Detectors
Industrial Radiography Fixed Location
Industrial Radiography Temporary Job Sites
Irradiators Self Shielded < 10,000 Curies
Irradiators Other < 10,000 Curies
Irradiators Self Shielded > 10,000 Curies
Irradiators Other > 10,000 Curies
Civil Defense
Special Nuclear Material, Plutonium Neutron Sources, <200g
'NUMBER OF
i AC1OT N '
LICENSEES
50
223
45
54
6
4
,772
1,529
112
596
83
10
48
19
18
5
70
49
1
1
0
33
6
63
0
45
13
26
64
192 '
172
19
33
20
30
92
D-6
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Table D-l. NRC licensees other than power reactors (continued).
C, mC-HCENSEBS EXCLPDBD UtOM 1KB RANDOM SORVBY AMJ> DESIGNATED SORVBY -
V " - SEALED SOURCES3
PROGRAM
CODE
22130
22140
22150
22151
22160
22161
22162
23100
PROGRAM CODE DESCRIPTION
Power Sources with Byproduct and/or Special Nuclear Material
Special Nuclear, Plutonium Sealed Source in Devices
Special Nuclear, Plutonium Sealed Sources, < Critical Mass
Special Nuclear, U-235, U-233 Sealed Sources, < Critical Mass
Pacemaker Byproduct/Special Nuclear - Medical Institution
Pacemaker Byproduct/Special Nuclear - Individual
Pacemaker Byproduct/Special Nuclear - Manufacturing &
Distribution
Fresh Fuel Storage at Reactor Sites
SUBTOTAL
TOTAL
NUMBER OF
ACTIVE
LICENSEES
0
10
15
3
68
4
1
5
4,609
8,244
3 Sealed source users are excluded from the Random Survey and Designated Survey data bases because the
potential for airborne radioactive effluents is essentially zero.
Radiopharmaceutical Users
The types of facilities that use radionuclides for medical purposes include hospitals,
clinics, and biomedical research facilities. The radionuclides used directly in patient therapy
and diagnosis are termed "radiopharmaceuticals," while those used in research are referred to
as "radionuclides." For simplicity, the term "radiopharmaceuticals" will be used to refer to
the radioactive materials used in both patient administration and research.
The radiopharmaceuticals used at medical facilities occur in all three basic physical
states: solid, liquid, and gas. The physical state of a particular radiopharmaceutical product
is determined by (1) the chemical form of the radionuclide and (2) the solution or other
mixture, if any, in which the radionuclide is dispensed. Both the radionuclide and the
substance in which it is mixed are chosen to suit specific therapeutic, diagnostic, and
research purposes.
D-7
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The mixing of the radionuclide with some other substance means that the physical
state of a radiophannaceutical product may be different than the physical state of the
radionuclide itself. In this document, discussions of the form of a particular radionuclide
refer to the radionuclide product. The physical states of these products are important in
assessing the potential for airborne release.
Most radionuclides used in medical facilities occur in liquid form. These liquids may
be administered either orally or intravenously. Orally administered radionuclides are usually
in the form of aqueous solutions. Many of these chemicals are ionic salts and thus occur in
liquid form as saline solutions. Radionuclides that are administered intravenously may occur
as solutions, colloids, or suspensions.
Solutions consist of molecules of solids or gaseous substances dissolved in a liquid.
Colloids involve the dispersion of larger particles (on the order of 10 nanometers to
1 micrometer in diameter) in a liquid medium; the larger particles are prevented from
aggregating and settling by being coated with a layer of gelatin (as is done with Au-198).
Suspensions are similar to colloids but involve the radionuclide labeling of still larger
particles (greater than 10 micrometers in diameter) of substances such as human serum
albumin. .
Gaseous radionuclides usually occur naturally in elemental form (e.g., Xe-133), and
are administered to patients as a pure gas or as a gas diluted by air. Patients normally inhale
the gas from a bag or from a gas "generator" through a respirator.
Solid radionuclides occur as gelatin capsules containing liquid solutions of the
radionuclide chemical. In some cases, the solution is absorbed in dry filler material. Solid
radionuclides are administered orally to patients.
The number of radionuclides with medical applications is extensive and increasing. In
the areas of diagnosis and therapy, the most commonly used radiopharmaceuticals include
Cr-51; Co-57, -58, and -60; Ga-67 and -68; Tc-99m; 1-123, -125, and -131; Se-75, Xe-127
and -133; and Tl-201. Biomedical researchers employ tritium, C-14, P-32, and S-35
extensively. The radiopharmaceuticals used in medical applications may be obtained from
radiopharmaceutical manufacturers or independent radiopharmacies, or they may be produced
on site from radiopharmaceutical generators. Because of the relatively short half-lives of the
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radionuclides used in medicine, shipments from vendors are received frequently (weekly or
daily), and[storage times are minimal.
Radiopharmaceuticals purchased from vendors may be in the form of pre-packaged
dose kits, radiopharmaceutical generators, or bulk supplies from which individual doses are
extracted and prepared. Handling of prepackaged dose kits may involve no more than
removing the material from the package and administering the radiopharmaceutical to the
patient either orally or by intravenous injection.
Handling of materials obtained in the form of bulk stocks or radiopharmaceutical
generators is more involved. In general, these materials are received and stored in a central
area where individual doses are prepared. In the case of liquids, dose preparation involves
extracting the required quantity from the stock solution by syringe or pipette and diluting the
material in a suitable sterile medium. These operations are conducted in a fume hood, and
the dose is administered to the patient either intravenously or orally.
Preparation of doses from radiopharmaceutical generators, of which Mo-99/Tc-99m
generators are the most common, involves elution of the product from the generator and
division of the elute into individual doses. The procedures for eluting a generator^depend on
whether it is a wet or dry column design. In a wet column generator, an evacuated
extraction vial is attached to the end of the generator column with a sterile needle. Using the
vacuum within the vial, the solvent is pulled from the generator reservoir through the column
and into the vial. The procedure for a dry column generator is similar. However, since dry
generators do not have a reservoir of solvent, solvent must be added to the column prior to
elution. The charge vial is attached to one end of the generator, and then the evacuated
extraction vial is attached to the other end. The solution is drawn through the generator
column and collected in the elution vial. These elution procedures and dose divisions are
conducted in a fume hood, with the generator shielded to prevent external irradiation of the
technicians.
t
Handling of radionuclides for biomedical research is more varied than that of
radiopharmaceuticals used for patient administration. Depending on the specific
radionuclides used and the goal of the experiment, the materials majr simply be extracted
from bulk stocks and administered, or the radionuclide may be subjected to additional
chemical or physical processing. .
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Radiopharmaceutical Producers and Suppliers
Radiopharmaceutical manufacturers produce the radionuclide-labeled compounds,
diagnostic kits, and radionuclide generators used in biomedical research and medical
diagnosis and therapy. The radiopharmaceutical products may be shipped directly to medical
users, or they may be shipped to independent radiopharmacies where individual doses are
prepared from the bulk supplies or generators and distributed to medical users. Individual
radiopharmaceutical manufacturers may specialize in only a few widely used ,
radiopharmaceuticals or may produce many of the radionuclides used in biomedical research
and patient diagnosis and therapy.
The radionuclides used in radiopharmaceuticals are produced either in nuclear reactors
or in accelerators. Radiopharmaceutical manufacturers may operate their own production
facilities or may purchase the bulk radionuclides from an outside vendor. In producing the
bulk radionuclides, a suitable target is first prepared and then bombarded with neutrons or
positive ions in the reactor core or accelerator. Once irradiation is complete, the target is
removed from the production device, and the product is recovered and purified in a hot cell
by appropriate chemical processing.
The production of the labeled compounds used in radiopharmaceuticals and
biomedical research is essentially a wet chemistry process. Depending on the specific
radiopharmaceutical, workers conduct these operations within laboratory fume hoods or
gloveboxes. The final products are generally assembled and packaged in assembly line
operations.
Radiopharmaceutical generators are designed and produced as closed aseptic systems
using some type of chromatographic column. Typically, this chromatograpbic column
consists of an inorganic ion exchange resin to which the generator (parent) radionuclide is
bound. As the parent radionuclide decays, the decay product, which has different
chemical/physical properties, is produced. The decay product is eluted from the column by
the user at specified intervals. Generators are manufactured in a hot cell, where the parent
radionuclide is packed in the column, and the column of the generator is surrounded by
absorbent materials and shielding. The absorbent materials minimize the consequences of
accidental breakage; the shielding reduces the radiation exposure of users. Once the
generator is loaded, final assembly and packaging are carried out on an assembly line.
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Independent radiopharmacies are a relatively recent phenomenon. Generally located
in large cities, these facilities serve as distribution facilities. Radiopharmacies purchase bulk
stocks and generators from radiophannaceutical manufacturers and provide hospitals and
clinics with individually prepared doses on an as-needed basis. The dose preparation
procedures at these facilities do not differ from those at medical facilities that obtain their
radiopharmaceuticals directly from the manufacturers.
D.1.2 Sealed Source Manufacturers
Manufacture of Self-Dluminating Devices
While facilities that use only sealed radiation sources are not covered by the
NESHAP, the industrial facilities that produce sealed sources are subject to the standard.
The facilities in this category fall into two broad classes: those that manufacture
encapsulated alpha-, beta-, or gamma-emitting radiation sources and those that manufacture
self-luminous devices. Only the latter is included as part of the Designated Survey.
Self-illuminating devices include watches, compasses, signs, and aircraft
instrumentation. Historically, Ra-226 was used in radio-luminescent products. However, the
well-documented hazards of working with radium and the advent of other materials with
inherently superior characteristics have largely eliminated the use of radium. Today, tritium
and, to a much lesser extent, Kr-85 and Pm-147 are used in the production of self-luminous
devices.
Two general types of self-illuminating devices are made: those in which the radio-
luminous material is incorporated into a paint which is used to coat the dial and/or instrument
hands; and those in which a radioactive gas (tritium or krypton) is contained in a phosphor-
coated glass ampule. Only the second type is included as part of the Designated Survey.
Manufacturers of self-illuminating devices obtain the bulk radionuclides in either
gaseous or (rarely) liquid form from a vendor. In the case of self-iUuminating sources, the
gaseous radionuclide (tritium or Kr-85) is transferred to a glass ampule and sealed. The
process is carried out in areas with high ventilation rates or in fume hoods to protect the
workers.
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D.1.3 Waste Receivers-Shippers and Disposal Facilities
Low-Level Radioactive Waste Processing and/or Packaging
The radioactive wastes generated by facilities that use radionuclides must be disposed
of in an approved manner. In general, wastes with high specific activities (such as uranium-
contaminated scrap at non-oxide fuel fabrication facilities) will be recycled and recovered.
However, virtually every user of unsealed radioactive materials will generate solid, low-level
radioactive wastes which require active disposal. Such wastes may be incinerated on site or
packaged and shipped off site to a licensed low-level waste disposal facility. This study
investigated incinerators and packing facilities.
Waste receivers and shippers (sometimes called "waste brokers") are primarily
collection and shipping agents for facilities generating low-level wastes. Most such
receiving-shipping facilities simply collect the wastes in shipping containers approved by the
Department of Transportation from a number of waste generating facilities, monitor the
packages for contamination, and hold the wastes at a warehouse until they arrange a shipment
to a licensed disposal site. The licenses of most such receiving and shipping facilities do not
allow the facility to repack or even open the waste packages. However, several such
facilities have been licensed to open, compact, and repackage waste materials before
shipment.
Incineration . '
Recently, a new low-level waste operation called incineration has been established.
Waste incinerators provide a volume reduction service by processing waste in the form of
paper, plastic, metal, liquid, or animal carcasses. Most of the radioactivity projected to be
burned is called dry active waste (DAW) from nuclear power plants. Much of the remainder
is industry and institutional DAW. The retaining most radionuclides is immobilized and
packaged for disposal. Some amount of radioactive material is discharged from the stack.
Burning waste can reduce volume by as much as 95 percent.
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Disposal
Until recently, low-level waste disposal in the United States was accomplished via
"shallow land burial," a method that does not rely on engineered barriers to isolate the
waste. Over the years several problems have developed and resulted in the closing of three
of the six operating facilities. New federal and state laws require for future facilities that
engineered barriers be used in addition to good siting practices. Some states have required a
design goal of zero release.
A low-level radioactive waste disposal facility has two distinct phases of operation,
pre-closure and post-closure. During the pre-closure phase, waste is received onsite, re-
packaged if necessary, and placed in its final resting place. In the post-closure phase,
monitoring of the facility is continued for a period of years into the future until institutional
controls can no longer be assumed to be available, usually 100 years.
D.2 NON-POWER REACTOR
D.2.1 Test and Research Reactors
(10 CFR 50, TYPE 104)
NRC licenses approximately 70 academic, research, and industrial facilities to operate
test and research reactors. Test and research reactors are used as teaching devices, to study
reactor designs, to conduct research on the effects of radiation on materials, and to produce
radioactive materials used by sealed source and radiopharmaceutical manufacturers.
The design of such reactors and their sizes vary widely. Approximately 15 research
reactors are used primarily as teaching devices and have very low power outputs (less than
15 watts). The nuclear cores of these reactors have their uranium fuel dispersed and fixed in
a plastic matrix. Given the design and use of these teaching reactors, airborne releases
cannot occur during normal operations.
Research and test reactors used for experimental and production purposes include both
light-water pool and heavy-water tank-type designs, ranging in power from 100 kilowatts to
10 megawatts. All of these facilities use highly .enriched uranium fuel, either in metal or
mixed carbide fuel elements.
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la these reactors, experiments and/or production activities are conducted by remotely
inserting the target containing the material to be irradiated into the experimental ports or
beam holes that penetrate the reactor core. The target material is subjected to the neutron
flux of the reactor core for an appropriate period of time and then withdrawn via shielded
transport devices (called "rabbit systems") to a hot cell. The irradiated material is examined
or the product is recovered in the hot cell. Product recovery may be as simple as dissolving
a soluble salt in water, or it may involve evaporation, precipitation, extraction, distillation,
and/or ion exchange.
Potential airborne releases from such facilities include the fission products in the core
of the reactor, activation products generated during the operation of the reactor, and releases
from the disassembly and recovery of target materials in the hot cell.
In general, the activation products, along with any gaseous fission products escaping
the coolant, are released directly to the atmosphere from the facility exhaust. Materials that
become airborne during processing in the hot cell will be vented through the hot cell's
exhaust system. The effluent from the hot cell is generally filtered through high efficiency
particulate air (HEPA) filters before release.
D.3 URANIUM FUEL CYCLE FACILITIES (10 CFR 40 and 70)
The uranium fuel cycle includes uranium mills, uranium hexafluoride conversion
facilities, uranium enrichment facilities, light-water reactor fuel fabricators, light-water
power reactors, and fuel reprocessing plants. With the exception of the uranium enrichment
facilities that are owned by the federal government and operated by contractors under the
supervision of the Department of Energy (DOE), these facilities are licensed by NRC or the
Agreement States. Nuclear power reactors and DOE enrichment facilities are not part of this
study.
D.3.1 Source Material Licensees (10 CFR 4G)
TTranfiim Mills
Uranium mills extract uranium from ores which contain only 0.01 to 0.3 percent
U3O8. Uranium milk, typically located near uranium mines in the western United States, are
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usually in areas of low population density. The product of the mills is shipped to conversion
plants, where it is converted to volatile uranium hexafluoride (UF6) which is used as feed to
uranium enrichment plants.
As of December 1988, of 27 uranium mills in the United States licensed by NRC or
the Agreement States, 4 were operating, 8 were shut down, 14 were being decommissioned,
and 1 had been built but never operated. The eight shut down mills could resume
operations, but the 14 mills that are being decommissioned will never operate again.
The operating mills have a capacity of 9,600 tons of ore per day. The number of
operating mills is down considerably from 1981, when 21 mills were processing
approximately 50,000 tons of ore per day. This reduction reflects the decrease in the
demand for yellowcake. The mined ore is stored on pads prior to processing. Crushing and
grinding and a chemical leaching process separate the uranium from the ore. The uranium
product is dried and packaged following recovery from the leach solution. The waste
product (mill tailings) is piped as a slurry to a surface impoundment area (tailings pile).
Radioactive materials released to the air during these operations include natural
uranium and thorium and their respective decay products (e.g., radium, lead, radon). These
radionuclides, with the exception of radon, are released as particulates.
Depleted ITrairhim Munitions Testing Facilities
The processing of natural uranium to obtain uranium enriched in the U-235 isotope
results in abundant tails referred to as depleted uranium. Its ownership, possession, and use
is licensed by NRC as source material. The density and low specific activity of depleted
uranium make it useful for several applications, including radiological shielding,
counterweights in aircraft, and in military munitions. This latter activity has the greatest
potential to result in airborne release of radioactive material.
Depleted uranium is used by the military in munitions designed to pierce armor
plating. The design of these munitions is developed and refined by the army based on "soft"
and "hard" testing. Soft testing is conducted to define and refine the accuracy of the
munitions, and is conducted on outdoor firing ranges where the depleted uranium round is
fired at the "target" located in a sand-filled testing enclosure located several kilometers from
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the gun. After impact, the depleted uranium "rod," which is generally intact, is simply left
in the ground as the risk from unexploded munitions makes retrieval too dangerous. Hard
testing is conducted to evaluate and refine the destructive capability of the munitions. In
hard testing, either actual munitions or scale mockups are fired at an armor-plated target. By
license conditions, all hard testing of depleted uranium munitions is conducted in indoor test
enclosures, the ventilation stacks of which are equipped with roughing and HEPA filters; the
exhaust is monitored during testing.
The Department of Defense conducts testing of depleted uranium munitions at a
number of proving grounds around the country. The U.S. Department of the Army's
Ballistic Research Laboratory and Combat Systems Test Activity facilities at the Aberdeen
Proving Ground in Aberdeen, Maryland conduct both hard and soft testing. Soft testing is
also conducted by the Army at the Yuma Proving Ground near Yuma, Arizona, and at the
Jefferson Proving Ground near Madison, Indiana, and the Navy conducts soft test firings at
the China Lake Weapons Testing Site near China Lake, California. Occasionally, on the
order of once every two or three years, the Army conducts an open-air hard test firing at the
Nevada Test Site. These munitions are used only during actual hostilities, not during
training or exercises.
Uranium Conversion Facilities
i • .
i .
The uranium conversion facility purifies and converts uranium oxide (U3O8 or
yellowcake) to volatile uranium hexafluoride (UF6),the chemical form in which uranium
enters the enrichment plant.
Currently 2 commercial uranium hexafluoride (UF6) production facilities are operating
in the United States: the Allied Chemical Corporation facility at Metropolis, Illinois and the
Kerr-McGee Nuclear Corporation facility at Sequoyah, Oklahoma. The Allied Corporation
facility, a dry-process plant in operation since 1968, has a capacity to produce about 12,600
mt of uranium per year in the form of TJF6. The Kerr-McGee Nuclear Corporation facility is
a wet-process plant in operation since 1970, with a capacity of about 9,100 mt per year.1
1 U.S. Atomic Energy Commission, Fuels and Materials Directorate of licensing, Environmental Survey
of the Uranium. Fuel Cycle, April 1984, and W. Dolezal, personal communication with D. Goldin, SC&A, Inc.,
September 1988. ,
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Two industrial processes are used for uranium hexafluoride production, the dry
hydrofiuor method and the wet solvent extraction method. Each method produces roughly
equal quantities of uranium hexafluoride; however, the radioactive effluents from the two
processes differ substantially. The hydrofiuor method releases radioactivity primarily in the
gaseous and solid states, while the solvent extraction method releases most of its radioactive
wastes dissolved in liquid effluents.
• Dry Hydrofluor Process
This process consists of reduction, hydrofluorination, and fluorination of concentrated
ore to produce crude uranium hexafluoride. Fractional distillation is used to obtain
purified UF6. Impurities are separated either as volatile compounds or as a relatively
concentrated and insoluble solid waste that is dried and drummed for disposal.
• Solvent Extraction Process
The solvent extraction process employs a wet chemical solvent extraction step at the
start of the process to prepare high purity uranium for the subsequent reduction,
hydrofluorination, and fluorination steps. The wet solvent extraction method
separates impurities by extracting the uranium from the organic solvent, leaving the
impurities dissolved in an aqueous solution. The raffinate is impounded in ponds at
the plant site.
Processing Facilities <•
Rare-earth elements are metals with distinct individual properties which make them
potentially valuable as alloying agents. The name rare earths is deceiving, however, because
they are neither rare nor earths. Rare earth minerals exist in many parts of the world, and
the overall potential supply is essentially unlimited. The term earth stems from the fact that
the elements were first isolated from their ores in the chemical form of oxides and that the
old chemical terminology for oxide is earth. The rare earths (also called Lanthanides) form
trivalent bonds, and when their salts are dissolved in water, they ionize to form trivalent ions
and the solutions exhibit very similar chemical properties, sharing a valence of three. Rare
earths are widely distributed in nature, although they generally occur in low concentrations.
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Approximately 10 facilities are engaged in the recovery of metals from source
materials. Rare earth facilities with NRC Source Material Licenses process natural and
synthetic ores which contain at least 0.05 percent, by weight, of naturally occurring uranium
and thorium. The principal environmental impacts of rare earth facility operations include
the potential release of radioactive particles and radon from the storage, handling, and
processing of the ores. The operation of a rare earth facility involves grinding, dissolving,
and processing the natural and synthetic ores. The ores are fed into a grinding machine
where they are reduced into particle size. Dust from this process goes to a dust collector
which recycles the dust back into the system, scrubs it, then releases it into the environment.
Because this process is relatively closed, it is generally believed that very limited amounts of
radioactivity escape. The reduced ores are transferred through pipes into digester tanks
which contain acid that selectively dissolves the ores. The unwanted uranium and thorium
react with the acid to form insoluble uranium and thorium fluorides. Different facilities have
different processes by which they store the radioactive wastes. It is often stored onsite in
barrels or slag piles.
D.3.2 Special Nuclear Material Licensees (10 CFR 7G) ' . .
LWR Fuel Fabrication Facilities
Light water reactor (LWR) fuels are fabricated from uranium which has been
enriched in U-235. At a gaseous diffusion plant, natural uranium in the form of UF6 is
processed to increase the U-235 content from 0.7 percent up to 2 to 4 percent by weight.
The enriched uranium hexafluoride product is shipped to LWR fuel fabrication plants where
it is converted to solid uranium dioxide pellets and inserted into zirconium alloy (Zircaloy)
tubes. The tubes are fabricated into fuel assemblies which are shipped to nuclear power
plants. There are seven licensed uranium fuel fabrication facilities in the United States which
fabricate commercial LWR fuel. Of the seven, only five had active operating licenses as of
January 1, 1988. Of those five facilities, two use enriched uranium hexafiupride to produce
completed fuel assemblies and two use uranium dioxide. The remaining facility converts
UF6 to UO2 and recovers uranium from scrap materials generated in the various processes of
the plant. .
The processing technology used for uranium fuel fabrications consists of three basic
operations: (1) chemical conversion of UF6 to UO2; (2) mechanical processing including
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pellet production and fuel element fabrication; and (3) recovery of uranium from scrap and
off-specification material. The most significant potential environmental impacts result from
converting UF6 to UO2 and from the chemical operations involved in scrap recovery.
Non-LWR Fuel Fabrication Facilities
Only a few facilities produce the metal and mixed carbide fuel used in test and
research reactors.
The non-oxide fuel fabrication process begins with highly enriched uranium metal.
The uranium metal may be mixed with an alloying metal in an induction furnace. The fuel is
then either rolled, punched, drilled, or crushed and compacted, and machined and shaped
into the proper dimensions. Once the fuel is properly formed, it is enclosed in aluminum or
stainless steel. The enclosing process may involve injection casting, loading into a can or
mold, or simply covering the fuel with side plates and rolling the metals together. Finished
fuel elements are then inspected and cleaned prior to assembly into fuel bundles:
The production of mixed carbide fuel starts with highly enriched uranium dioxide-
thorium dioxide powder (UO2-TbO^, This powder is mixed with graphite and heated to
form uranium-thorium carbide kernels. These kernels are formed into microspheres by
heating to a temperature in excess of the kernels' melting point. The microspheres are then
coated with carbon and silicon layers in a fluidized bed furnace. Fuel rods are formed by
injecting the coated kernels and a matrix material into a hot mold. The finished rods are
then inserted into a graphite block to form the final fuel assembly.
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APPENDIX E
QUALITY ASSURANCE CRITERIA FOR NUCLEAR POWER PLANTS
AND FUEL REPROCESSING PLANTS
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APPENDIX E
x
QUALITY ASSURANCE CRITERIA FOR NUCLEAR POWER PLANTS
AND FUEL REPROCESSING PLANTS
Quality assurance (QA) comprises all those planned and systematic actions necessary
to provide confidence that a component will perform satisfactorily in service. Given the
diversity of NRC-licensed facilities other than nuclear power reactors and the necessity to
structure QA programs suited to the function of a facility, QA programs are themselves
diverse, bearing closer resemblance to the highly structured power reactor programs as the
complexity and risk potential of a facility increases.
QA programs must be documented by written policies, procedures, or instructions and
must be carried out throughout the plant life. The QA program provides control over
activities affecting the quality of components to an extent consistent with their importance to
safely. The program must provide for the indoctrination and training of personnel
performing activities affecting quality.
The QA criteria applicable to the power reactor program are listed below.1 The
purpose of each of the 18 QA criteria is briefly explained in the following pages. Some or
all of the principles noted may apply in total or in part to NRC-licensed facilities other than
nuclear power reactors. Refer to the individual paragraphs in the Code of Federal
Regulations (10 CFR 30-39, 40, 50, and 70) for specific requirements.
Criterion 1 - Organization - To identify all activities affecting quality and to assure that the
responsibilities and authorities of key personnel are clear.
Criterion 2 - Quality Assurance Program - To cause the project manager to articulate the
actions necessary to plan and implement an effective quality assurance program.
Criterion 3 - Design Control - To control the following processes in accordance with the
requirements of Applicable and Relevant or Appropriate Requirements: (1) designing tests
and sampling patterns to characterize the geologic setting, to develop models to predict the
1 Appendix B to 10 CFR 50, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing
Plants." •
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performance and long-term stability of the site, and to predict the environmental interaction
between the site and its surroundings; (2) specifying requirements for design and
construction; and (3) designing computer codes.
Criterion 4 - Procurement Document Control - ,To provide the management controls to
manage the work activities of contractors and subcontractors and ensure acceptable quality of
the results. ~
Criterion 5 - Instructions, Procedures, and Drawings - To ensure the use of formal
instructions for work activities related to the accomplishment of performance objectives and
the design bases.
Criterion 6 - Document Control - To ensure that documents prescribing activities related to
the accomplishment of the performance objectives and the design bases are controlled during
review, approval, and distribution to ensure that those performing activities use approved and
up-to-date instructions. .
Criterion 7 - Control of Purchased Material, Equipment, and Services - To oversee and
control the work of contractors and suppliers and to ensure that the results are consistent with
performance objectives and design bases.
Criterion 8 - Identification and Control of Materials, Parts, and Components - To ensure
that all materials, parts, samples, and components important to the accomplishment of
performance objectives and the design bases are identified and controlled.
Criterion 9 - Control of Special Processes - To ensure that all work activities important to
the accomplishment of performance objectives and the design bases are controlled, including
the identification of activities that require specially trained personnel, or specialized
equipment or procedures.
Criterion 10 - Inspection - To ensure that independent, pre-planned inspections are performed
where it is deemed necessary to establish the acceptability of a product, process, or service,
either in progress or upon completion. _• '
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Criterion 11 - Test Control - To ensure that tests are conducted to determine if an item or
service is acceptable or to satisfy a need for more information.
Criterion 12 - Control of Measuring and Test Equipment - To ensure that measurements that
affect quality of work are taken only with instruments, tools, gauges, or other measuring
devices that are accurate, .controlled, calibrated, and adjusted at predetermined intervals to
maintain accuracy within necessary limits.
Criterion 13 - Handling, Storage, and Shipping - To ensure control over handling, storage,
cleaning, packaging, preservation, and shipping of items affecting quality of work.
Criterion 14 - Inspection, Test, and Operating Status - To ensure the identification of the
inspection and/or test status of samples, structures, systems, and components to prevent
inadvertent use of items found to be unacceptable for use.
Criterion 15 - Nonconforming Materials, Parts, or Components - To ensure that items not
conforming to specified requirements are identified and controlled to prevent inadvertent use.
Criterion 16 - Corrective Action - To ensure that management systems comprised by the QA
program are constantly monitored and that timely measures are taken to correct conditions
adverse to quality.
Criterion 17 - Quality Assurance Records - To ensure that records important to the
accomplishment of performance objectives and the design bases (including the data analysis
phase, hearings, permitting and licensing processes) are sufficient to demonstrate the quality
Of work performed. Records will also be needed should problems related to the performance
of the facility occur at a later date.
Criterion 18 - Audits - To ensure that audits, which are part of the management system's
sensors, are effective by being well planned, conducted by trained personnel familiar with the
work being audited, and designed to measure the potential of the activity or process being
audited to produce an acceptable product.
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APPENDIX F
NRC AGREEMENT STATES AND STATE DIRECTORS
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APPENDIX F
NRC AGREEMENT
-------
GRBEMEisrr STATES AKD STATE DIRECTORS*
STATE
7. Georgia
James L. Setser, Chief
Environmental Protection
Department of Natural Resources
Floyd Towers East 1166
205 Butler Street
Atlanta, Georgia 30309
(404)656-4713
8. Illinois
Thomas W. Ortcigar, Director
Department of Nuclear Safety
1035 Outer Park Drive
Springfield, Illinois 62704
(217)785-9868
9. Iowa
Donald A. Hater, Chief
Bureau of Radiological Health
Department of Public Health
Lucas State Office Building
Des Moines, Iowa 50319
(515)281-3478
10. Kansas
Mr. Gerald W. Allen, Chief
X-Ray & Radioactive Materials
Department of Health & Environment
109 S.W. 9th Street
Topeka, Kansas 66620
(913)296-1562
11. Kentucky
Mr. John Volpe, Manager
Radiation Control Branch
Department of Health Services
Cabinet for Human Resources
275 East Main Street
Frankfort, Kentucky 40621
(502)564-3700
12. Louisiana
Mr. Glenn Miller, Administrator
Radiation Protection Division
Office of Air Quality & Nuclear Energy
P.O. Box 82145
Baton Rouge, Louisiana 70884
(504)765-0160
13. Maryland
Mr. Roland G. Fletcher, Administrator
Radiological Health Program
Office of Toxics, Environmental Science and Health
Department of the Environment
2500 Broening Highway
Baltimore, Maryland 21224
(301)631-3300
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NEC AGREEMENT STAGES AND STATE DIRECTORS*
STATE
14. Mississippi
Mr. Eddie S. Fuente, Director
Division of Radiological Health
State Board of Health
3150 Lawson Street
P.O. Box 1700
Jackson, Mississippi 39215-1700
(601)354-6657/6670
15. Nebraska
Mr. Harold Borchert, Director
Division of Radiological Health
State Department of Health
301 Centennial Mall South
P.O. Box 95007
Lincoln, Nebraska 68509
(402)471-2168
16. Nevada
Mr. .Stanley Marshall, Supervisor
Radiological Health Section, Health Division
Department of Human Resources
505 East King Street, Room 202
Carson City, Nevada 89710
(702)885-5394
17. New Hampshire
Ms. Diane Tefft, Program Manager
Radiological Health Program
Bureau of Environmental Health
Division of Health Services
Health & Welfare Building, Hazen Drive
Concord, New Hampshire 03302
(603)271-4588
18. New Mexico
Benito J. Garcia, Chief
Community Services Bureau
Environmental Improvement Division
Department of Health & Environment
P.O. Box 968
Sante Fe, New Mexico 87504-0968
(505)827-2959 •
19. New York
Ms. Donna Ross, Energy Planner
Division of Policy Analysis and Planning
2 Rockefeller Plaza
Albany, New York 12223
(518)473-0048
20. North Carolina
Mr. Dayne H. Brown, Director
Department of Environment, Health and Natural Resources
Division of Radiation Protection
P.O. Box 27687
Raleigh, North Carolina 27603
(919)741-4283 - .
F-4
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3&C AGREEMENT STATES ANJJ STATE DIRECTORS*
STATE
21. North Dakota
22. Oregon i
23. Rhode Island
24. South Carolina
25. Tennessee
26. Texas
27. Utah
; •' , sTATBPHffiCTQR,
Mr. Dana Mount, Director
Division of Environmental Engineering
Radiological Health Program
State Department of Health
1200 Missouri Avenue ,
Bismarck, North Dakota 58502
(701)221-5188 -
Mr. Ray Paris, Manager
Radiation Control Section
Department of Human Resources
1400 South West Fifth Avenue .
Portland, Oregon 97201
(503)229-5797 .
Shelly Robinson, Acting Chief
Radioactive Materials & X-Ray Programs
'Department of Health
Cannon Building, Davis Street
Providence, Rhode Island 02908
(401)277-2438
Mr. Heyward G. Shealy, Chief
Bureau of Radiological Health , ,
Department of Health and Environmental Control
J. Marion Sims Building
2600 Bull Street
Columbia, South Carolina 29201
(803)734-4700 * . .
Mr. Michael H. Mobley, Director
Division of Radiological Health
TERRA Building, 150 9th Avenue North
Nashville, Tennessee 37219-5404
(615)741-7812
Mr. David K. Lacker, Chief -
Bureau of Radiation Control
Department of Health
1100 W. 49th Street (mail only)
Austin, Texas 78756
(512)835-7000
Mr. Larry Anderson, Director
Bureau of Radiation Control
State Department of Health
288 North 1460 West
P.O. Box 16690
Salt Lake City, Utah 84116-0690
(801)538-6734
F-5
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STATE
28. Washington
Mr. Teny R. Strong, Director
Office of Radiation Protection
Department of Health
Mail Stop LE-13
Olympia, Washington 98504
(206)586-8949
* As of August 1991
F-6
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APPENDIX G
RANDOM SURVEY QUESTIONNAIRE
G-l
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APPENDIX G
UNITED STATES ENVIRONMENTAL PROTECTION AGENCY
WASHINGTON, D.C. 20460
OFFICE OF
AIR AND RADIATION
On October 31, 1989, the U.S. Environmental Protection
Agency (EPA) promulgated standards controlling radionuclide air
emissions from facilities licensed by the Nuclear Regulatory
Commission, or certain agreement States. This regulation is
under reconsideration and the Agency needs to gather information
to determine whether or not these standards should be put into
effect. The facilities being studied are licensed to handle or
use radioactive materials in unsealed form. This facility has
been selected to take part in a study to determine the radiation
hazard to individuals residing outside the facility. Please fill
out the enclosed form and return it within 30 days of the receipt
of this request to:
Dale Hoffmeyer
U.S. Environmental Protection Agency
Mail Code ANR 460W
401 M Street, SW
Washington, DC 20460
This information is being requested under Section 114 of the
Clean Air Act. Under Section 114 of the Act, the Administrator
has the authority to require any person to submit information to
assist EPA in developing National Emission Standards for
Hazardous Air Pollutants under Section 112.
Please be advised that failure to provide all the
information required by this Reporting Requirement within the
time allowed, or to provide adequate written justification for
such failure, can result in enforcement action by EPA against you
under Section 113 of the Clean Air Act. Such enforcement may
include a civil action for the assessment of monetary penalties.
You should also be aware that Section 113 provides for possible
criminal sanctions for anyone who knowingly makes any false
statement, representation, or certification in a report required
by EPA. ,
G-2
Printed on flecycted Paperl
-------
You may assert a business confidentiality claim covering
part or all of. the information responsive to this Reporting
Requirement in the manner described in 40 C.F.R. Section
2.203(b). EPA will disclose information covered by such a claim
only to the extent and according to the procedures set forth in
40 C.F.R. Part 2, Subpart B. If you do not submit a
confidentiality claim with the information, EPA may disclose your
response ,without further notice to you. You should read the
above-cited regulations carefully before asserting a business
confidentiality claim, since certain categories of information
are not properly subject to such a claim.
If you have any questions concerning this letter or if you
would like assistance in completing the form, call (800-685-3339)
from 9 a.m. to 5 p.m. eastern standard time.
Sincerely,
Enclosure
William G. Roserierg
Assistant Administrator
for Air and iCadiation
G-3
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SURVEY FORM
Facility Name_
Address
City
State
Identify a person from whom clarification or additional information can be obtained, if necessary.
Name _ _ Telephone ( _ ) _ b _
^
Does your facility hand'9 only sealed radiation sources? .
Sealed sources include "Special Form" sources that are sealed and not intended to be opened in
their routine application; e.g.. density and thickness gages.
Yes _ STOP You do not have to complete the remainder of this form. However, you must return
the form to the EPA.
No _ CONTINUE You must complete the remainder of this form.
Indicate the principal activities conducted at your facility which involve unsealed forms of
radlonuclides (check all that apply): .
[ ] Accelerator
[ j Research/Test Reactor "
[ ] Nudear Medicine (Diagnostic only)
[ ] Nudear Medicine {Diagnostic and Therapeutic)
[] Manufacturer of Teletherapy Equipment
[ ] Manufacturer of Medical Implant Needles or Seeds
[ ] Manufacturer of Pacemakers
[ ] Manufacturer of Industrial/Scientific Gauging Equipment
[ j Manufacturer of Self-Illuminating Devices
[ ] Producer of RadiopharmaceuticaJs
[ ] Producer of Radio-Labelled Compounds for Research
[ J Producer of Munitions using Depleted Uranium
[ 1 Producer of Shielding using Depleted Uranium
[ ] Thorium/Rare Earth Processing/Recovery
[ ] Low-Level Waste Disposal Facfltty
[ ] Low-Level Waste Incinerator
[ ] Low-Level Waste Transfer Agent (Prepackage only _ .re-package _ )
[ 1 Research Laboratory (Indicate flekJ of research__ _ ; _ '. _ )
[ ] Other (please specify _ -. _ ' - _ — , - )
GENERAL INSTRUCTIONS
1. You must provide the information requested on pages 2-3 of this form separately fof
building at your facility where radtonuclkjes In unsealed torm are handled.
2. If a question does not apply to your facility, then mark the appropriate space "N/A\ if you
cannot answer a question, mark the appropriate space "IT.
G-4
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BUILDING NAME
Step 1. " Provide the approximate building dimensions in meters.
Length_
Width
Height
Step 2.
(if the building is irregularly shaped, the length and width should be those of the smallest
rectangle that in plan view would completely encompass the building; the height should be
the distance from the ground to the highest roof.)
Provide the following information for each stack/vent that serves an area in the building
where unsealed forms of radioactive materials are handled, «u»'«mg
Height
(m)'
Diameter
(m)a
Flow Rate
(m3/sec)
Temperature
Effluent Controls
(specify type)
/« *•" Iff7"!?1 ^^ *" """ft'*** h«.__ and provid. th. information for th. addition*
stacks/vents on a separate sheet of paper which deany designate* the buliding name.
'Distance from the ground to the top of the stack/vent
*lf the sta<*/vem is rectangular, give its lertgth and width.
l^ exituttmP«r«tur« i8 «PP«ainwWy th. same as th. ambi.nt t.mp.fatur.. simply «nt^ an A for ambi.nt.
NOTE H th. dau you hav. is in units othtf than tho« r*,u.st«d and you art uncartain of th. conversion, provid. your
data wrthth. units d»arly indicatsd; e.g. ft for fwt. CFM for f^/min. and °C for degrMs Celsius
Step 3.
Does anyone live in this building?
If YES. enter the distance along the buiding surface from the stack/vent to the nearest residence in the
bunding (meters)
If NO, enter the distance from the stack/vent to the nearest residence outside the building.
, (meters) Indicate the direction from the stack/vent to the nearest residence
(N.NNE.NE eta) '.-—
Step 4. Is there an office, school or business, not part of the facility covered by the NRC or
state license, in this building?
If YES. enter the distance along the buBding surface from the stack/vent to the nearest office, school
or business. (meters)
If NO. enter the distance from the stack/vert to the nearest office, school or business outside the
building. (meters) Indicate the direction from the stack/vent to the nearest office, school or
business.__ (N.NNE.NE etc.)
Step 5. Provide the distances In meters to the following sources of food production.
If the distance is greater than 2000 meters, then errter >2000.
Vegetables
Meat
MOk
G-5
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Step 6.
Complete the table below.
You must report §J1 radionuclkles in unsealed forms used in the building even if you do not
believe that they are being released.
If you cannot provide the emission or quantity used information requested below for each
stack/vent, then enter the values for the entire building under the column for stack 1.
MONITORED STACKS/VENTS
If you measure radionuclide emissions, enter the release rate for each radionudide in Cl/y.
Designate that it is a measured value by entering an M in the column headed *M/Q". and leave the
column headed 'Physical Form' Wank. If radionudide emissions are below the minimum detectable
level, enter that level (Cl/y) preceded by a < symbol.
UNMONITORED STACKS/VENTS
if emissions of any radionuclides used are not measured, enter the quantity of each radionudide
used (Cl/y) but not measured and enter a Q in the column headed 'M/Q'. In the column headed
•Physical Form', enter G for radtonudWes that are gases or subject to temperatures in excess of 100°;
L for radionudides that are handled in liquid or powder forms; and S for radtonudldes handled in solid
forms or capsules. (Mo-99 contained in a generator to produce Technetlum-99m can be assumed to
be a solid.) if a radionudide is used in more than one physical form, provide a separate entry for each.
Note: For both monitored or unmonftored emissions, if you know the lung dearance dass for a
radionudide. enter it in the table. Use D for days, W for weeks, and Y for years, if the chemical
species of a radionudide falls into more than one dearance dass, make a separate entry for each. If
the lung dearance dass is not known, enter the predominate chemical spedes tf known.
Nudide
M/Q
Physical
form
Clearance
dass
Dates Covered:
Stack 1
(Ci/Y)
Stack 2
(Cl/y)
From To
Stack 3
(Ci/y)
Stack 4
(ci/y)
Stacks
.(d/y).
!===
additional r»dlooudidw on a
p«p«c which dMity <**&*** trw.tHrikfing rwnw.
G-6
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APPENDIX H
DOSE CALCULATION ASSUMPTIONS
This appendix provides details of calculational assumptions made regarding
factors having a significant effect on dose.
H-l
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Contents
H.I Assumptions Related to Source Term . . ................... . • ..... H-3
H.2 Assumptions Related to Dispersion . ....... ...... • • • • .......... H~5
H.3 Assumptions Related to the Receptor . ....... ................... H-6
H.4 References . . ................... ...... ....... ........ H-ll
Tables:
Table H-l. Doses above 1 mrem/yr
Table H-2. Doses above 1 mrem/yr, no wind rose
H-5
H-2
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APPENDIX H
DOSE CALCULATION ASSUMPTIONS
H. 1 ASSUMPTIONS RELATED TO SOURCE TERM
Xe-133 Release from Hosmtals
All the Xe-133 used by hospitals was assumed to be released. This assumption is
appropriate for most hospitals but tends to overestimate the dose for others. While many
hospitals trap the Xe-133 exhaled by the patients and allow it to decay for a number of half-
lives, only a few of the hospitals surveyed indicated that they did so. Trapping the gas
reduces the amount available to be released into the environment.
In order to properly account for Xe-133 trapping, it would have been necessary to
contact each hospital to determine the details of its procedures. Because Xe-133 was the
principal contributor to dose for many of the hospitals, a reduction in Xe-133 release would
lower the median dose of the population. However, it would not have much effect on the
doses above 1 mrem/yr as shown below:
Table H-L Doses above 1 mrem/yr.
Facility
NH
NH
HN
H
H
NH
NH
As Calculated
1.1
1.7
1.8
2.0
3.9
5.3
8.0
AU3&-133 Tipped '
1.1
1.7
1.8
0.9
0.6
5.3
8.0
NH = non-hospital, H = hospital, HN = hospital, no Xe
Twenty-three facilities have estimated doses above 0.1 mrem/yr. Of these 23, Xe-133
is a contributor in five. Of these five, the Xe-133 contribution to the total dose is 30, 40,
50, 80, and 85 percent, for an average of approximately 60 percent.
H-3
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It is concluded that neglecting Xe-133 trapping by hospitals has a negligible effect
upon the distribution of estimated doses.
• Release Fractions
For those facilities that did not provide site-specific release rates, the default release
fraction of 1E-03, described in EPA89b, was used for liquids. As applied to nonvolatile
radionuclides, such as Tc-99m or buffered solutions of radioiodine, this assumption will
generally result in a higher estimate of the radionuclide release rate.
• » Xe-133 Release from Radiopharmacies
EPA89b specifies a release fraction of 1.0 for radionuclides in .gaseous form.
However, because radiopharmacies receive and distribute the Xe-133 in sealed vials, very
little is released. The Food and Drug Administration's limit on the leakage from these vials
is 0.5 percent per day; however, in practice, the measured leakage is a maximum of 0.1
percent per day (Mu91).
The total leakage is a function of both the release rate (percent per day) and the
length of time the vial is held in stock. Because Xe-133 has a half-life of only five days, it
is unlikely that it would be held in stock for very long. If it were to be held for 10 days, the
amount would have decayed to only one fourth the amount received by the radiopharmacy.
A 0.1 percent per day leakage rate and a holding time of 10 days was assumed. This
results in a release fraction of one percent. ,
• Emissions from Sources Other Than Stacks and Vents
Radionuclide air emissions from stacks and vents were considered in this study, but
emissions from diffuse sources were not covered. These include: fugitive emissions from
normal operations (e.g., releases from patients treated with radionuclidesj; spilling and
mishandling; and more catastrophic accidental releases such as fires and explosions.
Exposures from these sources could make some of the annual doses actually received by
members of the general public greater than those calculated.in this study. However, the
contribution of these sources to lifetime risks are generally believed to be low because the
occurrences are usually infrequent and the exposures are for a short duration.
H-4
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H.2 ASSUMPTIONS RELATED TO DISPERSION
• Closest Person Versus Maximally Exposed Person
In calculating the doses to the maximally exposed individual, the distance and
direction to the closest office, school, or business were used; It is possible that an individual
located at a greater distance • but in a sector toward which the wind blows more frequently,
could receive a higher dose. However, should such a circumstance arise, the dose would be
underestimated by no more than a factor of about 5.
The preceding discussion applies only to those cases in which a wind rose was used.
If the closest person was on the same building, a wind rose was not used. For this reason,
there is only a minimal effect on the doses above 1 mrem/yr as shown below.
Table H-2. Doses above 1 mrem/yr, no wind rose.
\ FaciStjr -'
MR
NR
RW
MR
MR
NR
MR
- As Calculated
1.1
1.7
1.8
2.0
3.9
5.3
8.0
, Person ia Max Sector
1.1
1.7
2.0
2.0
3.9
5.3
8.0
NR =• no wind rose used (same building); RW = wind rose, near wake
Of the 23 facilities having doses greater than 0.1 mrem/yr, 15 have the closest
residence, office, or classroom in the same building, five have them within the near-wake
region, and three have them outside the near-wake region. The ratios of the maximum to the
closest receptor for these are 1.0, 1.5, and 2.1, for an average of 1.5.
It is concluded that calculating the dose to the person in the closest residence, office
or classroom, rather than in the location of maximum dose, had a negligible effect upon the
distribution of doses. -
H-5
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The estimate of the air concentration when the source and receptor are on the same
building is quite conservative. The model used by COMPLY is based on NCRP
Commentary No. 3. NRCP based its model on a study by Wilson and Britter (Wi82), which
found that the concentration at various locations on a building was a function of the wind
speed and the distance (measured along the building surface between the source and the
receptor). The correlation is C/Q = B/ux2, where C/Q is the normalized concentration, B a
constant, u the wind speed, and x the distance between the source and the receptor.
Wilson and Britter suggest a value of 9 for B unless both the source and receptor are
on the lower third of the same or adjacent walls, in which case they suggest a value of 30.
The NCRP model uses 30 for all cases.
While the correlation based on these parameters seems reasonable, their data show a
great deal of scatter. With B = 9, more than 90 percent of the data points lie above the
correlating line. This means that their correlation encompasses more than 90 percent of the
data points; it is not a mean line. The mean line lies about a factor of 5 above their line.
That is, using the mean line would lead to B being about 1.4.
The NCRP method tends to overestimate dose. However, this has utility for
regulatory purposes, as it means there is only a small chance that a facility might appear to
be in compliance with the limit when the true concentration'would result in a dose exceeding
the limit.
H.3 ASSUMPTIONS RELATED TO THE RECEPTOR
• Age and Select Populations
Following the recommendations of the International Commission on Radiation
Protection (ICRP80), this study assumed that doses were delivered to a standard man. In
most cases, we have no information on the age or the predisposition of the exposed
population which would lead us to conclude that either doses or risks would be greater than
those estimated. However, for certain exposure groups considered in this study, such as
students attending school, the average age may be less than that of standard man and the
H-6
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annual dose may be greater than that calculated. The models used by EPA do not explicitly
account for these factors, but since the cancer risks are assumed to result from a lifetime of
exposure, underestimates assockted with age at time of exposure would tend to be mitigated.
• Dose Conversion Factors
The dose conversion factor (DCF) is one of the key parameters used to calculate the
doses assockted with radionuclide emissions from facilities licensed under 10 CFR 30. The
DCFs used in this report establish the relationship between a given intake of a radionuclide
or concentration in the environment and the dose to a person exposed to the radionuclide.
For radionuclides that are either inhaled or ingested, the DCF is expressed in units of the
dose per unit activity inhaled or ingested. The values are isotope specific and are typically
expressed in units of Sv/Bq1 or mrem/Ci inhaled or ingested. For external exposures, the
DCFs are expressed in units of dose rate per unit radionuclide concentration in the
environment. For example, the DCFs for external exposure assockted with immersion in a
cloud of radioactivity are often expressed in units of mrem/yr per Ci/m3. For external
exposure from activity deposited on the ground, the DCF is typically expressed in units of
mrem/yr per Ci/m2.
DCFs are convenient values because, once the inhalation rate or ingestion rate of a
given radionuclide is determined, the internal dose is readily obtained by multiplying by the
appropriate DCF. Similarly, once the concentration of a given radionuclide in air or on the
ground is determined, the external dose rate from immersion or direct radiation from
standing on the contaminated ground is readily obtained by multiplying by the appropriate
DCF.
Imbedded in the COMPLY code are default values for the DCFs for virtually all
radionuclides for inhalation, ingestion, and external exposure. The purpose of this section is
to explore the degree of conservatism, if any, inherent in these DCFs as used in this project.
The discussion is divided into three parts: Inhalation DCFs, Airborne Immersion DCFs, and
DCFs for External Exposure to Deposited Radionuclides. The discussions focus on the
radionuclides, pathways, and facilities found to be the most significant on this project.
1 Sievert (Sv) is the international system unit of any of the quantities expressed as dose equivalent. The dose
equivalent in sieverts is equal to the absorbed dose in grays multiplied by the quality factor (1 Sv = 100 rems).
One becquerel (Bq) is equal to 1 disintegration per second.
H-7
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Inhalation DCFs. In the biomedical community, which represents the majority of the
facilities addressed by this project, the principal radionuclides contributing to inhalation
exposures are Tc-99m and 1-131. The inhalation DCFs used by COMPLY are:
Tc-99m 3.26E+04 mrem/Ci or 8.80E-12 Sv/Bq inhaled
1-131 3.29E+07 mrem/Ci or 8.89E-09 Sv/Bq inhaled
These values were obtained from Table 2.1 of Federal Guidance Report No. 11,
which is the EPA guidance regarding DCFs (EPA88). Inspection of Federal Guidance
Report No. 11 reveals that these are committed effective dose equivalent factors (CEDE),
which means that the doses obtained using these DCFs represent the whole body dose
equivalent for the actual dose delivered to a specific organ. For example, exposure to 1-131
results predominantly in a dose to the thyroid gland. However, the 1-131 DCF includes a
weighting factor, which converts the dose to the thyroid gland to the whole body dose that is
equivalent, based on the effects of radioactive material in subsequent years following intake.
In the case of thyroid exposure, the DCF includes a weighting factor of 0.03. The weighting
factor for a tissue represents the proportion of stochastic risk resulting from irradiation of
that tissue compared to the total risk when the whole body is uniformly irradiated.
Therefore, as defined by ICRP, 3 percent of the total risk following whole body exposure is
attributable to the exposure of the thyroid.
The DCFs in Federal Guidance Report No. 11 are 50-year dose commitments. This
means, for a given intake of a radionuclide, the doses calculated using these DCFs are the
effective doses for the 50-year period following intake. Imbedded in these values are
assumptions regarding the clearance rate of the radionuclides from the body, which also bear
on the realism of the DCFs.
The following presents a closer look at the inhalation DCFs for Tc-99m and 1-131.
The Inhalation DCF for Tc-99m. The degree of conservatism inherent in the
inhalation DCF for Tc-99m used on this project must be discussed from two perspectives.
The first has to do with alternative DCFs provided in Federal Guidance Report No. 11 and
the specific alternative selected for use on this project. The second has to do with
conservatism inherent in the selected alternative.
H-8
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Alternatives. Inspection of Federal Guidance Report No. 11 reveals that two different
inhalation DCFs are provided for Tc-99m, 3.26E+04 and 2.67E+04 mrem/Ci inhaled. The
former is referred to as the DCF for lung clearance class D (days) and the latter as the DCF
for lung clearance class W (weeks) aerosols. Two different values are provided because the
DCF differs depending on the clearance class of the Tc-99m. The D value is to be used for
those forms of Tc-99m that are cleared from the lung relatively quickly, on the order of
days. The W value is to be used for those forms of Tc-99m that are cleared from the lung
more slowly, on the order of weeks. In COMPLY, the higher value was selected. As
discussed in the following, the higher DCF is the more appropriate value to use for the
chemical forms of Tc-99m used by the biomedical community.
Inspection of Federal Guidance Report No. 11 and ICRP 30 reveals that the inhalation
DCF for Tc-99m is based on an assumed aerosol size distribution of 1 micron activity
median aerodynamic diameter (AMAD) and a GI absorption fraction of 0.8, and that the
Tc-99m is in the pertechnetate form. The assumption that the aerosol is 1.0 pm AMAD does
not significantly affect the DCF because the DCF is based primarily on deposition and
retention of transportable technetium. The GI absorption fraction of 0.8 is conservative as
applied to many of the forms of Tc-99m that are not soluble, such as sulphur-colloid, but
appropriate for the pertechnetate form. As discussed below, since the pertechnetate form is
the most commonly used, this is a reasonable assumption. Finally, in developing the
metabolic models for Tc-99m, a broad range of different forms of Tc-99m was considered.
It was assumed that for both the W and D forms of inhaled Tc-99m, once absorbed, the
retention of Tc-99m will follow that of the pertechnetate form. Of the various forms of
Tc-99m, the dose equivalent for the pertechnetate form is generally higher than that of the
other forms (ICRP87). In addition, it is the most widely used form of Tc-99m.
Inhalation DCF for 1-131. The effective whole body DCF for the inhalation of 1-131
is 3.29E+07 mrem/Ci. The value is based on the assumption that 100 percent of the inhaled
iodine is absorbed (i.e., fl = 1), 30 percent goes to the thyroid gland (i.e., f2 = 0.3), and
the remainder is immediately excreted in the urine. The portion that goes to the thyroid
gland is assumed to have an effective half-life of 7.5 days. These values are best estimates
based on extensive experience with 1-131. The fl value of 100 percent is appropriate
because the iodine is easily absorbed. The f2 value of 0.3 is consistent with, though
somewhat more conservative than, the normal range of 0.05 to 0.25 referred to in ICRP90.
The effective half-life of 7.5 days is determined almost entirely by the 8.04-day radiological
half-life of 1-131 and is therefore highly reliable.
H-9
-------
Overall, the parameters used to calculate the thyroid dose to the typical adult from the
inhalation of 1-131 are realistic. However, the effective whole body DCF does have an
inherent degree of conservatism of about 2 up to 15 fold. The conservatism stems from the
way the thyroid DCF is converted to an effective whole body DCF.
As discussed above, the thyroid DCF is converted into an effective whole body DCF
by multiplying the thyroid DCF by 0.03. The 0.03 value represents the relative radiotoxicity
of a given dose of penetrating radiation to the thyroid gland as compared to the same dose
given to the whole body. The weighting factor of 0.03 was based on data that found that for
a given dose of external whole body radiation, approximately 0.03 of the cancer fatalities
caused by the radiation were due to thyroid cancer. Accordingly, an external dose delivered
to the thyroid gland alone is 0.03 as potentially harmful as the same dose delivered to the
whole body.
The 0.03 weighting factor is appropriate for external exposures. However, there is
evidence that the same dose of radiation delivered internally to the thyroid gland from 1-131
can be a factor of from 2 to as high as 15 less radiocarcinogenic (NAS90) (NRC85c). This
may be because a great majority of the dose to the thyroid gland from 1-131 is due to beta
particles, which deposit a large portion of their energy harmlessly in the colloid contained
within the follicles of the thyroid gland. Others disagree, finding 1-131 and x-rays equivalent
in inducing thyroid cancer. In any case, the effective whole body DCF for 1-131 may be
conservative by a factor of 2 to 15.
External Immersion DCF. The external dose from immersion in Xe-133, Tc-99m,
and 1-131 is an important contributor to the offsite doses associated with routine emissions
from hospitals and other materials licensees. COMPLY uses the external DCFs
recommended by EPA in Table 2.3 of Federal Guidance Report No. 11 and in a DOE
publication (DOE88). These DCFs are based on the assumption that the individual is
immersed in a semi-infinite cloud. In the real world, the cloud is of finite dimensions; the
assumption of a semi-infinite cloud could significantly overestimate the dose. The degree of
conservatism in the DCF depends on the size of the cloud and the energy of the photon
emitted by the radionuclide. For example, for a typical 0.7 MeV -gamma emitter, a plume of
about 1000 meters will act as an effectively semi-infinite cloud. However, the dose from a
plume of 100 meters will be about 1/2 the senu-infinite cloud dose, and the dose from a
plume of about 10 meters in diameter, will deliver a dose 1/10 the semi-infinite cloud dose
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(DOE84). For receptors close to the source, where the dimensions of the plume are
relatively small, the assumption of a semi-infinite cloud will likely introduce at least a two-
fold conservatism.
External DCF from Standing on Contaminated Ground. The external dose from
standing on ground contaminated with Tc-99m and 1-131 is another important contributor to
the offsite doses associated with routine emissions from hospitals and other materials
licensees. COMPLY uses the external DCFs recommended in DOE publications (DOE88).
These DCFs are based on the assumption that the individual is standing on an infinite,
smooth plane. In reality, the contaminated area is of a finite dimension and the ground is
generally not smooth. As a result, the doses derived using DCFs based on an infinite smooth
plane .may overestimate the dose by at least a factor of 2.
H.4 REFERENCES
DOE84 Anderson, D., Editor "Atmospheric Science and Power Production,"
DOE/TIC-27601, 1984.
DOE88 U.S. Department of Energy, "External Dose Rate Conversion Factors for
Calculation of Dose to the Public," EH-0070, July 1988.
EPA88 U.S. Environmental Protection Agency, "Limiting Values of Radionuclide
Intake and Air Concentrations and Dose ConversionFactors for Inhalation,
Submersion, and Ingestion," Federal Guidance Report No. 11, EPA-520/1-88-
020, September 1988.
EPA89b U.S. Environmental Protection Agency, "Procedures Approved for
Demonstrating Compliance with 40 CFR 61, Subpart I," EPA 520/1-89-001,
October 1989.
ICRP80 International Commission on Radiological Protection, "Limits for Intakes of
Radionuclides by Workers," ICRP 30, August 1980. , ,
ICRP87 International Commission on Radiological Protection, "Protection of the
Patient in Nuclear Medicine," ICRP 52, 1987.
ICRP90 International Commission on Radiological Protection, _" Age Dependent Doses
to Members of the Public from Intake of Radioactivity: Part 1," ICRP 56,
1990.
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Mu91
NAS90
NRC85c
Wi82
Mulliris, T. J., DuPont/Merck, letter to S. Beal, December 1991.
National Academy of Sciences, "Health Effects of Exposure to Low Levels of
Ionizing Radiation," BEIR V, NAS/NRC, 1990.
U.S. Nuclear Regulatory Commission, "Health Effects Model for Nuclear
Power Plant Accident Consequence Analysis, Part 1: Introduction,
Integration, and Summary; Part 2: Scientific Basis for Health Effects Model,"
NUREG/CR-4214, August 1985.
Wilson, D. J., and Britter, R. E., "Estimates of Building Surface
Concentrations from Nearby Point Sources," Atmospheric Environment. 16,
2631, 1982.
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