United States Air And Radiation EPA 430-R-92-011
Environmental Protection (6602J) November 1992
Agency
&EPA NESHAPS Rulemaking On
Nuclear Regulator/
Commission And Agreement
State Licensees Other Than
Nuclear Power Reactors
Background Information
Document
-------
-------
40 CFR Part 61
National Emission Standards
for Hazardous Air Pollutants
EPA 430-R-92-011
NESHAPS RULEMAKING ON NUCLEAR
REGULATORY COMMISSION AND AGREEMENT STATE LICENSEES
OTHER THAN NUCLEAR POWER REACTORS
BACKGROUND INFORMATION DOCUMENT
November 1992
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
Washington, DC 20460
-------
-------
PREFACE
The Environmental Protection Agency (EPA) is considering revisions to 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. 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 Eleanor Thornton at (202) 233-9773 or write to:
Director, Criteria and Standards Division
Office of Radiation and Indoor Air (6602J)
Environmental Protection Agency
401 M Street, SW
Washington, DC 20460
in
-------
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 Programs contributed to
the development and preparation of the BID.
Albert Coffi
Craig Conldin
Dale Hoffmeyer
Larry Gray
Chief, Air Standards and
Economics Branch
Health Physicist
Health Physicist
Environmental Scientist
Reviewer
Writer/Reviewer
Reviewer
Reviewer
An EPA contractor, S. Cohen & Associates, Inc., McLean, VA, provided signifi-
cant technical support in the preparation of the BID.
IV
-------
Contents
Preface iii
Disclaimer iv
List of Preparers iv
List of Tables . vii
List of Figures ix
Section
1. Introduction and Summary 1-1
1.1 History of Standards Development 1-1
1.2 Update Methodology 1-4
1.3 Purpose of the 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
2.2 The NRC's Regulatory Program under the Atomic Energy Act 2-4
2.3 Comparison of the NRC's Requirements with the NESHAP 2-7
2.4 NRC-Licensed Facility Program Analysis 2-9
3. Results of Designated Survey of NRC-Licensed Facilities 3-1
3.1 Uranium Fuel Cycle Facilities 3-2
3.2 Test and Research Reactors 3-13
3.3 Radiopharmaceutical and Radiolabeled Compound Manufacturers ... 3-17
3.4 Hospitals and Medical Research Facilities 3-20
3.5 Manufacturers of Sealed Sources 3-26
3.6 Testing of Depleted Uranium Munitions 3-29
3.7 Rare Earth and Thorium Processors (Source Material) 3-32
3.8 Commercial Low-Level Radioactive Waste Disposal and Incineration . 3-38
3.9 Summary of Results 3-41
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
4.3 Methods for Evaluating Data 4-6
4.4 Raw Results of the Survey , 4-7
4.5 Statistical Interpretation of the Results 4-9
5. Quality Control 5-1
-------
Contents (Continued)
Section
References [[[ R'l
Appendices
A - NRCs 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
-------
Tables
Number
1-1 Summary of estimated doses 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 3-12
3-6 Licensed test reactors in the United States as of August 1991 3-14
3-7 Effluent release rates (Ci/yr) for test and research reactors .. 3-15
3-8 DuPont Boston emission data 3-18
3-9 DuPont Billerica emission data 3-18
3-10 Mallincrodt emission data 3-19
3-11 Hospital and medical research facilities effluent release rates 3-23
3-12 Effluent release rates (Ci/yr) for sealed source manufacturers 3-28
3-13 Source term used for Aberdeen Proving Ground 3-31
3-14 Distances to receptors at Aberdeen Proving Ground 3-32
3-15 Rare earth processors' annual release rates 3-36
3-16 Summary of Designated Survey doses 3-42
VII
-------
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
vui
-------
Figures
Number Page
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 . 4-17
4-3 Frequency Distribution of Iodine Dose for 290 Facilities, with Fitted Models 4-18
4-4 Cumulative Dose Distribution 4-20
4-5 Extreme Tail of Sample Distribution 4-21
4-6 Cumulative Iodine Dose Distribution 4-22
4-7 Extreme Tail of Iodine Distribution 4-23
4-8 HLN-Probability Plot with Rho=0.14 4-24
4-9 HLN-Probability Plot with Rho=7.7 4-26
4-10 Cumulative Dose Distribution 4-27
4-11 Extreme Tail of Sample Distribution 4-28
4-12 Cumulative Iodine Dose Distribution 4-29
4-13 Extreme Tail of Iodine Distribution 4-30
IX
-------
-------
1. Introduction and Summary
1.1 HISTORY OF STANDARDS DEVELOPMENT
In 1977, Congress amended the Clean Air Act (CAA) to address emissions of
radioactive material. Before 1977, these emissions were either regulated under the
Atomic Energy Act or unregulated. Section 122 of the CAA required the Administrator
of the U.S. Environmental Protection Agency (EPA), after providing public notice and
an opportunity for public comment and public hearings, to determine whether emissions
of radioactive pollutants cause or contribute to air pollution that may reasonably be
expected to endanger public health. On December 27, 1979, the EPA published a notice
in the Federal Register (FR) listing radionuclides as hazardous air pollutants under
Section 112 of the CAA (44 FR 76738, December 27, 1979).
On April 6, 1983, the EPA published a notice in the Federal Register proposing
standards for four source categories: (1) Department of Energy (DOE) facilities, (2)
Nuclear Regulatory Commission (NRC) licensees, (3) underground uranium mines, and
(4) elemental phosphorus plants. The EPA determined that standards were not required
for the following source categories: (1) coal-fired boilers; (2) the phosphate industry; (3)
other mineral extraction industries; (4) uranium fuel cycle facilities, uranium mill tailings,
and high-level waste Management; and (5) low energy accelerators (48 FR 15077, April
6, 1983). To support these proposed standards and determinations, the EPA published a
draft report entitled "Background Information Document, Proposed Standards for
Radionuclides" (EPA83a).
On February 6, 1985, National Emission Standards for Hazardous Air Pollutants
(NESHAPs) were promulgated for emissions from DOE facilities, NRC-licensed and
non-DOE Federal facilities, and elemental phosphorus plants (50 FR 7280). Two
additional radionuclide NESHAPs, covering radon-222 emissions from underground
uranium mines and licensed uranium mill tailings, were promulgated on April 17, 1985
(50 FR 15385) and September 24, 1986 (51 FR 34056), respectively.
On July 28, 1987, the U.S. Court of Appeals for the D.C. Circuit remanded to the
Agency an emissions standard for vinyl chloride which had also been promulgated under
Section 112 of the CAA, In the case Natural Resources Defense Council, Inc. v. EPA,
1-1
-------
824 F.2d 1146 (D.C. Cir. 1987) (Vinyl Chloride), the Court concluded that the Agency
improperly considered cost and technological feasibility without first making a
determination based exclusively on risk to health.
In light of that decision, the EPA concluded that the standards for elemental
phosphorus plants, DOE facilities, NRC licensees, and underground uranium mines
should be repromulgated and on November 16, 1987, moved the D.C. Circuit Court for a
voluntary remand of the standards pursuant to pending litigation on the standards. On
December 8, 1987, the Court granted the EPA's motion for voluntary remand and
established a time schedule for the EPA to propose regulatory decisions for all
radionuclide source categories within 180 days and finalize them within 360 days. On
March 17, 1988, the Court granted a subsequent EPA motion and modified the order to
require proposed regulatory decisions by February 28, 1989, and final action by August
31, 1989.
On April 1, 1988, the EPA also requested a remand for its standard for licensed
uranium mill tailings. On August 3, 1988, the Court granted the EPA's motion and put
the uranium mill tailings NESHAP on the same schedule as the other radionuclide
NESHAPs.
On March 7, 1989, the EPA published a proposed NESHAP which described four
possible policy approaches for regulating emissions of radionuclides. Public hearings
were held on April 10, 11, 13, and 14, 1989. On July 14, 1989, the Court granted the
EPA's request for an extension until October 31, 1989, for final action.
On December 15, 1989, the EPA published final NESHAPs for the following
source categories: (1) DOE facilities, (2) NRC licensees and non-DOE Federal
facilities, (3) elemental phosphorus plants, (4) underground uranium mines, (5) operating
uranium mill tailings piles, (6) disposal of uranium mill tailings piles, (7) phosphogypsum
stacks, and (8) DOE faculties that release radon. The EPA decided that standards were
not required for coal-fired boilers, high-level nuclear waste disposal facilities, and surface
uranium mines. These NESHAPs are found in Part 61 of Chapter 40 of the Code of
Federal Regulations (CFR).
1-2
-------
Simultaneously with promulgating the NESHAPs, the 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 the
NRC and the National Institutes of Health (NIH) regarding the impacts of duplicative
regulations on licensees and the potential 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, the EPA continued to stay Subpart I in
its entirety.
While Subpart I was under reconsideration, the 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.
This section allows the Administrator of the EPA to decline to regulate NRC-
licensed facilities under Section 112 of the CAA if the Administrator determines, by
rulemaking and after consulting with the NRC, that the NRC's existing programs for
regulating these facilities ensure the protection of the public's health with an ample
margin of safety. Should the Administrator determine that an ample margin of safety
exists, the NESHAP for NRC licensees may be rescinded.
The 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), the EPA made an initial determination that the NRC's
1-3
-------
regulatory program limiting these emissions protects public health with an ample margin
of safely. Accordingly, on March 13, 1991, the EPA issued an Advanced Notice of
Proposed Rulemaking (ANPR) announcing the Agency's intention to enter into a
rulemaking to rescind Subpart I as it applies to nuclear power reactors (56 FR 37196).
At the same time, the EPA stayed subpart I for these facilities pending completion of the
recision rulemaking.
For all other categories of NRC licensees, the EPA concluded that it presently
lacks adequate information to characterize the facilities' emissions and thus embarked on
this information collection survey under Section 114. On April 15, 1991, the 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 the EPA is prepared to
make an initial determination under CAAA Section 112(d)(9) and conclude its
reconsideration (56 FR 18735, April 24, 1991).
Subpart I is presently in effect for non-DOE Federal facilities not licensed by the
NRC.
1.2 UPDATE METHODOLOGY
In previous evaluations (EPA83, EPA84, EPA89), the 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 faculties. Most of these large facilities
had emissions data providing a basis for reasonable estimates of doses and risks to public
health. Emissions from model faculties 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 the
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
1-4
-------
standards: first, determine an "acceptable" level of risk that considers only health factors,
and second, set a standard that provides an "ample margin of safety," in which cost,
feasibility, and other relevant factors in addition to health may be considered. The
Agency's methodology 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, the 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 (l.OE-04) is judged
to be presumptively acceptable.
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
MIR 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, the EPA seeks to assure that as many individuals as possible are at
an MIR of 1 in 1 million (l.OE-06) or less.
Using these criteria, the EPA found that the risk from all actual facilities
evaluated (both NRC-licensed 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 hi 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. The 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.
1-5
-------
For the large faculties previously evaluated by the 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, the EPA
has taken a statistical approach, based on a random sample of a subset of NRC and
Agreement State licensees. This sample was selected from the lists of facilities licensed
by the NRC and Agreement States that use radioactive materials. Licensees who only
use sealed sources were eligible for the sample but 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 THE BACKGROUND INFORMATION DOCUMENT
This BID provides background information from the Designated and Random
Surveys to assist the Administrator in detenmning 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 the 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 foUows:
Chapter 2 - A description of the EPA regulations that limit the effective
dose equivalent to members of the general public and the method for
detennining compliance with that dose limit. This chapter also summarizes
the organizational and administrative controls imposed by the 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.
1-6
-------
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 BID.
This BID also contains several appendices. Appendix A describes the
organization of the NRC and its regulations. Appendix B describes various NRC
Regulatory Guides pertinent to 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.
1-7
-------
3. The 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 the NRCs 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, Environmental Radiation Protection Standards for
Nuclear Power Operations.
4. The NRC has recently amended the requirements in 10 CFR Part 20. The
amendments, consistent with Federal guidance and the International Commission
on Radiological Protection (ICRP), 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) are based on 50 mrem/yr to account for multiple
pathways.
5. The Part 20 amendments also establish the requirement, previously just guidance,
to 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 The revisions to
Part 20, although still allowing a higher maximum permissible dose than the
NESHAP, is a stricter version of the regulation that has resulted heretofore in
actual licensee doses below the NESHAP limits. For this reason, the NRC
program should result in future emission levels no higher than current emissions
levels.
1 Per 10 CFR 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 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
SOCtoeconomic considerations, and the value of utilizing nuclear energy and licensed materials in the public
interest
1-8
-------
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 radioiodines. 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 cause doses in excess of the
NESHAP standard.
Estimated doses from the Designated Survey and the Random Survey are
summarized in Table 1-1.
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
mrem/Jr ede
2
7
6E-02
4
5
8
4
6E-04
2
7E-01
8
Maximum
Estimated
Iodine Dose
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 valup 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.
1-9
-------
-------
2. Description of Regulatory Programs
This chapter briefly summarizes the organizational and administrative controls
imposed by the EPA and the 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
The EPA regulations limit the estimated 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. The EPA does not license facilities; instead, each facility is
required to file an annual report showing that the dose from its emissions is less than the
limits. To minimize the burden on small users of radioisotopes, the EPA does not
require a report if the estimated dose is less than 10 percent of the standard.
The 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 the EPA for simple methods to assess compliance with the NESHAPs (NCRP86).
Commentary No. 3 was revised in January 1989. The EPA-approved methods for
demonstrating compliance with the NESHAPs are all based upon the January 1989
revision. In addition, the EPA allows the use of other methods for demonstrating
compliance, provided they have been approved by the Agency.
2-1
-------
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 the 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 fact to
the EPA. FacUities in compliance must file an annual report with the 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 NCRP 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
2-2
-------
about the size of the stack or vent. If the release rates are not measured,
the EPA has provided simple methods to estimate them. If desired, the
licensee may supply 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.
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 annual average frequencies for
each of 16 sectors are used along with the actual wind speed in that sector.
The licensee must supply distances to receptors in each of the 16 sectors.
COMPLY will determine 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
2-3
-------
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.
2.2 THE NRCS REGULATORY PROGRAM UNDER THE ATOMIC ENERGY
ACT
The regulatory programs established by the 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. The NRC implements its
programs either directly through licensing and inspection of facilities, or through the
Agreement State program, in which individual states perform the licensing and inspection
functions.
Facilities are regulated via the Code of Federal Regulations (CFR) and are
licensed by the 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 (enrichment 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.
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 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 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
-------
The NRCs regulations limiting routine radionuclide airborne emissions are
contained in 10 CFR Part 20, Standards for Protection Against Radiation, which applies
to all licensees, including those licensed by the Agreement States. For members of the
public, the basic dose limit was recently amended (effective January 1994) 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, the recent amendments to Part 20 require (effective January 1994) all
licensees to keep exposure as low as is reasonably achievable (ALARA).3 Formerly, this
ALARA requirement applied only to certain types of facilities, while the regulations
stated that other facilities "should" attempt to maintain exposures ALARA.
In addition to observing the 10 CFR Part 20 limits, licensees that are part of the
nuclear fuel cycle must comply with the EPA standard established in 40 CFR Part 190,
Environmental 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;
3 Per 10 CFR 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 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.
2-5
-------
identify the training and qualifications of the persons authorized to use the
material, and/or of the radiation safely 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.
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 (the 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 and the type of activity conducted.
2-6
-------
2.3 COMPARISON OF THE NRC'S REQUIREMENTS WITH THE NESHAP
The NESHAP established in 40 CFR 61, Subpart I, requires NRC-licensed facili-
ties 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 the 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 modification are equal to 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 ah 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.
2-7
-------
Table 2-1. Summary of regulatory requirements.
Regulatory
Activity
Licensing or
Approval
Dose Limit
Records
Reports
Inspections
Enforcement
NRC Requirements for
Large NRC Liccnsee$
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, collective
actions, fines, orders,
license revocation;
citizens may petition the
NRC to enforce, but if
the EDO does not agree,
no action is taken.
NRC Current Part 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 than 10 times
20.105 or 20.106.
Once every 1 to 7 years
depending on type of
license and activities
conducted.
Same as for large
facilities.
EPA's NESHAP
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
than 3 mrem/y due to
radioiodines.
Effluent monitoring data
or annual possession of
materials data used to
determine compliance.
Annual dose calculations
if greater than 10 percent
of limits.
Under development,
subject of possible MOU
between the EPA and the
NRC
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 than 2
mrem/hr. Licensees
subject to 40 CFR
190 must comply with
that, 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
excccdence of dose
limits must be
reported within 30
days.
No change.
No change.
2-8
-------
However, because licensees typically assess these concentrations in the stack, the
concentrations in open areas are lower due to dispersion.
This notwithstanding, the basic limits imposed by the 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 the NRC and the EPA to estimate dose. The
first is that the NRC's MPCs are based primarily on the inhalation pathway. By contrast,
the EPA concentrations consider doses received via the immersion, inhalation, ingestion,
and ground-surface pathways. The second difference is that the NRC's MPCs are based
on ICRP II recommendations, and the EPA's are based on ICRP 26 and 30. With the
revision to Part 20, the differences between the NRC and the EPA are lessened since
the new Part 20 uses ICRP 26 and 30 methodology.
2.4 NRC-IJCENSED 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 10 CFR Part 20.106, need not submit their calculations to
the 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 the NRC if concentrations exceed 10
times the allowable DACs.
Although the monitoring and inspection process for these facilities are not
rigorous, the 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
2-9
-------
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 II
of Appendix B to 10 CFR Part 2Q even in the event of minor operational errors.
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 NRCs 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.
The NRCs 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
faculties, manufacturing facilities, and medical institutions. However, because effluents
are not actually measured by stack instruments, the 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, the EPA decided to
conduct an analysis of the doses caused by these facilities to judge the adequacy of the
NRCs program.
2-10
-------
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), 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 tq 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.
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 the EPA's most
current 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);
3-J.
-------
and telephone 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 siimmarized 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 (yeUowcake) to uranium hexafluoride; the enrichment plants (owned
by the DOE but not regulated by the 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.
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.
3-2
-------
3.1.1.1 Previous Evaluations. The 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. The dose obtained for the model mill's tailings pile was 25.8
mrem/yr ede.1
3.1.1.2 Evaluations of Specific Facilities Made During the Reconsideration Period.
Since the dose reported in EPA89 for the model mill's tailings 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. The 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 NRC'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
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
-------
&
04
5
CO
en
?
CO
CO
I
P
t-i
d s>
*-i O
TO
0)
13
8
s
8
en
S3
o
Is
E-. -a
o
O
O
"S
o 01 g\
T-5 CO
Tt-
o
£
VO 01
» 0\ 04 VI
"
o~
ea
'S3
O
ol
O Hi
"o o u
ea 5 ^ O
*S Z -2 ^
Slf!
C Ci O 03'
< Dffi M
- -i
-1^-
4> PH cd -<
co ^£1! f? o
> c3 o t-
111
w o ffi
Tf CO OO
r-i r-5 VO
9-9-9"
8S8
o\ t^t t>
T-H i-H O\
co co cQ
^3 ^j ^3
o\
§
00
o\
^n
S":
88
II
3 ts
§ §
H-J »-»
11
t-t !-c
O O
I
>8
vo
\q 04.
888SS88S
i i i i i i i i
co
T-H VO i-5 CS CO CO
co"
8
O O
OO O4
VO
CO VO
o
!*>.
O
O
Q
o
Western Nucle
a S
a s
o >^
Mt^;
J3^ S
S s3
S3?
cl
s
bJ
a
IS
K
CQ
^
13
0
g
(O di a> r ^
rtc!u5C3iflcold!< >r
«
s
o
PH
11
£3 *j O
2 1 "s
u "o" « ^
D 6; . "* .23
rH
-------
farms. These assumptions are consistent with those made in the Random Survey portion
of the study.
Source Term Determination
The EPA derived airborne source terms for exposed tailings using site
meteorology taken from EPA89 and the methodology suggested by the 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 from 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
effective dose equivalent (ede) calculated using COMPLY is 2 mrem/yr. This dose is
primarily from the inhalation and ingestion 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, as shown in Table 3-16.
3-5
-------
3.1.2 Uranium Conversion Facilities
A 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.
3.1.2.1 Previous Evaluations. Currently, two commercial uranium hexafluoride (UF6)
production faculties 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, m 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 hi 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 the 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.
3-6
-------
Table 3-2. Atmospheric radioactive emissions assumed for reference dry and wet
process uranium conversion facilities
Facility
Radionudide
Emissions
(Ci/y)
Solubility Class (%)*
D
W
Reference
Allied Corp.
Metropolis, IL
U-Natural2
Th-2302
Ra-2262
0.10000
0.00050
0.00001
56
0
0
30
0
100
14
100
0
NRC84
Sequoyah Fuels
Sequoyah, OK
U-Natural3
Th-2303
Ra-2263
0.050
0.005
0.005
65
0
0
5
0
100
30
100
0
NRC85b
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.
3.
Particle size 3.4 um.
Particle size (urn)
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: 1980-19841
93
9.7
5.5
6.5
13.5
55.3
Data taken from NUREG-1157 (NRC85b).
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 faculties 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.
3-7
-------
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 "stocker 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 stocker 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.
3.1.3 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.
3.13.1 Previous Evaluations.
Non-Light Water Reactor HLWRI Fuel Fabrication Facilities. None of the
faculties 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 hexafluoride 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.
3-8
-------
i
,g»^
3 '**"''
§ h 'd
Produc
Fi
o o "
8°ftf
£ sS
OO
Ope
CO
(U
'i
H
(D
t)
o
o
§
Complete fuel
assembh'es
-
LEU
(UF^ to
Fabrica
Reco
LWR
i*
,
Q
LEU Conversion
S"
1
1
a
.2
"1
<3
J
o
s
«
O
3
; si
aw
s
(UF6 to UO2) &
Scrap Recovery
O
bO
j3
8
"5b
W
8
^
&
LEU Conversion
«"
I
.f
i
"3
§
o
f
CO
s §
y
&a
u
z
_y
0
W
Recovery: Commercial
LWR Fuel
T3
«e«
J3^
jaS
&a
a u
,«? s
O CQ
0
^
b
p
.1^
i2<75
o _
j N \C
w ^5
jio
.s"
5
CJ on
u
CO
3
5
}l
^ w
»*
Fabrication & Scrap
Recovery; Commercia
LWR Fuel
o
3
.a
1
i
s
o
.53
I
I
1
I
I
iS
3-9
-------
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 TLWRt 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 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, the 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. All other 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 the 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 start .with uranium hexafluoride, while the
other two facilities begin the fuel fabrication process with UO2.
3-10
-------
Table 3-4. Light water reactor commercial fuel fabrication facilities reported annual
uranium effluent releases for 1983 through 1987 in
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<2>
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
L5E+03
1.2E+03
1.1E+03
l.OE+03
U-235
2.1E-01
2.5E-01
2.1E-01
2.5E-01
1.7E-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
53E+01
1.2E+02
7.2E+01
53E+01
5.6E+01
U-236
2.1E-02
23E-02
2.1E-02
2.6E-02
1.7E-02
NA
NA
NA
NA
NA
NA
NA
NA
NA
NA
13E+02
5.7E+00
5.7E+00
1.6E+01
2.0E+00
NR<4>
NR
NR
NR
NR
U-238
1.1E+00
13E+00
1.1E+00
13E+00
9.1E-01
NA
NA
NA
NA
NA
NA
NA
NA
NA
NA
4.5E+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
73E+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
73E+01
6.7E+02
2.8E+02
4.6E+02
6.0E+02
5.9E+02
1.6E+03
23E+02<3>
L5E+03
1.9E+03
1.6E+03
1.5E+03
1.4E+03
1. Taken from semi-annual licensee environmental monitoring reports submitted to the
NRC.
2. Not available; only total curies of uranium released reported to the 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.
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.
3-11
-------
Table 3-5. Atmospheric radioactive emissions assumptions for
reference fuel fabrication facility.
RSdionuclide
U-234
U-235
U-236
U-238
Emissions (Ci/yr)
1.2E-03
6.7E-05
1.6E-05
3.0E-04
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 the NRC by the General
Electric facility at Wilmington, North Carolina. The effluent release height used in this
analysis is 10 m (EPA89).
Meteorologic, Demographic, afld Agricultural Information
The cMmatological data used originally in EPA89 are based on measurements
taken at the U.S. Weather Bureau Station at Columbia 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.
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
3-12
-------
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, the 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 the NRC in the United States (NRC88a).
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).
3-13
-------
Table 3-6. Licensed test reactors in the United States as of August 1991.1
NRC
Docket
No.
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
NBS
GEEVESR
Exp. Superheat
B&W BAWTR
SEFOR Sodium Cooled
Thermal
Power (Mw)
60
60
50
28
' 10
17
6
20
4
Location
Waltz Mill, PA
Sandusky, OH
Alameda County, CA
Saxton, PA
Gaithersburg, MD
Alameda County, CA
Lynchburg, VA
Strickler, 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.
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 Standards2). 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 hi lesser
amounts. The emissions result in a maximum estimated dose of 0.7 mrem/yr ede
(EPA89).
2 NBS is now known as the National Institute of Science and Technology.
3-14
-------
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 [NIST] reactors). These three
remaining reactors were included in the Designated Survey. As part of the reevaluation,
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.
Faei%
University of Missouri
National Institute of Standards & Technology
Massachusetts Institute of Technology
Radionuclide
H-s
1.6E+01
1.6E+02
-
Ar-4t
2.5E+03
4.7E+02
4.2E+03
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
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 MIT reactors are for Columbia, Missouri; Fort Meade, Maryland; and Boston,
3-15
-------
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 MTT, 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.
3.2.3 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.
' -3-16
-------
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). Mallincrodt's Maryland Heights,
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 faculty 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.
3-17
-------
Table 3-8. Dupont Boston emission data.1
Nudide
H-3
C-14
C-14
S-35
% ,vt f « * *****
Release Rate (Ci/yr)
1987
97.7
4.7
3.8
038
- 1989
18m
12.9
1.9
2.8
0.2
24m
91.23
4.9
10.1
0.3
1. Data obtained from SCA91.
For Dupont Billerica, both 1987 and 1989 releases of iodine-125 were known.
These include estimates of the releases from a waste storage warehouse. These values
are based on DuPont's engineering estimates of the potential releases from the
warehouse based on ambient air monitoring results and estimated air turnover rates.
Available release data for calendar years 1987 and 1989 are presented in Table 3-9. The
emission data for 1987 were used for this evaluation.
Table 3-9. Dupont Billerica emission data.
f .
Nudide "
Xe-133
P-32
S-35
1-125
1-131
Kr-85
Release Rajte (Ci/yr)
1987
2.84
1.6E-02
1.6E-02
2.0E-02
2.5E-02
9.5E-01
1989
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 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.
3-18
-------
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.
Nudide
1-131
1-125
1-123
Tc-99m
Mo-99
In-111
Ga-67
Release Kate (Ci/yt)
33-m Stack
1J5E-02
-
1.6E-03
-
-
.
-
19-a* Stack
2.2E-01
7.0E-04
L5E-03
7.7E-02
63E-03
l.OE-03
6.0E-04
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
3-19
-------
#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 m 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.
3.4 HOSPITALS AND MEDICAL RESEARCH FACILITIES
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), the EPA focused on large hospitals and medical research
faculties. 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
3,20
-------
assessment code AIRDOS-EPA When the 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, the 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 rulemaking, the 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 "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
uncertainly 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 the 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 NIH and other qpmmentators on the
NESHAP focused on- the stringency of the 3 mrem/yr ede limit for doses from radioio-
3-21
-------
dines, the EPA again attempted to identify medical facilities using large quantities of
radioiodines. Beginning with information supplied by the medical facilities, the 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 the EPA was requested. Information
on quantities of radioactive materials used, effluent concentrations, effluent controls
employed, and locations of nearby individuals was obtained from SCA91 for each facility.
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 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.
3-22
-------
Table 3-11. Hospital and medical research facilities effluent release rates.
Facility Name/Location
Johns Hopkins
Baltimore, MD
University of Wisconsin
Madison, WI
Release Point/ Stack
Height
WBS Building
51m
P-B Building
13m
T Building (incinerator)
51m
Incinerator
10m
Nuclide
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
In-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-111
Sn-113
Ce-141
Se-75
Ci/yr
5.0E-fOO
5.0E-01
2.1E-10
L7E-03
2.1E-05
2.5E-05
15.6
1.4E-02
9.5E-04
4.0E-03
73E-03
8.5E-02
2.0E-03
13E-01
7.2E-02
63E-02
3.1E-02
7.7E-02
3.5E-02
7.9E-02
4.7E-02
3.9E-01
2.0E-02
33E-02
8.5E-04
2.0E-03
1.7E-03
1.4E-03
l.OE-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
-------
Table 3-11. (Continued)
UCLA Hospital
Los Angeles, CA
UC San Francisco
San Francisco, CA
UC Irvine
Irvine, CA
NIH
Bethesda, MD
Hospital
5m
MS Building
56m
Nuclear Med. Building
5m
NIH Complex
42m
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
2.2E-03
7.0E-03
l.OE-04
32.8
6.7E-03
3.2E-03
5.0E-04
1.3E-03
1.5E-03
3.0E-03
1.6E-04
7.0E-01
6.2E+00
6.5E-03
2.5E-03
2.0E-03
l.OE-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
13E-02
3.4E-04
2.2E-02
1.9E-02
2.1E-02
Meteorological. Demographic, and Agricultural Data
Johns Hopkins (SCA91): 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)
3-24
-------
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 (SCA91): 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.
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 0.3 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.
3-25
-------
NIH6: 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 NIH. The total ede calculated for
NIH 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.
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 faculties are located in industrial areas. They rely
heavily on engineered safeguards to prevent releases of radionuclides.
6 Personal correspondence between R. Zoon (NIH) and A. Colli (EPA), November 1989.
3-26
-------
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 gone out of production. The remaining two were re-evaluated using updated,
detailed demographic data. Calculations supporting 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 site-specific, updated, close-in demography information
supplied by the licensee, and the findings are incorporated in this study.
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 is equipped with a continuous vent
monitoring system, which was modified recently. The sampling is isokinetic, 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 IE-12 //Ci/ml, approximately 0.3 percent of the MPC for insoluble forms
of cobalt-60. All measurements with this new system show activity below the MDL. The
3-27
-------
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 used in the analysis are also shown in Table 3-12.
Table 3-12. Effluent release rates (Ci/yr) for sealed source manufacturers.
Radionuclide
H-3
Co-60
Ni-63
Po-210
Am-241
Neutron Products
_
1.2E-05
-
-
-
NRD, Inc.
3.4E+02
-
8.0E-06
1.4E-04
6.1E-05
Safety Light Corp.
2.2E+03
-
-
-
-
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.
3-28
-------
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 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 for 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.
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 mrem/yr. Inhalation is the primary pathway of exposure.
3.6 TESTING OF DEPLETED URANIUM MUNITIONS
The processing of natural uranium to obtain uranium enriched hi the uranium-235
isotope results in abundant tails referred to as depleted uranium. The NRC licenses
ownership, possession, and use 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 release airborne radioactive material.
3-29
-------
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.
The Aberdeen Proving Ground conducts the greatest number of test firings.
Because it is also very close to many residences, the 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 respirable range. However, to remove any uncertainty, this source category was
evaluated in this study.
3-30
-------
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
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.
Table 3-13. Source term used for Aberdeen Proving Ground.
Operation
Range 9
Range 14
Range 14A
Range HOE
Abrasive Blaster
BTD Enclosure
Superbox
Cut Box
U-23S Release Rate, Ci/jrl l
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.
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.
3-31
-------
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, m
5000 (R)1
7000 (R)
7000 (R)
200 (R)
1200 (B)
1100 (B)
1000 (B)
1QOO(B)
1. R indicates residence; B indicates business.
All farms are located at greater than 2,000 m; however, all vegetables were
assumed to be grown at the home of the closest individuals.
3.6.3 Results of the Designated Survey for Testing of Depleted Uranium Munitions
The dose received by the maximally exposed individual in proximity to the
Aberdeen Proving Grounds is 6E-04 mrem/yr-ede.
3.7 RARE EARTH AND THORIUM PROCESSORS (SOURCE MATERIAL)
Approximately 10 facilities 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
ShieldaUoy-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
3-32
-------
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
The EPA conducted a screening assessment in 1983 and concluded that rare earth
and thorium processors did not pose a public health risk (EPA83). However, the EPA
decided to reduce the uncertainty associated with the 1983 evaluation.
The 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 m),
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).
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 columbium. 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.
3-33
-------
Raw ores are ground into a flour-like consistency and then transferred into
digester tanks which selectively dissolve the tantalum and columbium. The unwanted
uranium and -thorium react with the acid to form insoluble uranium and thorium
fluorides. Particles less than 10 ^m 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 trackload 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. The 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
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.
3-34
-------
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.
Shieldalloy-Newfield: This active facility is located in a rural area. Shieldalloy
manufactures a variety of specialty ferro-alloys, using the raw material ferro-columbium
(Fe-Cb). 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 m3/min baghouse dust collectors before its release
to the environment. The maximum amount processed per day would be about 400 ^Ci
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.
In July 1988, ORAU performed a radiological survey of the Newfield site. It
found that "there hasn't been adequate segregation and control of potentially radioactive
materials at this site in the past" (ORAU88).
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.
3-35
-------
The 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
Each operating rare earth processor provided release rate information. Molycorp
and Shieldalloy supplied process source term data in response to the EPA's survey.
Comparable information was not available for Cabot Corporation. Instead, site
meteorology was used in conjunction with the methods hi Regulatory Guide 3.59 to
derive the airborne source term for sludge that is stored in open-air vented mausoleums.
Table 3-15 presents the source terms used in this study.
Table 3-15. Rare earth processors' annual release rates.
Facility
Cabot Corp.
Molycorp, Inc.
Shieldalloy
Release Point
Mausoleums
Tank Room
Waste Treatment
Moly Building
Department 111
Stack
Height (m)
1
10
5
2
12.2
Nataral
Thorium1
(0/yr)
7.6E-07
1.5E-06
7.0E-05
2.0E-07
3.0E-043
Natural
Uranium?
-------
Demographic data obtained for Cabot Corporation indicated that the nearest
resident is approximately 270 m from the mausoleums used to store 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.
3.7.3 Results of the Designated 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.7 The maximum
ede is calculated for Shieldalloy Metallurgical Corporation; the dose received by this
individual is 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.
7 The value of 0.3 is based upon 10 hours per day, 5 days per week, 52 weeks per year
(10x5x52/8760=0.3).
3-37
-------
3.8 COMMERCIAL LOW-LEVEL RADIOACTIVE WASTE DISPOSAL AND
INCINERATION
Virtually every user 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. LLW contains only low
concentrations of radioactive material.
Waste Brokers
Waste receivers and shippers (sometimes called "waste brokers") are primarily'
collection and shipping agents for faculties 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.
3-38
-------
Disposal Facilities
Some radiomiclides may also be emitted from LLW. disposal sites. Currently,
only three sites (Barnwell, South Carolina; Beatty, Nevada; and Richland, Washington)
are operating. As required by the Low-Level Radioactive Waste Policy Act of 1982, any
state that wishes to dispose of its LLW should either be part of an interstate compact or
be designing its own facility in accordance with 10 CFR 61.
LLW disposal sites do not accept special nuclear materials, transuranics, and
spent reactor fuels. 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, the EPA's evaluation of incinerators was limited to those which were
part of hospital and medical research faculties 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.
3-39
-------
3.8.2 Evaluations of Specific Facilities Made During the Reconsideration Period
The 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, Barawell, and Beatty), the EPA
confirmed its prior analyses (EPA79, EPA84) through conversations with state radiation
control officers.8
For compact sites, the 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, the EPA investigated the SEG incinerator located in Knoxville,
Tennessee. Quarterly data include radionuclide 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 one of the operating LLW disposal sites,
reported in Table 3-16, indicates that releases above background have not been detected.
As a result, no COMPLY calculations were made for that site.
For the proposed compact 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).
8 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.
3-40
-------
For incineration operations, the dose to the maximally exposed individual is
established at less than 7E-03 mrem/yr ede, with 3E-04 mrem/yr 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.
3-41
-------
Table 3-16. Summary of Designated Survey doses.
Type of Facility ,
Uranium Tailings
Company^ Facility Name
American Nuclear Corp
Federal American Mill
Riverton, WY
Anaconda
Anaconda Mill
Bluewater, NM
Adas
MoabMill
Moab, UT
Exxon
Highland Mill
Douglas, WY
Homestake
Homestake Mill
Grants, NM
Kerr McGee
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
LaSal,UT
Umetco Minerals
Gas Hills Mill
Riverton, WY
Umetco Minerals
White Mesa Mill
Blanding,UT
Umetco Minerals
Uravan Mill
Uravan, CO
Total
ede* (mrem/yr)
3E-01
6E-01
1E-01
1E-01
2E-01
4E-02
4E-02
6E-02
2E-01
2E+00
4E-02
3E-01
2E-01
8E-03
loditte %
ede
(mrem/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
3-42
-------
Table 3-16. (Continued)
Type of Facility
UF6 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, EL
- 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 Billerica
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 (mrem/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
Iodine
ede
fmrtenj/vrl
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
3-43
-------
Table 3-16. (Continued)
Type of Facility
Manufacturers of Sealed
Sources
Testing of Depleted
Uranium Munitions
Rare Earth & Thorium
Processors
Commercial Low-Level
Radioactive
Waste Disposal &
Incineration
Company & Facility Name
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
Hanford Site
Richland, WA
Total
ede1 (mrem/yr)
3E-022
4E+00
5E-02
7E-03
6E-04
6E-012
1E-02
2E+00
7E-01
7E-01
<7E-03
Iodine
ede
ftnrem/vrt
< 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 03 for people in businesses or offices.
3-44
-------
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
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 the 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 the EPA, and presumed to represent the "worst cases"
in terms of MIR, actually do represent the upper-bound of the doses caused by 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 the 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 NRC-licensed
facilities.
In making its evaluation, the 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
-------
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 faculties provided to the EPA by the NRC and Agreement States.
Facilities 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 the NRC.
4.2.2 Data Requirements
la 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 from
both databases includes licensed facilities using both sealed and unsealed sources.
The database distinguished between those facilities licensed directly by the 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 faculties) due to the differing sources of information on licensees in these
strata. Initial sampling of these strata permitted estimation of the relative proportion of
4-2
-------
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 Sites3
(item 5 as percent of item 4)
NRC
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%
Total
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.
unsealed source sites in each 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
-------
Some or all of the following information was obtained through the 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.
4-4
-------
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).
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 residents in the release building, 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 not a resident, but was in an office, school, or business,
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.
4-5
-------
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 hi the airborne emissions
and by the atmospheric dispersion and deposition processes.
Computer codes are commonly used to model dispersion and deposition processes
that determine human exposure. The 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 this study were run using the most realistic level (Level 4).
4-6
-------
4.4 RAW RESULTS OF THE SURVEY
4.4.1 Results
The 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 the 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.
Table 4-2. Number of facilities having doses in various ranges.
Maximum Individual Effective
£tos& Equivalent (ntrera/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
IE-OS to 1E-07
1E-07 to 1E-06
1E-06 to IE-OS
IE-OS to 1E-04
1E-04 to IE-OS
IE-OS to 1E-02
1E-02 to 1E-01
1E-01 to 1.0
1.0 to 10
> 10
Total
KromAlI
Naciides
0
0
3
2
4
7
16
29
56
79
82
66
16
7
0
367
From
ItadioSodHte
1
2
7
24
18
23
30
47
33
36
32
28
9
0
0
290
4-7
-------
4.4.2 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.
4.4.3 Population Dose Estimates
As discussed in Section 1.2, the multi-factor approach adopted by the 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 (240 million, i.e., 100 percent of the U.S.
population) 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 faculties 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.
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 faculties evaluated may have emissions associated with risks in
excess of 1E-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,
0.3 to 3.0, and 3.0 to 10 mrem/yr. The numbers of people in the annuli defined by these
distances were estimated using Census Bureau data (CB88). The number of people at
the various levels of exposure is shown hi Table 4-3. The analysis assumes that an
exposure of 3 mrem/yr is the equivalent of a lifetime cancer risk of 1E-04.
4-8
-------
Table 4-3. Population dose estimates.
Dose (mrem/yr)
0.03 to 03
0.3 to 3
3 to 10
Approximate Risk
1E-06 to IE-OS
IE-OS to 1E-04
1E-04 to 1E-03
Estimated Number
of People In
Sample of 367
2,000
89
1
Estimated Number
of People for 6,200
Facilities
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).
4.4.4 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). The EPA adopted this system because it is simple,
related to risk, and widely accepted by leading national and international advisory
bodies.
The 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 the EPA are not
strictly proportional to the ede derived using the ICRP quality factors and organ
weighting factors. While the risk methodology underlying the ICRP ede differs from that
used by the EPA in the past, the EPA believes that 3 mrem/yr ede is approximately
equal to a lifetime individual risk of 1 in 10,000.
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
4-9
-------
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 uncertainly 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.
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 is 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
4-10
-------
Table 4-4. Estimated distribution of maximum individual doses.
A. Selected characteristics of the random sample
Sample Characteristic
Minimum Dose
Geometric Mean
Median Dose
Arithmetic Mean
Maximum Dose
Sample Size
Value
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 perceatiles of the estimated dose distribution
Sample or Population
Percentile
10
20
30
40
50
60
70
80
90
95
99.0
99.7
Boss
(mrem/yr}
1.6E-06
1.7E-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
Estimated Number of Facilities at or Exceeding
Has Dose:
In Sample
331
294
257
220
184
147
no
73
36
18
3
1
la Population
5,538
4,922
4^07
3,692
3,077
2,461
1,846
1,231
615
308
62
18
4-11
-------
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
13E-02
6.7E-01
290
B. Selected percentues of the estimated: radioiodute dose distribution
Sample or Population
Percentile
10
20
30
40
50
60
70
80
90
95
99.0
99.7
" , Dose
(rarem/yr)
8.0E-10
3.9E-08
2.4E-07
1.8E-06
8.1E-06
4.2E-05
3.4E-04
2.4E-03
2.0E-02
6.0E-02
3.9E-01
6.7E-01
Estimated Komber of Facilities at or Exceeding
" " This Dose:
la Sample
261
232
203
174
145
116
87
58
29
14
3
1
In Population
4,376
3,890
3,403
2,917
2,431
1,945
1,459
972
486
243
49
19
4-12
-------
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 than 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:
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, at 1.3E-02 mrem/yr.
4-13
-------
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 faculties 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 faculties in the population exceeding this iodine
dose level.
These point estimates of the iodine dose distribution have an associated
uncertainly 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
4-14
-------
-------
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 lognormal curve
underestimates the sample distribution in the shoulder region and overestimates the
sample distribution in the upper tail.
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, workers 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 looks like a normal distribution above rho X = 1 and like 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 mrem/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.
4-16
-------
CO
ULJ
o
£
Q
111
_J
Q_
^
<
CO
111
13
O
111
DC
U_
CM
CM
E
CD
i_
E
c
LU
CO
O
Q
O
o
I
CM
I
-------
T3
O
4-18
-------
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 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 + In(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.
4-19
-------
CtL
LU
CO
UJ
Q.
CO
"«*
a
Q
Q
2
I
o>
CO
CO
in
-------
og -
- o
- to
CM
-
en
Q
(U
f(
&.
CO
<4-<
o
»~(
S
-------
liJ
1
cc.
LJ
CO
-------
+ +
CD
co r-
f 4-
CO
CD
(£>
O
CM
4-
XQ
LJ
+
+
CN|
o*
-------
UJ
00
*
-------
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 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/yr 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
4-25
-------
UJ
(O
O
O
*
O
01
o
T
OLJ
O '
K
01 CO
o
CM
O
CO
I
J
*o
o
4>
O
s
3
I
o
s
£>
r-<
i
2
>
3ivos
4-26
-------
LJ
Q
O
Ldm
eoii
§1
O
O
O
s
ffi
1
a
i
§
^4
4-*
I
I
ea
1
£
JO lN30W3d
4-27
-------
co fc £ £
CO CM ;: 2
_ CM
- O
- CO
£<
£k
to 0^5
f^\ *
1 co
xa
CJ
q
in
q
co
q
CM
s?
o
cJ?
q
CD
a
o
-^-t
*-*
M
.23
Q
-------
Ul
Q
O
(D
o
o.
3
ffi
13
G
.tj
4->
«-H
O
t-i
-*-»
CO
(U
CO
O
Q
.1
T3
O
»(
-------
£<=»
e
§
'B
C/3
.1
-------
Table 4-6. Estimated percentage and number of facilities exceeding specified
dose using the lognormal and hybrid-lognormal models.
A. Model-based estimates for dose from, all nudities
Percent of Facilities Exceeding 10
mrem/yr
Lognormal
0.54%
BLN
0.22%
Number of Facilities Exceeding 10
mrem/yjr
(Out of 6,153 Facilities)
Lognormal
33
HLN
14
EL Model-based estimates for dose from radioiodine
Percent of Facilities Exceeding 3
mreni/yr
Logaotaaal1
1.26%
ULN
< 0.0000001%
Number of FacUifles Exceeding 3
- mrem/yr
(Out oT 4353 Facilities}
Ixjgnojttoal1
61
BUST
< 1
1. Estimated values from the lognormal distribution are overstated due to
over-estimation by the model of the upper tail of the sample
radioiodine distribution.
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.
4-31
-------
Short of obtaining information and determining the doses to the public from every
one of the roughly 6,000 facilities licensed by the 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.
4-32
-------
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, the EPA prepared a sample questionnaire to assess the licensees' ability to
interpret the 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, the 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
-------
-------
References9
CB88 U.S. Bureau of the Census, "County and City Data Book 1988," U.S.
Government Printing Office, 1988.
Ce81 Centaur Associates, Inc., "Economic Study of the Radionuclides Industry,"
prepared for the U.S. Nuclear Regulatory Commission, NUREG/CR-4048,
Washington, B.C., 1981.
Co83 Corbit, C.D., et al., "Background Information on Sources of Low-Level
Radionuclide Emissions to Air," Pacific Northwest Laboratory, PNL-4670,
Richland, WA, 1983.
DA92 Letter from M.F. Flannery, Department of the U.S. Army, Directorate of
Safety, Health and Environment, Aberdeen Proving Ground, to D.
Hoffmeyer, U.S. EPA, February 27, 1992.
EPA73a U.S. Environmental Protection Agency, "Environmental Analysis of the
Uranium Fuel Cycle - Part I - Fuel Supply," EPA 520/9-73-003C, Office of
Radiation Programs, Washington, D.C., 1973.
EPA73b U.S. Environmental Protection Agency, "Environmental Analysis of the
Uranium Fuel Cycle - Part H - Nuclear Power Reactors," EPA 520/9-73-
003C, Office of Radiation Programs, Washington, D.C., 1973.
EPA78 U.S. Environmental Protection Agency, "A Radiological Emissions Study at
a Fuel Fabrication Facility," EPA 520/5-77-004, Office of Radiation
Programs, Washington, D.C., 1978.
EPA79 U.S. Environmental Protection Agency, "Radiological Impact Caused by
Emissions of Radionuclides into Air in the United States," EPA 520/7-79-
006, Washington, D.C., 1979.
EPA82 U.S. Environmental Protection Agency, "Final Environmental Impact
Statement for Remedial Action Standards for Inactive Uranium Processing
Sites," EPA 520/4-82-013-1, Washington, D.C., October 1982.
EPA83 U.S. Environmental Protection Agency, "Final Environmental Impact
Statement for Standards for the Control of Byproduct Materials from
Uranium Ore Processing," EPA 520/1-83-008-1, Washington, D.C.,
September 1983.
References are also listed at the end of each appendix.
R-l
-------
EPA83a U.S. Environmental Protection Agency, "Background Information
Document, Proposed Standards for Radionuclides," EPA 520/1-83-001,
Office of Radiation Programs, U.S. EPA, Washington, D.C., March 1983.
EPA84 U.S. Environmental Protection Agency, "Radionuclides Background
Information Document for Final Rules," EPA 520/1-84-022, Office of
Radiation Programs, Washington, D.C., October 1984.
EPA84a U.S. Environmental Protection Agency, "Occupational Exposure to Ionizing
Radiation in the United States," EPA 520/1-84-005, Office of Radiation
Programs, Washington, D.C., 1984
EPA86 U.S. Environmental Protection Agency, "Final Rule for Radon-222
Emissions from Licensed Uranium Mill Tailings," Background Information
Document, EPA 520/1-86-009, Washington, D.C., August 1986.
EPA89 U.S. Environmental Protection Agency, "Environmental Impact Statement,
NESHAPs for Radionuclides, Background Information Document -
Volume 2," EPA 520/1-89-006-1, Washington, D.C., September 1989.
EPA89a U.S. Environmental Protection Agency, "A Guide for Determining
Compliance with the Clean Air Act Standards for Radionuclide Emissions
from NRC-Licensed and Non-DOE Federal Facilities," EPA 520/1-89-002,
October 1989.
EPA89b U.S. Environmental Protection Agency, "Procedures Approved for
Demonstrating Compliance with 40 CFR 61, Subpart I," EPA 520/1-89-001,
October 1989.
EPA91 U.S. Environmental Protection Agency, "Background Information
Document to Support NESHAPs Rulemaking on Nuclear Power Reactors,"
EPA 520/1-91-019, Washington, D.C., August 1991.
EPA92 U.S. Environmental Protection Agency, 'Technical Support for Amending
40 CFR 192," Draft, May 1992.
G178 Glick, N., "Breaking Records and Breaking Boards," American
Mathematical Monthly. Mathematical Association of America, Washington,
D.C., Vol. 85, No. 1, pp. 2-26, 1978.
Ku81 Kumazawa, S. and Numakunai, T., "A New Theoretical Analysis of
Occupational Dose Distributions Indicating the Effect of Dose Limits,"
Health Physics. Vol. 41, No. 3, pp. 465-475, 1981.
R-2
-------
MIT87
Mu91
NCRP86
Ne82
NOAA90
NRC82
NRC84
NRC84a
NRC85a
NRC85b
NRC87
NRC88
M.I.T. Research Reactor Staff, "Annual Report to United States Nuclear
Regulatory Commission for the Period July 1, 1986-June 30, 1987,"
Cambridge, MA, August 29, 1987.
Mullins, T.J., DuPont/Merck, letter to S. Beal, December 1991.
National Council on Radiation Protection and Measurements, NCRP
Commentary No. 3, "Screening Techniques for Determining Compliance
with Environmental Standards: Releases of Radionuclides to the
Atmosphere," 1986.
Nelson, D., Kumazawa, S., Tokai-mura, and Richardson, A.C.B., "Hybrid
Lognormal Analysis of U.S. Occupational Exposure in 1975 and 1980,"
presented at the 27th Annual Meeting of the Health Physics Society, Las
Vegas, NV (P/71), 1982.
National Oceanic and Atmospheric Administration, "Stability Array
Tabulation (STAR)," TD-9773, August 1990.
U.S. Nuclear Regulatory Commission, "Technology, Safety, and Costs of
Decommissioning Reference Nuclear Research and Test Reactors: Main
Report," NUREG/CR-1756-Vl, PNL, March 1982.
U.S. Nuclear Regulatory Commission, "Environmental Impact Appraisal for
the Renewal of Source Material License No. SUB-526," NUREG-1071,
May 1984.
U.S. Nuclear Regulatory Commission, "Incineration by Material Licensees,"
Policy and Guidance Directive, FC 84-21, December 1984.
U.S. Nuclear Regulatory Commission, "Environmental Assessment for
Renewal of Special Nuclear Material License No. SNM-1107," NUREG-
1118, May 1985.
U.S. Nuclear Regulatory Commission, "Environmental Assessment for
Renewal of Special Nuclear Materials License No. SUB 1010," Docket No.
40-8027, Sequoyah Fuels Corporation, NUREG-1157, August 1985.
U.S. Nuclear Regulatory Commission, "Methods for Estimating Radioactive
and Toxic Airborne Source Terms for Uranium Milling Operations,"
Regulatory Guide 3.59, March 1987.
U.S. Nuclear Regulatory Commission, "Environmental Impact Appraisal for
Renewal of License No. SMB-920," NUREG-1027, November 1988.
R-3
-------
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.
R-4
-------
APPENDIX A
NRC's ORGANIZATION, REGULATIONS, AND CONTROLS
This appendix explains the origins and need for the NRC and its
predecessor, the Atomic Energy Commission, dating back to the Atomic
Energy Act. It describes how the NRC's organization promotes the
discharge of its responsibilities and its ability to fulfill its legislative charter.
Regulations and effluent controls for NRC-licensed faculties other than
nuclear power reactors are described.
A-l
-------
CONTENTS
A.1 Organization and Rsponsibilities of NRC A-3
A. 1.1 Basic Functions A-3
A.1.2 Organization A-4
A.l.2.1 The Commission A-4
A.l.2.2 NRC Offices A-4
A.l.2.3 State Programs A-7
A.2 Controls Applicable to Licensees - General A-9
A.2.1 Establishing Airborne Emission Controls A-9
A.2.1.1 Rulemaking and Regulatory Guides A-10
A.2.1.2 Generic Letters, Bulletins and Information Notices A-ll
A.2.1.3 NRC Reports A-ll
A.2.2 Licensing Program A-12
A.2.2.1 Part 30 Licenses A-12
A.2.2.2 Part 40 Licenses A-14
A.2.2.3 Part 50 (Type 104) Licenses A-16
A.2.2.4 Part 70 Licenses A-16
A.2.3 Airborne Emissions Monitoring A-18
A.2.3.1 Part 30 Licenses A-18
A.2.3.2 Part 40 Licenses A-20
A.2.3.3 Part 50 (Type 104) Licenses: Changes, Tests, and
Experiments (10 CFR 50.59) A-21
A.2.3.4 Part 70 Licenses -. A-22
A.2.4 Inspection Programs A-22
A.2.4.1 Part 30 Licenses A-22
A.2.4.2 Part 40 Licenses A-23
A.2.4.3 Part 50 (Type 104) Licenses A-23
A.2.4.4 Part 70 Licenses A-23
A.2.5 Enforcement Programs A-23
A.3 Controls Applicable to Airborne Emissions A-24
A3.1 10 CFR 20 A-24
A.4 References A-26
A-2
-------
APPENDIX A
NRC's ORGANIZATION, REGULATIONS, AND CONTROLS
A.1 ORGANIZATION AND RESPONSIBILITIES OF NRC
The origins of the present NRC lie in the early days of the Manhattan Project in
1942. Over time, the NRC's responsibilities have evolved from overseeing the post-war
regulation and development of uses of atomic energy to the current licensing and
regulation of nuclear facilities and materials under the authority of the Atomic Energy
Act (AEA)(EEI86).
The NRC regulates the civilian uses of source, byproduct, and special nuclear
materials arid nuclear reactb'rs in the United States. This mission is accomplished
through the development 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 licensing the possession, use, and disposal of nuclear materials;
and inspection and enforcenlent activities to ensure compliance with these controls and
the conditions imposed through permits and licenses.
A.1.1 Basic Functions
The NRC's responsibilities include protecting public health arid safety; protecting
the environment; protectirig and safeguarding rnaterials and plarits in the interest of
national security; and ensuring conformity with antitrust laws. During fiscal year 1990,
the NRC had approximately 3,200 employees arid a budget of Over $400 million to carry
out three basic functions: regulatory research and standards development, licensing, and
inspection and eriforcement.
As part of its regulatory research and standards development function, the NRC is
mandated by law to conduct an extensive confirmatory research program in the areas of
safety, safeguards, and environmental assessment. The Commission establishes regula-
tions, standards, and guidelines governing the various licensed u§es" of nuclear facilities
and materials.
A-3
-------
In its licensing function, the agency reviews and issues licenses for the construc-
tion 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.
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.l.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 spokes-
man for the 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.l.2.2 NRC Offices. The 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.
A-4
-------
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 (NMSSX 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
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 faculties 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 spent fuel for recycle.1
Most of the manufacturing operations that make up the nuclear fuel cycle are
licensed by the 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.
The NRC regulates approximately 8200 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. Most of the NRC's licenses are administered by NRC's Regional Offices.
1 The latter step is not being performed in the United States.
A-5
-------
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
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 (AEODX This Office
independently analyzes and evaluates operational safety data associated with NRC-
licensed activities to identify issues that require action by the NRC or 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 enforce-
ment 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 modifica-
tion, suspension, or revocation.
Regional Offices. The 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, the 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.
A-6
-------
A. 1.2.3 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. The 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 the 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.
A-7
-------
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 facilities.
Each agreement provides that the State will use its best efforts to maintain
continuing compatibility with the NRC's program. The 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.
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.
A-8
-------
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 State Programs in the
Office of Governmental and Public Affairs. This Commission-level office was
established as part of an overall NRC reorganization which became effective April 12,
1987. The 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 the NRC. This is a general
index, and actual staffing needs will vary according to the particular circumstances in any
given State.
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 the 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-9
-------
A.2.1.1 Rulemaking and Regulatory Guides. The term nilemaking 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 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.
When the development of a rule or a guide reaches the point where it is present-
ed 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.
A-10
-------
The RES staff, in consultation with the affected prpgram office, evaluates public
comments received on a proposed rule. The Commission has used both rulemaking
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 receiyed and evaluated, the staff
modifies the rule as necessary, repeats the entire review prqpess followed for the
proposed rule, and returns the rulemaking 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, the NRC is calling for or recommending that the licensees take specific
steps.
A.2.1.3 NRC Reports. NRC reports (usually referred to generically as NUREGs) are
prepared by the NRC's staff, contractors, or national laboratories and provide the
technical basis for decision making. Special categories of such reports include Safely
Evaluation Reports (SERs), Environmental Impact Statements (EISs), and Standard
Review Plans (SRPs). The 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 staff's review
process.
Standard Review Plans address the responsibilities of the persons performing the
review, the masters that are reviewed, the Commission's regulatiqns and acceptance
A-ll
-------
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 the NRC's ability to
establish and maintain 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.
The 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.
A-12
-------
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. The 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, the NRC will issue a specific license
authorizing the possession and use of byproduct material on NRC Form 374, "Byproduct
Material License."
To help license applicants prepare the application and design their radiation
safety programs, the 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
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
A-13
-------
NUREG-0267
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 ah environmental report discussing the operation's impact on
the health and safety of the public and on the environment. The Commission us"es this
information to determine whether the applicant's proposed activities will, among other
things, 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 40.31, "Application for Specific Licenses," of 10 CFR Part 40, arid in
10 CFR Part 51, "Licensing arid 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 safely, 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 Envifbnmerital Report (ER) pursuant to 10 CFR Part 51.
Based on the inforrriation provided in these reports, the NRC will in turn develop a
Safety Evaluation Report (SER) arid an Envkonmental 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 Enviroririiental
Report if the 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.
A-14
-------
To help licensees develop the application, the 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
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
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
Standard Format and Content for the Health and
Safely 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 Mining
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
A-15
-------
NUREG/CR-2011
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.
The NRC initiates a comprehensive technical review of the license application
and any supporting documents after initial acceptance review and docketing. During this
period, the 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 the NRC's staff and a formal letter of recom-
mendation from the ACRS to the Chairman of the NRC.
In determining whether to grant a construction permit, the 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 authorizing construction to
begin.
Prior to anticipated completion of construction, the applicant submits an updated
license application to the NRC in support of obtaining a license to operate. The NRC's
staff and the ACRS again conduct technical reviews which, if favorable, result in the
issuance of a Safety Evaluation Report by the NRC's staff and a formal letter of
recommendation from the ACRS to the Chairman of the 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
A-16
-------
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 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,
the 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 if the NRC's 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, the 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
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
A-17
-------
Regulatory Guide 4.9
Regulatory Guide 8.10
Regulatory Guide 10.3
Preparation of Environmental Reports for Commercial
Uranium Enrichment Facilities
Operating Philosophy for Maintaining Occupational
Radiation Exposures As Low As Is Reasonably
Achievable
Guide for the Preparatipn of Applications for Special
Nuclear Material Licenses of Less Than Critical Mass
Quantities
A.2.3 Airborne Emissions Monitoring
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.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 charcbal 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-18
-------
(b)
(c)
(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.
A licensee is required to administer radioactive gases only in rooms that are at
negative pressure compared to surrounding rooms>
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 hi (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 II of Appendix B to 10 CFR Part 20. The
NRC or 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.
If aerosols arid 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
A-19
-------
gases are directly vented to the atmosphere, airborne effluent concentrations must be
calculated. For medical institutions, the 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.
(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
A-20
-------
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, (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 effluence 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 the 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 50.59X
Once a license to operate has been issued, the 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 con-
tained in the FSAR.
A-21
-------
Records and Reports (10 CFR 50.71). Each licensee and each holder of a con-
struction 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.
A.2.3.4 Part 70 Licenses. The requirements for airborne emissions monitoring are
essentially the same as for Part 40 licensees.
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 the 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-22
-------
A2.4.2 Part 40 Licenses. Initial 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.
A2.4.4 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 Prograins
The objective of the NRC's enforcement programs is to protect the public health
arid safety by ensuring that licensees comply with regulatory requirements. The NRC's
enforcement policy, contained in 10 CFR Part 2, Appendix C, calls for strong enforce-
ment 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
the NRC's requirements.
used in all instances of noncompliance with
Civil penalties -- considered for licensees who evidence significant or
repetitive instances of noncompliance, especially if a previous Notice of
A-23
-------
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-kind 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.
A3 CONTROLS APPLICABLE TO AIRBORNE EMISSIONS
Current regulations limiting routine radionuclide airborne emissions from NRC-
licensed faculties 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 gaseous and liquid effluents and considers exposures from all sources.
Part 20 also imposes the requirement that exposures be as low as reasonably achievable
(ALARA). Licensees may demonstrate compliance with this limit using the effluence
concentrations set forth in Table 2 of Appendk 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.3.1 10
20
The portions of Part 20 that apply to airborne radionuclide emissions from
licensed facilities are those which set permissible levels of radiation (in mrem per unit
time) in unrestricted areas and those which establish limits (in curies released to the
environment) on radioactivity in effluents to unrestricted areas. Part 20,1105 states that
the Commission will approve an application for a license to possess or use radioactive
materials and any other source of radiation if the applicant can demonstrate that
radionuclide releases are not likely to cause any individual in an unrestricted area to
receive a dose to the whole body in excess of 100 mrem/yr. Part 20 also requires that
A-24
-------
no individual in an unrestricted area receive a dose in excess of 2 mrern in any 1 hour or
100 mrem in any 7 consecutive days.
Part 20 limits, in part, the release of radioactive material to unrestricted areas to
levels that will not result in average annual radionuclide concentrations in air and water
that exceed the limits set forth in Table 2 of Appendix B of Part 20. This is a secondary
standard designed to provide a level of assurance that the primary health-based standard
of 100 mrem/yr ede is not exceeded.
In 1981, Part 20 was amended to adopt EPA standards set forth hi 40 CFR Part
190. (Part 190 requires that the radiation doses to real individuals from all uranium fuel
cycle sources^ including all gaseous and liquid effluent pathways and direct radiation, not
exceed 25 mrem/yr to the whole body or any organ, except the thyroid, which is limited
to 75 mrem/yr.)
In addition to these numerical standards, Part 20 requires that each licensee make
every reasonable effort to maintain radiation exposures, and releases of radioactive
material in effluents to unrestricted areas, as low as is reasonably achievable. The term
"as low as is reasonably achievable," as defined in the glossary of 10 CFR Part 20, means
"as low as is reasonably achievable taking into account the state of technology, and the
economics of improvement in relation to benefits to the public health and safety, and
other societal and socioeconomic considerations in relation to the utilization of atomic
energy in the public interest." Thus, the explicit consideration of cost is intended.
On January 9, 1986, major revisions to Part 20 (51 FR 1092) were proposed to
keep pace with changes in the scientific knowledge underlying radiation protection that
have occurred since Part 20 was originally issued more than 30 years ago. The revised
rule implements the 1987 Presidential Guidance on occupational radiation protection
and the recommendations of scientific organizations to establish risk-based limits and a
system of dose limitation in accordance with the guidance published by the International
Committee on Radiation Protection.
Revised Part 20 requires that (1) the total effective dose equivalent to individual
members of the public shall not exceed 100 mrem/yr, and 2 mrem in any 1 hour for
external exposures, and (2) a licensee or applicant may apply for prior NRC authori-
A-25
-------
zation to operate up to an annual limit for an individual member of the public of 500
mrem. Part 20 also requires appropriate surveys to ensure that the dose limits are not
exceeded. In addition, Tabl6 2 of the revised Part 20 presents Derived Air Concen-
trations (DACs) that licensees may use to demonstrate compliance with the limits. The
values for air concentrations are derived to assure that doses will be less than 50
mrem/yr. The revised rule fetainS the requirement for uranium fuel cycle facilities to
comply with the standards set forth' in 40 CFR 190.
A.4 REFERENCES
EEI86 Edison Electric Institute, "A Report on the Management Structure of the
Nuclear Regulatdry Commission Prepared for the Edison Electric
Institute," June 19*86.
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-26
-------
APPENDIX B
SELECTED NRC REGULATORY GUIDES
This appendix provides a partial list of the regulatory guides published by
the NRC that are relevant to airborne effluents from nonreactor NRC-
licensed facilities.
B-l
-------
NO.
TITLE
REV. DATE
DIVISION 2 - RESEARCH AND TEST REACTORS
22
32
33
35
3.6
3.7
3.8
3.12
335
326
332
333
334
335
339
3.42
3.44
3.46
3.48
3.49
3.51
3.52
355
356
Development of Technical Specifications for Experiments in Research Reactors
DIVISION 3 - FUELS AND MATERIALS FACOJnES
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-Basin 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
-
-
11/73
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
-------
3.59
3.60
3.61
3.62
3.63
3.64
3.65
4.1
4.5
4.6
4.9
4.13
4.14
4.15
4.16
4.17
4.18
5.4
55
5.13
5.18
5.24
5.33
5.42
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
DIVISION 4 - ENVIRONMENTAL AND SITING
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 Thennoluminescence Dosimetry: Envi-
ronmental 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
DIVISION 5 - MATERIALS AND PLANT PROTECTION
Standard Analytical Methods for the Measurement of Uranium Tetraflouride (UF4) and Uranium
Hexaflouride (UF6)
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
-
.
-
-
-
-
-
1
.
-
1
1
1
1
1
1
-
-
-
_
-
.
-
-
03/87
03/87
02/89
02/89
03/88
06/89
08/89
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
02/73
02/73
11/73
01/74
06/74
06/74
01/75
B-3
-------
5.45
551
558
5.62
82
8.10
8.18
821
833
824
825
830
831
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)
Management Review of Nuclear Material Control and Accounting Systems
Considerations for Establishing Traceability of Special Nuclear Material Accounting Measurements
Reporting of Safeguards Events
DIVISION 8 - OCCUPATIONAL HEALTH
Guide for Administrative Practices in Radiation Monitoring
Operating Philosophy for Maintaining Occupational Radiation Exposures As Low As Is Rea-
sonably Achievable
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
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
-
1
1-R
1
1
1
1
-
-
-
12/74
06/75
11/78
02/80
02/81
11/87
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
102
103
10.4
105
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 o,f 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
an/16
04/77
07/76
03/77
12/87
09/76
12/80
, 02/77
08/79
01/79
10/80
(38/87
03/87
-------
APPENDIX C
DESCRIPTION OF NRC AND AGREEMENT STATE
LICENSED ACTIVITIES
This appendix (describes the activities for which an NRC or Agreement
State license is required (NRC91).
C-l
-------
CONTENTS
C.1 General c'3
C.2 Byproduct Material Program (10 CFR 30) C-5
O3 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
-------
APPENDIX C
DESCRIPTION OF NRC AND AGREEMENT STATE
LICENSED ACTIVITIES
C.1 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, and 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 the NRC to make day-to-day decisions about
the program.
*
The program codes referred to are designated by NRC and may or may not be used by Agreement
States.
C-3
-------
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.
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 II. 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
-------
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. & 35)
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
humans. Separate licenses are issued to authorize teletherapy. Radioactive
material administered to patients is an in-vivo procedure.
C-5
-------
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 - Well-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
-------
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 IT-192.
Radiography can be conducted either hi 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 hi 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
-------
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.
C3 SOURCE MATERIAL PROGRAM (10 CFR 40)
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
C-8
-------
processing or fabrication 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 milling 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.
Uranium Hexafluoride Production Plants - These licenses are issued for the
possession and use of uranium to allow the conversion of yellowcake and/or ore
concentrates to uranium hexafluoride (UF6).
Solution Mining - These licenses are issued for the extraction of uranium from
uranium ores. The only mining operation licensed by the NRC is solution mining,
which is leaching of ore by injection of liquid chemicals into the geologic
formation.
C-9
-------
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
-------
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 the NRC of their intent to terminate a
portion or all of their activities involving special nuclear material and/or have
submitted to the 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
-------
Decommissioning of Uranium Fuel Fabrication Plants - These licenses are issued
to facilities that have notified the NRC of their intent to terminate a portion or
all of their activities involving special nuclear material and/or has submitted to
the 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.
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.
Decommissioning of Critical Mass - Other Than Universities - These licenses are
issued to faculties that have notified the NRC of theirs intent to terminate a
portion or all their its activities involving special nuclear material and/or has
submitted to the NRC a plan and schedule for the facilities, properly, 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 instruments".
C-12
-------
Special Nuclear Material. Plutonium Neutron Sources. Less Than 200 Grams -
These licenses aje issued for the possession and use of small quantities of
plutonium (less than 200 grams total) usuaUy 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
-------
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
-------
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.1.3 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 Faculties (10 CFR 40 and 70) D-14
D3.1 Source Material Licensees (10 CFR 40) D-15
D.3.2 Special Nuclear Material Licensees (10 CFR 70) D-19
Table D-l. NRC licensees other than power reactors D-4
D-2
-------
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. The 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 L" 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 faculties 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
D-3
-------
Table D-l. NRC licensees other than power reactors.
J.
PROGRAM
CODE
01100
01110
01120
01200
02110
02120
02121
02200
02201
02209
02220
02400
02410
02500
02511
02532
02513
03211
03212
03213
03214
03218
03232
03234
03610
03611
03612
03613
03620
11200
11210
11230
11300
11500
11600
11800
21130
21135
21215
21240
\. NRC-IIOENSEES CoVi^
PRO0R^^DEDESCRIP%(5N '**"
' '
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
Li-Vitro Testlab
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
,
., % ., f~f'.-f . v f *
NUMBER QF^
ACtlVe:
LICENSEES ;
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
-------
Table D-l. NRC licensees other than power reactors (continued).
A. NRC4JECENSEES COVERED BY THE RANDOM SURVEY*
PROGRAM
CODE
21310
21320
21325
22110
22111
22170
25110
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 U-235 < 350g,
U-233 < 200g
Special Nuclear, General License Distribution
Transport - Private Carriage
Miscellaneous
SUBTOTAL
NUMBER OF
ACTIVE
LICENSEES
10
4
0
16
12
0
2
13
3509
1 10 CFR 50 licensees (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) the
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. NRC-L1CENSEES INCLUDED IN THE DESIGNATED SURVEY - PJHOR WAI£TATI0NS
UPDATED AND NEW SOURCE CATEGORIES ADBED*
PJtOGRAM
CODE
03231
03233
03235
06100
11100
11220
11400
11700
21210
^^Wta?se*. .' .y
Test and Research Reactors
Waste Disposal - Burial
Waste Disposal Service - Incineration
Incineration, Non-Commercial
Low Level Waste Storage
Mills
Source Material Military Munition Testing
Uranium Hexaflouride Production Plants
Rare Earth Extraction and Processing
Uranium Fuel Processing Plants
SUBTOTAL
NUMBER OF
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) the EPA has
evaluated in prior studies and in need of updating (e.g,, site-specific demographics), and {2)' specific
facilities in which the EPA has developed an interest (e.g., rare earth processors).
D-5
-------
Table D-L NRC licensees other than power reactors (continued).
C NRC-LICENSEES EXCLUDED FROM THE RANDOM SURVEY AND DESIGNATED
SURVEY -SiAUED 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
: PROGRAM CODE DESCRIPTION
j X. w
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
Reid 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
NUMBER OF
ACTIVE
LICENSEES
50
223
45
54
6
4
772
1529
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
D-6
-------
Table D-l. NRC licensees other than power reactors (continued).
C. NRC-LICENSEES EXCLUDED FROM THE RANDOM SURVEY AND DESIGNATED
SURVEY - SEALED SOURCES3
PROGRAM
CODE
22120
22130
22140
22150
22151
22160
22161
22162
23100
PROGRAM CODE DESCRIPTION
Special Nuclear Material, Plutonium Neutron Sources, <200g
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
92
Q
10
15
3
68
4
1
5
4609
8244
Sealed source users are excluded from the Random. Survey and Designated Survey data bases
because the potential for airborne radioactive effluents is essentially zero.
research and patient administration of radippharrnaceuticals for both diagnostic and
therapeutic purposes.
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 administratiQn and research.
D-7
-------
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.
The mixing of the radionuclide with some other substance means that the physical
state of a radiopharmaceutical 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.
D-8
-------
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 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 pre-
paration 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.
D-9
-------
These elution procedures and dose divisions are conducted in a fume hood, with the
generator shielded to prevent external irradiation of the technicians.
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 may simply be extracted
from bulk stocks and administered, or the radionuclide may be subjected to additional
chemical or physical processing.
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 then: own
production faculties 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.
D-10
-------
Radiopharmaceutical generators are designed and produced as closed aseptic
systems using some type of chromatographic column. Typically, this chromatographic
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 minimi/.^
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.
Independent radiopharmacies are a relatively recent phenomenon. Generally
located in large cities, these facilities serve as distribution faculties. 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-Illuminating 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.
D-ll
-------
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-iUuminating devices^ obtain the bulk radionuclides in either
gaseous or (rarely) liquid form from a vendor. In the case of self-illuminating 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.
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.
D-12
-------
Incineration ^ s
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.
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 LICENSEES (10 CFR 50, TYPE 104)
D.2.1 Test and Research Reactors
The 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 ma-
terials, and to produce radioactive materials used by sealed source and
radiopharmaceutical manufacturers.
D-13
-------
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 dis-
persed 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.
In 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 FACHITIES (10 CFR 40 and 70)
The uranium fuel cycle includes uranium mills, uranium hexafhioride conversion
facilities, uranium enrichment facilities, light-water reactor fuel fabricators, light-water
power reactors, and fuel reprocessing plants. With the exception of the uranium
D-14
-------
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 the 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 40)
Uranium Mills
Uranium mills extract uranium from ores which contain only 0.01 to 0.3 percent
U3O8. Uranium mills, typically located near uranium mines in the western United States,
are 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 the
NRC or 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 milk
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 then: respective decay products (e.g., radium, lead, radon).
These radionuclides, with the exception of radon, are released as particulates.
D-15
-------
Depleted Uranium 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 the 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 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.
D-16
-------
Uranium Conversion Facilities
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,
fllinois 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 UF6. 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
Two industrial processes are used for uranium hexafluoride production, the dry
hydrofluor 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 hydrofluor 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 UF^ Impurities are separated either as volatile
compounds t>r as a relatively concentrated and insoluble solid waste that is dried
and drummed for disposal.
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.
D-17
-------
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 raffmate is
impounded in ponds at the plant site.
Rare Earth Extraction and 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 iinlimited. 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.
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 qf 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
D-18
-------
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 70}
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 faculties 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 hexafluoride 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 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
D-19
-------
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-ThO2). 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 hi a fluidized bed furnace. Fuel rods are
formed by injecting the coated kernels and a matrix material into a hot mold. The fin-
ished rods are then inserted into a graphite block to form the final fuel assembly.
D-20
-------
APPENDIX E
QUALITY ASSURANCE CRITERIA FOR NUCLEAR POWER PLANTS
AND FUEL REPROCESSING PLANTS
E-l
-------
APPENDIX E
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 safety. 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 faculties
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 - Organisation - 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.
1 Appendix B to 10 CFR 50, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing
Plants."
E-2
-------
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 performance and long-term stability of the site, and to predict the environ-
mental 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.
E-3
-------
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 predeter-
mined 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.
E-4
-------
APPENDIX F
NRC AGREEMENT STATES AND STATE DIRECTORS
F-l
-------
APPENDIX F
NRC AGREEMENT STATES AJSiD STATE DIRECTORS1
STATE
1. Alabama
2. Arizona
3. Arkansas
4. California
5. Colorado
6. Florida
, STATE DOCTOR
Mr. Aubrey V. Godwin, Chief
Bureau of Radiological Health Environmental Health
Administration
Room 314, State Office Building
Montgomery,, Alabama 36130
(205)261-5313
Mr. Charles F. Tedford, Director
Radiation Regulatory Agency
4814 South 40th Street
Phoenix, Arizona 85040
(602)255-4845
Ms. Greta Dicus, Director
Division of Radiation Control and Emergency Management
Department of Health
4815 West Markam
Little Rock, Arkansas 72205-3867
(501)661-2301
Mr. Jack McGurk, Chief
Environmental Health Branch
State Department of Health
714/744 P Street, Room 498
Sacramento, California 95814
(916)332-2073 or 3482
Mr. Robert Quillin, Director
Radiation Control Division
Office of Health Protection
Department of Public Health
4210 East llth Avenue
Denver, Colorado 80220
(303)331-8480
Mary E. Clark, Ph.D., Director
Office of Radiation Control
Department of Health & Rehabilitative Services
1317 Winewood Blvd.
Tallahassee, Florida 32399-0700
(904)487-1004
F-2
-------
NRC AGREEMENT STATES, AND STATE DIRECTORS1
STATE
7. Georgia
8. Illinois
9. Iowa
10. Kansas
11. Kentucky
12. Louisiana
13. Maryland
STATE DIRECTOR
James L. Setser, Chief
Environmental Protection
Dept of Natural Resources
Floyd Towers East 1166
205 Butler Street
Atlanta, Georgia 30309
(404)656-4713
Thomas W. Ortcigar, Director
Department of Nuclear Safety
1035 Outer Park Drive
Springfield, Illinois 62704
(217)785-9868
Donald A. Flater, Chief
Bureau of Radiological Health
Department of Public Health
Lucas State Office Building
Des Moines, Iowa 50319
(515)281-3478
Mr. Gerald W. Allen, Chief
X-Ray & Radioactive Materials
Department of Health & Environment
109 S.W. 9th Street
Topeka, Kansas 66620
913)296-1562
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
Mr. Glenn Miller, Administrator
Radiation Protection Division
Office of Air Quality & Nuclear Energy
P.O. Box 82145
Baton Rouge, Louisiana 70884
(504)765-0160
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
F-3
-------
JmCAGREEME^^^sfATESA3S]DASTA^^EDmECTpRS1 '
STATE
14. Mississippi
15. Nebraska
16. Nevada
17. New Hampshire
18, New Mexico
19. New York
,, ' ^ , STATE DIRECTOR
Mr. Eddi& 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
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
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
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
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
Ms. Donna Ross, Energy Planner
Division of Policy Analysis and Planning
2 Rockefeller Plaza
Albany, New York 12223
(518)473-0048
F-4
-------
NRC AGREEMENT STATES AND STATE DIRECTORS1
STATE
20. North Carolina
21. North Dakota
22. Oregon
23. Rhode Island
24. South Carolina
25. Tennessee
STATE DIRECTOR
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
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
F-5
-------
NRG AGREEMENT! :ST%$S 1$& ''STATE DIRECTORS1
STATE
26. Texas
27. Utah
28. Washington
-.*,V W,1J,,7 _, STATE DIRECTOR.
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
Mr. Terry R. Strong, Director
Office of Radiation Protection
Department of Health
Mail Stop LE-13
Olympia, Washington 98504
(206)586-8949
1. As of August 1991.
F-6
-------
APPENDIX G
RANDOM SURVEY QUESTIONNAIRE
G-l
-------
APPENDIX G
UNITED STATES ENVIRONMENTAL PROTECTION AGENCY
WASHINGTON, O.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
Printtd on Recycled Paper
-------
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,
William G. Roserfberg
Assistant Administrator
for Air and Radiation
Enclosure
G-3
-------
SURVEY FORM
Facility Name_
Address
City
State
Identify a person from whom clarification or additional .information can be obtained, if necessary.
Name , Telephone ( )_
Does your facility handle 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 activtti*s conducted at your facility which involve unsealed forms of
radlonucllde* (check all that apply):
[ ] Accelerator
[ ] Research/Test Reactor
[ ] Nuclear Medicine (Diagnostic only)
[ j Nuclear Medicine (Diagnostic and Therapeutic)
[ j Manufacturer of Teletherapy Equipment
[ j Manufacturer of Medical Implant Needles or Seeds
[ j Manufacturer of Pacemakers
[ j Manufacturer of Industrial/Scientific Gauging Equipment
[ j Manufacturer of Self-Illuminating Devices
[ ] Producer of Radlopnarmaceuticals
[ ] Producer of Radio-Labelled Compounds for Research
[ j Producer of Munitions using Depleted Uranium
[ j Producer of Shielding using Depleted Uranium
[ j Thorium/Rare Earth Processing/Recovery
[ ] Low-Level Waste Disposal Facility
[ ] Low-Level Waste Incinerator
[ j Low-Level Waste Transfer Agent (Prepackage only .re-package )
[ ] Research Laboratory (indicate fleld of research )
[ ] Other (please specify )
GENERAL INSTRUCTIONS
1. You mutt provide th« information requested on pages 2-3 of this form separately for each
building at your facility where radionuciides in unsealed form 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 *U*.
G-4
-------
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.
STACK/VENT
1
2
3
4
5
Height
(m)1
Diameter
(m)2
Flow Rate
(m3/sec)
Temperature
(°F)3
Effluent Controls
(specify type)
If there are more than 5 stacks/vents scoring this building, check here and provide the information for the additional
stacks/vents on a separata sheet of paper which dearly designates the building name.
' Distance from the ground to the top of the stack/vent
2lf the stack/vent is rectangular, give its length and width.
3lf the exit temperature is approximately the same as the ambient temperature, simply enter an A for ambient.
NOTE: If the data you have is in units other than those requested and you are uncertain of the conversion, provide your
data with the units clearly indicated; e.g. ft. for feet. CFM for ft3/min, and °C for degrees Celsius.
Step 3.
Does anyone live in this building?
If YES, enter the distance along the buBding surface from the stack/vent to the nearest residence in the
building (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 etc.)
Step 4. Is there en office, school or business, not part of the facility covered by the NRC or
stats license, in this building?
If YES, enter the distance along the buftdlng surface from the stack/vent to the nearest office, school
or business. (meters)
If NO, enter the distance from the stack/vent 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 enter >2000.
Vegetables^
Meat
Milk
G-5
-------
Step 6.
Complete the table below.
You must report ail radionuclidea in unsealed forms used in the building even if you do not
believe that they are being released.
If you cannot provide th
-------
APPENDIX H
DOSE CALCULATION ASSUMPTIONS
This appendix provides details of calculational assumptions made regarding
factors having a significant effect on dose.
H-l
-------
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 H-3
Table H-2. Doses above 1 mrem/yr, no wind rose H-5
H-2
-------
APPENDIX H
DOSE CALCULATION ASSUMPTIONS
H.1 ASSUMPTIONS RELATED TO SOURCE TERM
Xe-133 Release from Hospitals
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 hospitals indicated on the survey form that they did trap
Xe-133. Trapping the gas reduces the amount available to be released into the envi-
ronment.
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
53
8.0
AE3&K133 Trapped I
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
-------
It is concluded that neglecting Xe-133 trapping by hospitals has a negligible effect
upon the distribution of estimated doses.
Release Fractions
For those faculties that did not provide site-specific release rates, the default
release fraction of IE-OS, 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
radiophannacy.
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 radionuclides); 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
H-4
-------
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.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.
Fadfify
MR
MR
RW
NR
NR
NR
NR
As Calculated
1.1
1.7
1.8
2.0
3.9
53
8.0
Person in Max Sector
1.1
1.7
2.0
2.0
3.9
53
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
-------
Same Building Effect
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 he
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,
H-6
-------
such as students attending school, the average age may be less than that of standard man
and the annual dose may be greater than that calculated. The models used by the EPA
do not explicitly account for these factors, but since the cancer risks are assumed to
result from a lifetime of exposure, underestimates associated 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 associated 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/Oi 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 associated 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
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
-------
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.
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 I-
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.
H-8
-------
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.
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
/um 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.
H-9
-------
Inhalation DCF for 1-131. The effective whole body DCF for the inhalation of I-
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., £2
= 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 £2 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.
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 I-
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.
H-10
-------
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 the 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 semi-
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 (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 Conversion Factors for Inhalation,
Submersion, and Ingestion," Federal Guidance Report No. 11, EPA-520/1-
88-020, September 1988.
H-ll
-------
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.
Mu91 Mullins, T. J., DuPont/Merck, letter to S. Deal, December 1991.
NAS90 National Academy of Sciences, "Health Effects of Exposure to Low Levels
of Ionizing Radiation," BEIR V, NAS/NRC, 1990.
NRC85c 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.
Wi82 Wilson, D. J., and Britter, R. E., "Estimates of Building Surface
Concentrations from Nearby Point Sources." Atmospheric Environment. 16,
2631, 1982.
H-12
-------
-------
------- |