Regulatory Guide 1.183
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U.S. NUCLEAR REGULATORY COMMISSION July 2000
REGULATORY
GUIDE
OFFICE OF NUCLEAR REGULATORY RESEARCH
REGULATORY GUIDE 1.183
(Draft was issued as DG-1081)
ALTERNATIVE RADIOLOGICAL SOURCE TERMS FOR
EVALUATING DESIGN BASIS ACCIDENTS
AT NUCLEAR POWER REACTORS
ii
AVAILABILITY INFORMATION
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iii
TABLE OF CONTENTS
A. INTRODUCTION ........................................................1
B. DISCUSSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
C. REGULATORY POSITION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1. IMPLEMENTATION OF AST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1.1 Generic Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
1.2 Scope of Implementation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
1.3 Scope of Required Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
1.4 Risk Implications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
1.5 Submittal Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
1.6 FSAR Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
2. ATTRIBUTES OF AN ACCEPTABLE AST . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
3. ACCIDENT SOURCE TERM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.1 Fission Product Inventory . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
3.2 Release Fractions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
3.3 Timing of Release Phases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
3.4 Radionuclide Composition . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
3.5 Chemical Form . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15
3.6 Fuel Damage in Non-LOCA DBAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
4. DOSE CALCULATIONAL METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
4.1 Offsite Dose Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
4.2 Control Room Dose Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17
4.3 Other Dose Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
4.4 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
5. ANALYSIS ASSUMPTIONS AND METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . . . . 20
5.1 General Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20
5.2 Accident-Specific Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
5.3 Meteorology Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
6. ASSUMPTIONS FOR EVALUATING THE RADIATION DOSES FOR EQUIPMENT
QUALIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
D. IMPLEMENTATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23
REFERENCES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .24
iv
APPENDICES
A. Assumptions for Evaluating the Radiological Consequences of a LWR
Loss-of-Coolant Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1
B. Assumptions for Evaluating the Radiological Consequences of a Fuel
Fuel Handling Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1
C. Assumptions for Evaluating the Radiological Consequences of a BWR
Rod Drop Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-1
D. Assumptions for Evaluating the Radiological Consequences of a BWR Main
Steam Line Break Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-1
E. Assumptions for Evaluating the Radiological Consequences of a PWR Main
Steam Line Break Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-1
F. Assumptions for Evaluating the Radiological Consequences of a PWR Main
Steam Generator Tube Rupture Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . F-1
G. Assumptions for Evaluating the Radiological Consequences of a PWR Locked
Rotor Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . G-1
H. Assumptions for Evaluating the Radiological Consequences of a PWR Rod
Ejection Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . H-1
I. Assumptions for Evaluating Radiation Doses for Equipment Qualification . . . . . . . . . . . . I-1
J. Analysis Decision Chart . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . J-1
K. Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . K-1
1 Applicants for a construction permit, a design certification, or a combined license that do not reference a standard design
certification who applied after January 10, 1997, are required by regulation to meet radiological criteria provided in 10 CFR
2 As defined in 10 CFR 50.2, design bases means information that identifies the specific functions to be performed by a structure,
system, or component of a facility and the specific values or ranges of values chosen for controlling parameters as reference
bounds for design. These values may be (1) restraints derived from generally accepted "state of the art" practices for achieving
functional goals or (2) requirements derived from analysis (based on calculation or experiments or both) of the effects of a
postulated accident for which a structure, system, or component must meet its functional goals. The NRC considers the accident
source term to be an integral part of the design basis because it sets forth specific values (or a range of values) for controlling
parameters that constitute reference bounds for design.
1.183-1
A. INTRODUCTION
This guide provides guidance to licensees of operating power reactors on acceptable
applications of alternative source terms; the scope, nature, and documentation of associated
analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals.
This guide establishes an acceptable alternative source term (AST) and identifies the significant
attributes of other ASTs that may be found acceptable by the NRC staff. This guide also identifies
acceptable radiological analysis assumptions for use in conjunction with the accepted AST.
In 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” Section
50.34, “Contents of Applications; Technical Information,” requires that each applicant for a
construction permit or operating license provide an analysis and evaluation of the design and
performance of structures, systems, and components of the facility with the objective of assessing
the risk to public health and safety resulting from operation of the facility. Applicants are also
required by 10 CFR 50.34 to provide an analysis of the proposed site. In 10 CFR Part 100,
“Reactor Site Criteria,” Section 100.11,1 “Determination of Exclusion Area, Low Population Zone,
and Population Center Distance,” provides criteria for evaluating the radiological aspects of the
proposed site. A footnote to 10 CFR 100.11 states that the fission product release assumed in
these evaluations should be based upon a major accident involving substantial meltdown of the
core with subsequent release of appreciable quantities of fission products.
Technical Information Document (TID) 14844, “Calculation of Distance Factors for Power
and Test Reactor Sites” (Ref. 1), is cited in 10 CFR Part 100 as a source of further guidance on
these analyses. Although initially used only for siting evaluations, the TID-14844 source term has
been used in other design basis applications, such as environmental qualification of equipment
under 10 CFR 50.49, “Environmental Qualification of Electric Equipment Important to Safety for
Nuclear Power Plants,” and in some requirements related to Three Mile Island (TMI) as stated in
NUREG-0737, “Clarification of TMI Action Plan Requirements” (Ref. 2). The analyses and
evaluations required by 10 CFR 50.34 for an operating license are documented in the facility final
safety analysis report (FSAR). Fundamental assumptions that are design inputs, including the
source term, are to be included in the FSAR and become part of the facility design basis.2
Since the publication of TID-14844, significant advances have been made in understanding
the timing, magnitude, and chemical form of fission product releases from severe nuclear power
plant accidents. A holder of an operating license issued prior to January 10, 1997, or a holder of a
renewed license under 10 CFR Part 54 whose initial operating license was issued prior to January
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10, 1997, is allowed by 10 CFR 50.67, “Accident Source Term,” to voluntarily revise the accident
source term used in design basis radiological consequence analyses.
In general, information provided by regulatory guides is reflected in NUREG-0800, the
Standard Review Plan (SRP) (Ref 3). The NRC staff uses the SRP to review applications to
construct and operate nuclear power plants. This regulatory guide applies to Chapter 15.0.1 of the
SRP.
The information collections contained in this regulatory guide are covered by the
requirements of 10 CFR Part 50, which were approved by the Office of Management and Budget
(OMB), approval number 3150-0011. The NRC may not conduct or sponsor, and a person is not
required to respond to, a collection of information unless it displays a currently valid OMB control
number.
B. DISCUSSION
An accident source term is intended to be representative of a major accident involving
significant core damage and is typically postulated to occur in conjunction with a large loss-of-coolant
accident (LOCA). Although the LOCA is typically the maximum credible accident, NRC staff
experience in reviewing license applications has indicated the need to consider other accident
sequences of lesser consequence but higher probability of occurrence. The design basis accidents
(DBAs) were not intended to be actual event sequences, but rather, were intended to be surrogates to
enable deterministic evaluation of the response of a facility’s engineered safety features. These
accident analyses are intentionally conservative in order to compensate for known uncertainties in
accident progression, fission product transport, and atmospheric dispersion. Although probabilistic
risk assessments (PRAs) can provide useful insights into system performance and suggest changes in
how the desired depth is achieved, defense in depth continues to be an effective way to account for
uncertainties in equipment and human performance. The NRC’s policy statement on the use of PRA
methods (Ref. 4) calls for the use of PRA technology in all regulatory matters in a manner that
complements the NRC’s deterministic approach and supports the traditional defense-in-depth
philosophy.
Since the publication of TID-14844 (Ref. 1), significant advances have been made in
understanding the timing, magnitude, and chemical form of fission product releases from severe
nuclear power plant accidents. In 1995, the NRC published NUREG-1465, “Accident Source Terms
for Light-Water Nuclear Power Plants” (Ref. 5). NUREG-1465 used this research to provide
estimates of the accident source term that were more physically based and that could be applied to the
design of future light-water power reactors. NUREG-1465 presents a representative accident source
term for a boiling-water reactor (BWR) and for a pressurized-water reactor (PWR). These source
terms are characterized by the composition and magnitude of the radioactive material, the chemical
and physical properties of the material, and the timing of the release to the containment. The NRC
staff considered the applicability of the revised source terms to operating reactors and determined that
the current analytical approach based on the TID-14844 source term would continue to be adequate to
protect public health and safety. Operating reactors licensed under that approach would not be
required to re-analyze accidents using the revised source terms. The NRC staff also determined that
some licensees might wish to use an AST in analyses to support cost-beneficial licensing actions.
3 The NUREG-1465 source terms have often been referred to as the “revised source terms.” In recognition that there may be
additional source terms identified in the future, 10 CFR 50.67 addresses “alternative source terms.” This regulatory guide
endorses a source term derived from NUREG-1465 and provides guidance on the acceptable attributes of other alternative source
terms.
1.183-3
The NRC staff, therefore, initiated several actions to provide a regulatory basis for operating reactors
to use an AST3 in design basis analyses. These initiatives resulted in the development and issuance of
10 CFR 50.67 and this regulatory guide.
The NRC’s traditional methods for calculating the radiological consequences of design basis
accidents are described in a series of regulatory guides and SRP chapters. That guidance was
developed to be consistent with the TID-14844 source term and the whole body and thyroid dose
guidelines stated in 10 CFR 100.11. Many of those analysis assumptions and methods are
inconsistent with the ASTs and with the total effective dose equivalent (TEDE) criteria provided in 10
CFR 50.67. This guide provides assumptions and methods that are acceptable to the NRC staff for
performing design basis radiological analyses using an AST. This guidance supersedes corresponding
radiological analysis assumptions provided in other regulatory guides and SRP chapters when used in
conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67. The affected
guides will not be withdrawn as their guidance still applies when an AST is not used. Specifically,
the affected regulatory guides are:
Regulatory Guide 1.3, “Assumptions Used for Evaluating the Potential Radiological Consequences of
a Loss of Coolant Accident for Boiling Water Reactors” (Ref. 6)
Regulatory Guide 1.4, “Assumptions Used for Evaluating the Potential Radiological Consequences of
a Loss of Coolant Accident for Pressurized Water Reactors” (Ref. 7)
Regulatory Guide 1.5, “Assumptions Used for Evaluating the Potential Radiological Consequences of
a Steam Line Break Accident for Boiling Water Reactors” (Ref. 8)
Regulatory Guide 1.25, “Assumptions Used for Evaluating the Potential Radiological Consequences
of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized
Water Reactors” (Ref. 9)
Regulatory Guide 1.77, “Assumptions Used for Evaluating a Control Rod Ejection Accident for
Pressurized Water Reactors” (Ref. 10)
The guidance in Regulatory Guide 1.89, “Environmental Qualification of Certain Electric
Equipment Important to Safety for Nuclear Power Plant.” (Ref. 11), regarding the radiological source
term used in the determination of integrated doses for environmental qualification purposes is
superseded by the corresponding guidance in this regulatory guide for those facilities that are
proposing to, or have already, implemented an AST. All other guidance in Regulatory Guide 1.89
remains effective.
This guide primarily addresses design basis accidents, such as those addressed in Chapter 15
of typical final safety analysis reports (FSARs). This guide does not address all areas of potentially
significant risk. Although this guide addresses fuel handling accidents, other events that could occur
during shutdown operations are not currently addressed. The NRC staff has several ongoing
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initiatives involving risks of shutdown operations, extended burnup fuels, and risk-informing current
regulations. The information in this guide may be revised in the future as NRC staff evaluations are
completed and regulatory decisions on these issues are made.
C. REGULATORY POSITION
1. IMPLEMENTATION OF AST
1.1 Generic Considerations
As used in this guide, an AST is an accident source term that is different from the accident
source term used in the original design and licensing of the facility and that has been approved for use
under 10 CFR 50.67. This guide identifies an AST that is acceptable to the NRC staff and identifies
significant characteristics of other ASTs that may be found acceptable. While the NRC staff
recognizes several potential uses of an AST, it is not possible to foresee all possible uses. The NRC
staff will allow licensees to pursue technically justifiable uses of the ASTs in the most flexible
manner compatible with maintaining a clear, logical, and consistent design basis. The NRC staff will
approve these license amendment requests if the facility, as modified, will continue to provide
sufficient safety margins with adequate defense in depth to address unanticipated events and to
compensate for uncertainties in accident progression and analysis assumptions and parameter inputs.
1.1.1 Safety Margins
The proposed uses of an AST and the associated proposed facility modifications and changes
to procedures should be evaluated to determine whether the proposed changes are consistent with the
principle that sufficient safety margins are maintained, including a margin to account for analysis
uncertainties. The safety margins are products of specific values and limits contained in the technical
specifications (which cannot be changed without NRC approval) and other values, such as assumed
accident or transient initial conditions or assumed safety system response times. Changes, or the net
effects of multiple changes, that result in a reduction in safety margins may require prior NRC
approval. Once the initial AST implementation has been approved by the staff and has become part
of the facility design basis, the licensee may use 10 CFR 50.59 and its supporting guidance in
assessing safety margins related to subsequent facility modifications and changes to procedures.
1.1.2 Defense in Depth
The proposed uses of an AST and the associated proposed facility modifications and changes
to procedures should be evaluated to determine whether the proposed changes are consistent with the
principle that adequate defense in depth is maintained to compensate for uncertainties in accident
progression and analysis data. Consistency with the defense-in-depth philosophy is maintained if
system redundancy, independence, and diversity are preserved commensurate with the expected
frequency, consequences of challenges to the system, and uncertainties. In all cases, compliance with
the General Design Criteria in Appendix A to 10 CFR Part 50 is essential. Modifications proposed
for the facility generally should not create a need for compensatory programmatic activities, such as
reliance on manual operator actions.
Proposed modifications that seek to downgrade or remove required engineered safeguards
equipment should be evaluated to be sure that the modification does not invalidate assumptions made
in facility PRAs and does not adversely impact the facility’s severe accident management program.
4 This planning basis is also addressed in NUREG-0654, “Criteria for Preparation and Evaluation of Radiological Emergency
Response Plans and Preparedness in Support of Nuclear Power Plants” (Ref. 13).
1.183-5
1.1.3 Integrity of Facility Design Basis
The design basis accident source term is a fundamental assumption upon which a significant
portion of the facility design is based. Additionally, many aspects of facility operation derive from
the design analyses that incorporated the earlier accident source term. Although a complete reassessment of all facility radiological analyses would be desirable, the NRC staff determined that
recalculation of all design analyses would generally not be necessary. Regulatory Position 1.3 of this
guide provides guidance on which analyses need updating as part of the AST implementation
submittal and which may need updating in the future as additional modifications are performed.
This approach would create two tiers of analyses, those based on the previous source term and
those based on an AST. The radiological acceptance criteria would also be different with some
analyses based on whole body and thyroid criteria and some based on TEDE criteria. Full
implementation of the AST revises the plant licensing basis to specify the AST in place of the
previous accident source term and establishes the TEDE dose as the new acceptance criteria.
Selective implementation of the AST also revises the plant licensing basis and may establish the
TEDE dose as the new acceptance criteria. Selective implementation differs from full
implementation only in the scope of the change. In either case, the facility design bases should clearly
indicate that the source term assumptions and radiological criteria in these affected analyses have
been superseded and that future revisions of these analyses, if any, will use the updated approved
assumptions and criteria.
Radiological analyses generally should be based on assumptions and inputs that are consistent
with corresponding data used in other design basis safety analyses, radiological and nonradiological,
unless these data would result in nonconservative results or otherwise conflict with the guidance in
this guide.
1.1.4 Emergency Preparedness Applications
Requirements for emergency preparedness at nuclear power plants are set forth in 10 CFR
50.47, “Emergency Plans.” Additional requirements are set forth in Appendix E, “Emergency
Planning and Preparedness for Production and Utilization Facilities,” to 10 CFR Part 50. The
planning basis for many of these requirements was published in NUREG-0396, “Planning Basis for
the Development of State and Local Government Radiological Emergency Response Plans in Support
of Light Water Nuclear Power Plants”4 (Ref. 12). This joint effort by the Environmental Protection
Agency (EPA) and the NRC considered the principal characteristics (such as nuclides released and
distances) likely to be involved for a spectrum of design basis and severe (core melt) accidents. No
single accident scenario is the basis of the required preparedness. The objective of the planning is to
provide public protection that would encompass a wide spectrum of possible events with a sufficient
basis for extension of response efforts for unanticipated events. These requirements were issued after
a long period of involvement by numerous stakeholders, including the Federal Emergency
Management Agency, other Federal agencies, local and State governments (and in some cases, foreign
governments), private citizens, utilities, and industry groups.
Although the AST provided in this guide was based on a limited spectrum of severe accidents,
the particular characteristics have been tailored specifically for DBA analysis use. The AST is not
1.183-6
representative of the wide spectrum of possible events that make up the planning basis of emergency
preparedness. Therefore, the AST is insufficient by itself as a basis for requesting relief from the
emergency preparedness requirements of 10 CFR 50.47 and Appendix E to 10 CFR Part 50.
This guidance does not, however, preclude the appropriate use of the insights of the AST in
establishing emergency response procedures such as those associated with emergency dose
projections, protective measures, and severe accident management guides.
1.2 Scope of Implementation
The AST described in this guide is characterized by radionuclide composition and magnitude,
chemical and physical form of the radionuclides, and the timing of the release of these radionuclides.
The accident source term is a fundamental assumption upon which a large portion of the facility
design is based. Additionally, many aspects of facility operation derive from the design analyses that
incorporated the earlier accident source term. A complete implementation of an AST would upgrade
all existing radiological analyses and would consider the impact of all five characteristics of the AST
as defined in 10 CFR 50.2. However, the NRC staff has determined that there could be
implementations for which this level of re-analysis may not be necessary. Two categories are
defined: Full and selective implementations.
1.2.1 Full Implementation
Full implementation is a modification of the facility design basis that addresses all
characteristics of the AST, that is, composition and magnitude of the radioactive material, its
chemical and physical form, and the timing of its release. Full implementation revises the plant
licensing basis to specify the AST in place of the previous accident source term and establishes the
TEDE dose as the new acceptance criteria. This applies not only to the analyses performed in the
application (which may only include a subset of the plant analyses), but also to all future design basis
analyses. At a minimum for full implementations, the DBA LOCA must be re-analyzed using the
guidance in Appendix A of this guide. Additional guidance on analysis is provided in Regulatory
Position 1.3 of this guide. Since the AST and TEDE criteria would become part of the facility design
basis, new applications of the AST would not require prior NRC approval unless stipulated by 10
CFR 50.59, “Changes, Tests, and Experiments,” or unless the new application involved a change to a
technical specification. However, a change from an approved AST to a different AST that is not
approved for use at that facility would require a license amendment under 10 CFR 50.67.
1.2.2 Selective Implementation
Selective implementation is a modification of the facility design basis that (1) is based on one
or more of the characteristics of the AST or (2) entails re-evaluation of a limited subset of the design
basis radiological analyses. The NRC staff will allow licensees flexibility in technically justified
selective implementations provided a clear, logical, and consistent design basis is maintained. An
example of an application of selective implementation would be one in which a licensee desires to use
the release timing insights of the AST to increase the required closure time for a containment isolation
valve by a small amount. Another example would be a request to remove the charcoal filter media
from the spent fuel building ventilation exhaust. For the latter, the licensee may only need to reanalyze DBAs that credited the iodine removal by the charcoal media. Additional analysis guidance
is provided in Regulatory Position 1.3 of this guide. NRC approval for the AST (and the TEDE dose
criterion) will be limited to the particular selective implementation proposed by the licensee. The
5 Dose guidelines of 10 CFR 100.11 are superseded by 10 CFR 50.67 for licensees that have implemented an AST.
1.183-7
licensee would be able to make subsequent modifications to the facility and changes to procedures
based on the selected AST characteristics incorporated into the design basis under the provisions of
10 CFR 50.59. However, use of other characteristics of an AST or use of TEDE criteria that are not
part of the approved design basis, and changes to previously approved AST characteristics, would
require prior staff approval under 10 CFR 50.67. As an example, a licensee with an implementation
involving only timing, such as relaxed closure time on isolation valves, could not use 10 CFR 50.59
as a mechanism to implement a modification involving a reanalysis of the DBA LOCA. However,
this licensee could extend use of the timing characteristic to adjust the closure time on isolation
valves not included in the original approval.
1.3 Scope of Required Analyses
1.3.1 Design Basis Radiological Analyses
There are several regulatory requirements for which compliance is demonstrated, in part, by
the evaluation of the radiological consequences of design basis accidents. These requirements
include, but are not limited to, the following.
� Environmental Qualification of Equipment (10 CFR 50.49)
� Control Room Habitability (GDC-19 of Appendix A to 10 CFR Part 50)
� Emergency Response Facility Habitability (Paragraph IV.E.8 of Appendix E to 10
� Alternative Source Term (10 CFR 50.67)
� Environmental Reports (10 CFR Part 51)
� Facility Siting (10 CFR 100.11)5
There may be additional applications of the accident source term identified in the technical
specification bases and in various licensee commitments. These include, but are not limited to, the
following from Reference 2, NUREG-0737.
� Post-Accident Access Shielding (NUREG-0737, II.B.2)
� Post-Accident Sampling Capability (NUREG-0737, II.B.3)
� Accident Monitoring Instrumentation (NUREG-0737, II.F.1)
� Leakage Control (NUREG-0737, III.D.1.1)
� Emergency Response Facilities (NUREG-0737, III.A.1.2)
� Control Room Habitability (NUREG-0737, III.D.3.4)
1.3.2 Re-Analysis Guidance
Any implementation of an AST, full or selective, and any associated facility modification
should be supported by evaluations of all significant radiological and nonradiological impacts of
the proposed actions. This evaluation should consider the impact of the proposed changes on the
facility’s compliance with the regulations and commitments listed above as well as any other
facility-specific requirements. These impacts may be due to (1) the associated facility
modifications or (2) the differences in the AST characteristics. The scope and extent of the re-
6 For example, a proposed modification to change the timing of a containment isolation valve from 2.5 seconds to 5.0 seconds
might be acceptable without any dose calculations. However, a proposed modification that would delay containment spray
actuation could involve recalculation of DBA LOCA doses, re-assessment of the containment pressure and temperature transient,
recalculation of sump pH, re-assessment of the emergency diesel generator loading sequence, integrated doses to equipment in
the containment, and more.
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evaluation will necessarily be a function of the specific proposed facility modification6 and
whether a full or selective implementation is being pursued. The NRC staff does not expect a
complete recalculation of all facility radiological analyses, but does expect licensees to evaluate all
impacts of the proposed changes and to update the affected analyses and the design bases
appropriately. An analysis is considered to be affected if the proposed modification changes one
or more assumptions or inputs used in that analysis such that the results, or the conclusions drawn
on those results, are no longer valid. Generic analyses, such as those performed by owner groups
or vendor topical reports, may be used provided the licensee justifies the applicability of the
generic conclusions to the specific facility and implementation. Sensitivity analyses, discussed
below, may also be an option. If affected design basis analyses are to be re-calculated, all affected
assumptions and inputs should be updated and all selected characteristics of the AST and the
TEDE criteria should be addressed. The license amendment request should describe the licensee’s
re-analysis effort and provide statements regarding the acceptability of the proposed
implementation, including modifications, against each of the applicable analysis requirements and
commitments identified in Regulatory Position 1.3.1 of this guide.
The NRC staff has performed an evaluation of the impact of the AST on three
representative operating reactors (Ref. 14). This evaluation determined that radiological analysis
results based on the TID-14844 source term assumptions (Ref. 1) and the whole body and thyroid
methodology generally bound the results from analyses based on the AST and TEDE methodology.
Licensees may use the applicable conclusions of this evaluation in addressing the impact of the
AST on design basis radiological analyses. However, this does not exempt the licensee from
evaluating the remaining radiological and nonradiological impacts of the AST implementation and
the impacts of the associated plant modifications. For example, a selective implementation based
on the timing insights of the AST may change the required isolation time for the containment
purge dampers from 2.5 seconds to 5.0 seconds. This application might be acceptable without
dose calculations. However, evaluations may need to be performed regarding the ability of the
damper to close against increased containment pressure or the ability of ductwork downstream of
the dampers to withstand increased stresses.
For full implementation, a complete DBA LOCA analysis as described in Appendix A of
this guide should be performed, as a minimum. Other design basis analyses are updated in
accordance with the guidance in this section.
A selective implementation of an AST and any associated facility modification based on
the AST should evaluate all the radiological and nonradiological impacts of the proposed actions
as they apply to the particular implementation. Design basis analyses are updated in accordance
with the guidance in this section. There is no minimum requirement that a DBA LOCA analysis be
performed. The analyses performed need to address all impacts of the proposed modification, the
selected characteristics of the AST, and if dose calculations are performed, the TEDE criteria. For
selective implementations based on the timing characteristic of the AST, e.g., change in the closure
timing of a containment isolation valve, re-analysis of radiological calculations may not be
7 In performing screenings and evaluations pursuant to 10 CFR 50.59, it may be necessary to compare dose results expressed in
terms of whole body and thyroid with new results expressed in terms of TEDE. In these cases, the previous thyroid dose should
be multiplied by 0.03 and the product added to the whole body dose. The result is then compared to the TEDE result in the
screenings and evaluations. This change in dose methodology is not considered a change in the method of evaluation if the
licensee was previously authorized to use an AST and the TEDE criteria under 10 CFR 50.67.
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necessary if the modified elapsed time remains a fraction (e.g., 25%) of the time between accident
initiation and the onset of the gap release phase. Longer time delays may be considered on an
individual basis. For longer time delays, evaluation of the radiological consequences and other
impacts of the delay, such as blockage by debris in sump water, may be necessary. If affected
design basis analyses are to be re-calculated, all affected assumptions and inputs should be updated
and all selected characteristics of the AST and the TEDE criteria should be addressed.
1.3.3 Use of Sensitivity or Scoping Analyses
It may be possible to demonstrate by sensitivity or scoping evaluations that existing
analyses have sufficient margin and need not be recalculated. As used in this guide, a sensitivity
analysis is an evaluation that considers how the overall results vary as an input parameter (in this
case, AST characteristics) is varied. A scoping analysis is a brief evaluation that uses
conservative, simple methods to show that the results of the analysis bound those obtainable from
a more complete treatment. Sensitivity analyses are particularly applicable to suites of calculations
that address diverse components or plant areas but are otherwise largely based on generic
assumptions and inputs. Such cases might include postaccident vital area access dose calculations,
shielding calculations, and equipment environmental qualification (integrated dose). It may be
possible to identify a bounding case, re-analyze that case, and use the results to draw conclusions
regarding the remainder of the analyses. It may also be possible to show that for some analyses the
whole body and thyroid doses determined with the previous source term would bound the TEDE
obtained using the AST. Where present, arbitrary “designer margins” may be adequate to bound
any impact of the AST and TEDE criteria. If sensitivity or scoping analyses are used, the license
amendment request should include a discussion of the analyses performed and the conclusions
drawn. Scoping or sensitivity analyses should not constitute a significant part of the evaluations
for the design basis exclusion area boundary (EAB), low population zone (LPZ), or control room
dose.
1.3.4 Updating Analyses Following Implementation
Full implementation of the AST replaces the previous accident source term with the
approved AST and the TEDE criteria for all design basis radiological analyses. The
implementation may have been supported in part by sensitivity or scoping analyses that concluded
many of the design basis radiological analyses would remain bounding for the AST and the TEDE
criteria and would not require updating. After the implementation is complete, there may be a
subsequent need (e.g., a planned facility modification) to revise these analyses or to perform new
analyses. For these recalculations, the NRC staff expects that all characteristics of the AST and the
TEDE criteria incorporated into the design basis will be addressed in all affected analyses on an
individual as-needed basis. Re-evaluation using the previously approved source term may not be
appropriate. Since the AST and the TEDE criteria are part of the approved design basis for the
facility, use of the AST and TEDE criteria in new applications at the facility do not constitute a
change in analysis methodology that would require NRC approval.7
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This guidance is also applicable to selective implementations to the extent that the affected
analyses are within the scope of the approved implementation as described in the facility design
basis. In these cases, the characteristics of the AST and TEDE criteria identified in the facility
design basis need to be considered in updating the analyses. Use of other characteristics of the
AST or TEDE criteria that are not part of the approved design basis, and changes to previously
approved AST characteristics, requires prior NRC staff approval under 10 CFR 50.67.
1.3.5 Equipment Environmental Qualification
Current environmental qualification (EQ) analyses may be impacted by a proposed plant
modification associated with the AST implementation. The EQ analyses that have assumptions or
inputs affected by the plant modification should be updated to address these impacts. The NRC
staff is assessing the effect of increased cesium releases on EQ doses to determine whether
licensee action is warranted. Until such time as this generic issue is resolved, licensees may use
either the AST or the TID14844 assumptions for performing the required EQ analyses. However,
no plant modifications are required to address the impact of the difference in source term
characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the evaluation of the
generic issue. The EQ dose estimates should be calculated using the design basis survivability
period.
1.4 Risk Implications
The use of an AST changes only the regulatory assumptions regarding the analytical
treatment of the design basis accidents. The AST has no direct effect on the probability of the
accident. Use of an AST alone cannot increase the core damage frequency (CDF) or the large early
release frequency (LERF). However, facility modifications made possible by the AST could have
an impact on risk. If the proposed implementation of the AST involves changes to the facility
design that would invalidate assumptions made in the facility’s PRA, the impact on the existing
PRAs should be evaluated.
Consideration should be given to the risk impact of proposed implementations that seek to
remove or downgrade the performance of previously required engineered safeguards equipment on
the basis of the reduced postulated doses. The NRC staff may request risk information if there is a
reason to question adequate protection of public health and safety.
The licensee may elect to use risk insights in support of proposed changes to the design
basis that are not addressed in currently approved NRC staff positions. For guidance, refer to
Regulatory Guide 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to the Licensing Basis” (Ref. 15).
1.5 Submittal Requirements
According to 10 CFR 50.90, an application for an amendment must fully describe the
changes desired and should follow, as far as applicable, the form prescribed for original
applications. Regulatory Guide 1.70, “Standard Format and Content of Safety Analysis Reports
for Nuclear Power Plants (LWR Edition)” (Ref 16), provides additional guidance. The NRC
staff’s finding that the amendment may be approved must be based on the licensee’s analyses,
1.183-11
since it is these analyses that will become part of the design basis of the facility. The amendment
request should describe the licensee’s analyses of the radiological and nonradiological impacts of
the proposed modification in sufficient detail to support review by the NRC staff. The staff
recommends that licensees submit affected FSAR pages annotated with changes that reflect the
revised analyses or submit the actual calculation documentation.
If the licensee has used a current approved version of an NRC-sponsored computer code,
the NRC staff review can be made more efficient if the licensee identifies the code used and
submits the inputs that the licensee used in the calculations made with that code. In many cases,
this will reduce the need for NRC staff confirmatory analyses. This recommendation does not
constitute a requirement that the licensee use NRC-sponsored computer codes.
1.6 FSAR Requirements
Requirements for updating the facility’s final safety analysis report (FSAR) are in 10 CFR
50.71, “Maintenance of Records, Making of Reports.” The regulations in 10 CFR 50.71(e) require
that the FSAR be updated to include all changes made in the facility or procedures described in the
FSAR and all safety evaluations performed by the licensee in support of requests for license
amendments or in support of conclusions that changes did not involve unreviewed safety
questions. The analyses required by 10 CFR 50.67 are subject to this requirement. The affected
radiological analysis descriptions in the FSAR should be updated to reflect the replacement of the
design basis source term by the AST. The analysis descriptions should contain sufficient detail to
identify the methodologies used, significant assumptions and inputs, and numeric results.
Regulatory Guide 1.70 (Ref. 16) provides additional guidance. The descriptions of superseded
analyses should be removed from the FSAR in the interest of maintaining a clear design basis.
2. ATTRIBUTES OF AN ACCEPTABLE AST
An acceptable AST is not set forth in 10 CFR 50.67. Regulatory Position 3 of this guide
identifies an AST that is acceptable to the NRC staff for use at operating power reactors. A
substantial effort was expended by the NRC, its contractors, various national laboratories, peer
reviewers, and others in performing severe accident research and in developing the source terms
provided in NUREG-1465 (Ref. 5). However, future research may identify opportunities for
changes in these source terms. The NRC staff will consider applications for an AST different from
that identified in this guide. However, the NRC staff does not expect to approve any source term
that is not of the same level of quality as the source terms in NUREG-1465. To be considered
acceptable, an AST must have the following attributes:
2.1 The AST must be based on major accidents, hypothesized for the purposes of design
analyses or consideration of possible accidental events, that could result in hazards not
exceeded by those from other accidents considered credible. The AST must address events
that involve a substantial meltdown of the core with the subsequent release of appreciable
quantities of fission products.
8 The uncertainty factor used in determining the core inventory should be that value provided in Appendix K to 10 CFR Part 50,
typically 1.02.
9 Note that for some radionuclides, such as Cs-137, equilibrium will not be reached prior to fuel offload. Thus, the maximum
inventory at the end of life should be used.
1.183-12
2.2 The AST must be expressed in terms of times and rates of appearance of radioactive fission
products released into containment, the types and quantities of the radioactive species
released, and the chemical forms of iodine released.
2.3 The AST must not be based upon a single accident scenario but instead must represent a
spectrum of credible severe accident events. Risk insights may be used, not to select a
single risk-significant accident, but rather to establish the range of events to be considered.
Relevant insights from applicable severe accident research on the phenomenology of
fission product release and transport behavior may be considered.
2.4 The AST must have a defensible technical basis supported by sufficient experimental and
empirical data, be verified and validated, and be documented in a scrutable form that
facilitates public review and discourse.
2.5 The AST must be peer-reviewed by appropriately qualified subject matter experts. The
peer-review comments and their resolution should be part of the documentation supporting
the AST.
3. ACCIDENT SOURCE TERM
This section provides an AST that is acceptable to the NRC staff. The data in Regulatory
Positions 3.2 through 3.5 are fundamental to the definition of an AST. Once approved, the AST
assumptions or parameters specified in these positions become part of the facility’s design basis.
Deviations from this guidance must be evaluated against Regulatory Position 2. After the NRC
staff has approved an implementation of an AST, subsequent changes to the AST will require NRC
staff review under 10 CFR 50.67.
3.1 Fission Product Inventory
The inventory of fission products in the reactor core and available for release to the
containment should be based on the maximum full power operation of the core with, as a
minimum, current licensed values for fuel enrichment, fuel burnup, and an assumed core power
equal to the current licensed rated thermal power times the ECCS evaluation uncertainty.8 The
period of irradiation should be of sufficient duration to allow the activity of dose-significant
radionuclides to reach equilibrium or to reach maximum values.9 The core inventory should be
determined using an appropriate isotope generation and depletion computer code such as ORIGEN
2 (Ref. 17) or ORIGEN-ARP (Ref. 18). Core inventory factors (Ci/MWt) provided in TID14844
and used in some analysis computer codes were derived for low burnup, low enrichment fuel and
should not be used with higher burnup and higher enrichment fuels.
10 The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak
burnup up to 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel.
1.183-13
For the DBA LOCA, all fuel assemblies in the core are assumed to be affected and the core
average inventory should be used. For DBA events that do not involve the entire core, the fission
product inventory of each of the damaged fuel rods is determined by dividing the total core
inventory by the number of fuel rods in the core. To account for differences in power level across
the core, radial peaking factors from the facility’s core operating limits report (COLR) or technical
specifications should be applied in determining the inventory of the damaged rods.
No adjustment to the fission product inventory should be made for events postulated to
occur during power operations at less than full rated power or those postulated to occur at the
beginning of core life. For events postulated to occur while the facility is shutdown, e.g., a fuel
handling accident, radioactive decay from the time of shutdown may be modeled.
3.2 Release Fractions10
The core inventory release fractions, by radionuclide groups, for the gap release and early
in-vessel damage phases for DBA LOCAs are listed in Table 1 for BWRs and Table 2 for PWRs.
These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1.
For non-LOCA events, the fractions of the core inventory assumed to be in the gap for the
various radionuclides are given in Table 3. The release fractions from Table 3 are used in
conjunction with the fission product inventory calculated with the maximum core radial peaking
factor.
Table 1
BWR Core Inventory Fraction
Released Into Containment
Gap Early
Release In-vessel
Group Phase Phase Total
Noble Gases 0.05 0.95 1.0
Halogens 0.05 0.25 0.3
Alkali Metals 0.05 0.20 0.25
Tellurium Metals 0.00 0.05 0.05
Ba, Sr 0.00 0.02 0.02
Noble Metals 0.00 0.0025 0.0025
Cerium Group 0.00 0.0005 0.0005
Lanthanides 0.00 0.0002 0.0002
11 The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peak
burnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rod
average power for burnups exceeding 54 GWD/MTU. As an alternative, fission gas release calculations performed using NRCapproved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected
power history that will bound the limiting projected plant-specific power history for the specific fuel load. For the BWR rod
drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.
12 In lieu of treating the release in a linear ramp manner, the activity for each phase can be modeled as being released
instantaneously at the start of that release phase, i.e., in step increases.
1.183-14
Table 2
PWR Core Inventory Fraction
Released Into Containment
Gap Early
Release In-vessel
Group Phase Phase Total
Noble Gases 0.05 0.95 1.0
Halogens 0.05 0.35 0.4
Alkali Metals 0.05 0.25 0.3
Tellurium Metals 0.00 0.05 0.05
Ba, Sr 0.00 0.02 0.02
Noble Metals 0.00 0.0025 0.0025
Cerium Group 0.00 0.0005 0.0005
Lanthanides 0.00 0.0002 0.0002
Table 311
Non-LOCA Fraction of Fission Product Inventory in Gap
Group Fraction
I-131 0.08
Kr-85 0.10
Other Noble Gases 0.05
Other Halogens 0.05
Alkali Metals 0.12
3.3 Timing of Release Phases
Table 4 tabulates the onset and duration of each sequential release phase for DBA LOCAs
at PWRs and BWRs. The specified onset is the time following the initiation of the accident (i.e.,
time = 0). The early in-vessel phase immediately follows the gap release phase. The activity
released from the core during each release phase should be modeled as increasing in a linear
fashion over the duration of the phase.12 For non-LOCA DBAs in which fuel damage is projected,
the release from the fuel gap and the fuel pellet should be assumed to occur instantaneously with
the onset of the projected damage.
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Table 4
LOCA Release Phases
Phase Onset Duration Onset Duration
Gap Release 30 sec 0.5 hr 2 min 0.5 hr
Early In-Vessel 0.5 hr 1.3 hr 0.5 hr 1.5 hr
For facilities licensed with leak-before-break methodology, the onset of the gap release
phase may be assumed to be 10 minutes. A licensee may propose an alternative time for the onset
of the gap release phase, based on facility-specific calculations using suitable analysis codes or on
an accepted topical report shown to be applicable to the specific facility. In the absence of
approved alternatives, the gap release phase onsets in Table 4 should be used.
3.4 Radionuclide Composition
Table 5 lists the elements in each radionuclide group that should be considered in design
basis analyses.
Table 5
Radionuclide Groups
Group Elements
Noble Gases Xe, Kr
Halogens I, Br
Alkali Metals Cs, Rb
Tellurium Group Te, Sb, Se, Ba, Sr
Noble Metals Ru, Rh, Pd, Mo, Tc, Co
Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr
Sm, Y, Cm, Am
Cerium Ce, Pu, Np
3.5 Chemical Form
Of the radioiodine released from the reactor coolant system (RCS) to the containment in a
postulated accident, 95 percent of the iodine released should be assumed to be cesium iodide (CsI),
4.85 percent elemental iodine, and 0.15 percent organic iodide. This includes releases from the
gap and the fuel pellets. With the exception of elemental and organic iodine and noble gases,
fission products should be assumed to be in particulate form. The same chemical form is assumed
in releases from fuel pins in FHAs and from releases from the fuel pins through the RCS in DBAs
other than FHAs or LOCAs. However, the transport of these iodine species following release from
the fuel may affect these assumed fractions. The accident-specific appendices to this regulatory
guide provide additional details.
13 The prior practice of basing inhalation exposure on only radioiodine and not including radioiodine in external exposure
calculations is not consistent with the definition of TEDE and the characteristics of the revised source term.
1.183-16
3.6 Fuel Damage in Non-LOCA DBAs
The amount of fuel damage caused by non-LOCA design basis events should be analyzed
to determine, for the case resulting in the highest radioactivity release, the fraction of the fuel that
reaches or exceeds the initiation temperature of fuel melt and the fraction of fuel elements for
which the fuel clad is breached. Although the NRC staff has traditionally relied upon the departure
from nucleate boiling ratio (DNBR) as a fuel damage criterion, licensees may propose other
methods to the NRC staff, such as those based upon enthalpy deposition, for estimating fuel
damage for the purpose of establishing radioactivity releases.
The amount of fuel damage caused by a FHA is addressed in Appendix B of this guide.
4. DOSE CALCULATIONAL METHODOLOGY
The NRC staff has determined that there is an implied synergy between the ASTs and total
effective dose equivalent (TEDE) criteria, and between the TID-14844 source terms and the whole
body and thyroid dose criteria, and therefore, they do not expect to allow the TEDE criteria to be
used with TID-14844 calculated results. The guidance of this section applies to all dose
calculations performed with an AST pursuant to 10 CFR 50.67. Certain selective implementations
may not require dose calculations as described in Regulatory Position 1.3 of this guide.
4.1 Offsite Dose Consequences
The following assumptions should be used in determining the TEDE for persons located at
or beyond the boundary of the exclusion area (EAB):
4.1.1 The dose calculations should determine the TEDE. TEDE is the sum of the
committed effective dose equivalent (CEDE) from inhalation and the deep dose equivalent (DDE)
from external exposure. The calculation of these two components of the TEDE should consider all
radionuclides, including progeny from the decay of parent radionuclides, that are significant with
regard to dose consequences and the released radioactivity.13
4.1.2 The exposure-to-CEDE factors for inhalation of radioactive material should be
derived from the data provided in ICRP Publication 30, “Limits for Intakes of Radionuclides by
Workers” (Ref. 19). Table 2.1 of Federal Guidance Report 11, “Limiting Values of Radionuclide
Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and
Ingestion” (Ref. 20), provides tables of conversion factors acceptable to the NRC staff. The
factors in the column headed “effective” yield doses corresponding to the CEDE.
4.1.3 For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be assumed to be
3.5 x 10-4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rate
should be assumed to be 1.8 x 10-4 cubic meters per second. After that and until the end of the
accident, the rate should be assumed to be 2.3 x 10-4 cubic meters per second.
14 With regard to the EAB TEDE, the maximum two-hour value is the basis for screening and evaluation under 10 CFR 50.59.
Changes to doses outside of the two-hour window are only considered in the context of their impact on the maximum two-hour
EAB TEDE.
1.183-17
4.1.4 The DDE should be calculated assuming submergence in semi-infinite cloud
assumptions with appropriate credit for attenuation by body tissue. The DDE is nominally
equivalent to the effective dose equivalent (EDE) from external exposure if the whole body is
irradiated uniformly. Since this is a reasonable assumption for submergence exposure situations,
EDE may be used in lieu of DDE in determining the contribution of external dose to the TEDE.
Table III.1 of Federal Guidance Report 12, “External Exposure to Radionuclides in Air, Water, and
Soil” (Ref. 21), provides external EDE conversion factors acceptable to the NRC staff. The factors
in the column headed “effective” yield doses corresponding to the EDE.
4.1.5 The TEDE should be determined for the most limiting person at the EAB. The
maximum EAB TEDE for any two-hour period following the start of the radioactivity release
should be determined and used in determining compliance with the dose criteria in 10 CFR
50.67.14 The maximum two-hour TEDE should be determined by calculating the postulated dose
for a series of small time increments and performing a “sliding” sum over the increments for
successive two-hour periods. The maximum TEDE obtained is submitted. The time increments
should appropriately reflect the progression of the accident to capture the peak dose interval
between the start of the event and the end of radioactivity release (see also Table 6).
4.1.6 TEDE should be determined for the most limiting receptor at the outer boundary of
the low population zone (LPZ) and should be used in determining compliance with the dose
criteria in 10 CFR 50.67.
4.1.7 No correction should be made for depletion of the effluent plume by deposition on
the ground.
4.2 Control Room Dose Consequences
The following guidance should be used in determining the TEDE for persons located in the
control room:
4.2.1 The TEDE analysis should consider all sources of radiation that will cause exposure
to control room personnel. The applicable sources will vary from facility to facility, but typically
will include:
� Contamination of the control room atmosphere by the intake or infiltration of the
radioactive material contained in the radioactive plume released from the facility,
� Contamination of the control room atmosphere by the intake or infiltration of
airborne radioactive material from areas and structures adjacent to the control room
envelope,
� Radiation shine from the external radioactive plume released from the facility,
15 The iodine protection factor (IPF) methodology of Reference 22 may not be adequately conservative for all DBAs and control
room arrangements since it models a steady-state control room condition. Since many analysis parameters change over the
duration of the event, the IPF methodology should only be used with caution. The NRC computer codes HABIT (Ref. 23) and
RADTRAD (Ref. 24) incorporate suitable methodologies.
16 This occupancy is modeled in the χ/Q values determined in Reference 22 and should not be credited twice. The ARCON96
Code (Ref. 26) does not incorporate these occupancy assumptions, making it necessary to apply this correction in the dose
calculations.
1.183-18
� Radiation shine from radioactive material in the reactor containment,
� Radiation shine from radioactive material in systems and components inside or
external to the control room envelope, e.g., radioactive material buildup in
recirculation filters.
4.2.2 The radioactive material releases and radiation levels used in the control room dose
analysis should be determined using the same source term, transport, and release assumptions used
for determining the EAB and the LPZ TEDE values, unless these assumptions would result in nonconservative results for the control room.
4.2.3 The models used to transport radioactive material into and through the control
room,15 and the shielding models used to determine radiation dose rates from external sources,
should be structured to provide suitably conservative estimates of the exposure to control room
personnel.
4.2.4 Credit for engineered safety features that mitigate airborne radioactive material
within the control room may be assumed. Such features may include control room isolation or
pressurization, or intake or recirculation filtration. Refer to Section 6.5.1, “ESF Atmospheric
Cleanup System,” of the SRP (Ref. 3) and Regulatory Guide 1.52, “Design, Testing, and
Maintenance Criteria for Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air
Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants” (Ref. 25), for
guidance. The control room design is often optimized for the DBA LOCA and the protection
afforded for other accident sequences may not be as advantageous. In most designs, control room
isolation is actuated by engineered safeguards feature (ESF) signals or radiation monitors (RMs).
In some cases, the ESF signal is effective only for selected accidents, placing reliance on the RMs
for the remaining accidents. Several aspects of RMs can delay the control room isolation,
including the delay for activity to build up to concentrations equivalent to the alarm setpoint and
the effects of different radionuclide accident isotopic mixes on monitor response.
4.2.5 Credit should generally not be taken for the use of personal protective equipment or
prophylactic drugs. Deviations may be considered on a case-by-case basis.
4.2.6 The dose receptor for these analyses is the hypothetical maximum exposed
individual who is present in the control room for 100% of the time during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after
the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days.16
For the duration of the event, the breathing rate of this individual should be assumed to be 3.5 x 10-
4 cubic meters per second.
1.183-19
4.2.7 Control room doses should be calculated using dose conversion factors identified in
Regulatory Position 4.1 above for use in offsite dose analyses. The DDE from photons may be
corrected for the difference between finite cloud geometry in the control room and the semiinfinite cloud assumption used in calculating the dose conversion factors. The following
expression may be used to correct the semi-infinite cloud dose, DDE�, to a finite cloud dose,
DDEfinite, where the control room is modeled as a hemisphere that has a volume, V, in cubic feet,
equivalent to that of the control room (Ref. 22).
finite = ∞ 0 338
1173
.
4.3 Other Dose Consequences
The guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, in
re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in
NUREG-0737 (Ref. 2). Design envelope source terms provided in NUREG-0737 should be
updated for consistency with the AST. In general, radiation exposures to plant personnel identified
in Regulatory Position 1.3.1 should be expressed in terms of TEDE. Integrated radiation exposure
of plant equipment should be determined using the guidance of Appendix I of this guide.
4.4 Acceptance Criteria
The radiological criteria for the EAB, the outer boundary of the LPZ, and for the control
room are in 10 CFR 50.67. These criteria are stated for evaluating reactor accidents of exceedingly
low probability of occurrence and low risk of public exposure to radiation, e.g., a large-break
LOCA. The control room criterion applies to all accidents. For events with a higher probability of
occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 6.
The acceptance criteria for the various NUREG-0737 (Ref. 2) items generally reference
General Design Criteria 19 (GDC 19) from Appendix A to 10 CFR Part 50 or specify criteria
derived from GDC-19. These criteria are generally specified in terms of whole body dose, or its
equivalent to any body organ. For facilities applying for, or having received, approval for the use
of an AST, the applicable criteria should be updated for consistency with the TEDE criterion in 10
17 For PWRs with steam generator alternative repair criteria, different dose criteria may apply to steam generator tube rupture
and main steam line break analyses.
1.183-20
Table 617
Accident Dose Criteria
Accident or Case
EAB and LPZ
Dose Criteria Analysis Release Duration
LOCA 25 rem TEDE 30 days for containment, ECCS, and
BWR Main Steam Line Break Instantaneous puff
Fuel Damage or Pre-incident Spike 25 rem TEDE
Equilibrium Iodine Activity 2.5 rem TEDE
BWR Rod Drop Accident 6.3 rem TEDE 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
PWR Steam Generator Tube Rupture Affected SG: time to isolate; Unaffected
SG(s): until cold shutdown is established Fuel Damage or Pre-incident Spike 25 rem TEDE
Coincident Iodine Spike 2.5 rem TEDE
PWR Main Steam Line Break Until cold shutdown is established
Fuel Damage or Pre-incident Spike 25 rem TEDE
Coincident Iodine Spike 2.5 rem TEDE
PWR Locked Rotor Accident 2.5 rem TEDE Until cold shutdown is established
PWR Rod Ejection Accident 6.3 rem TEDE 30 days for containment pathway; until
cold shutdown is established for
secondary pathway
Fuel Handling Accident 6.3 rem TEDE 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
The column labeled “Analysis Release Duration” is a summary of the assumed
radioactivity release durations identified in the individual appendices to this guide. Refer to these
appendices for complete descriptions of the release pathways and durations.
5. ANALYSIS ASSUMPTIONS AND METHODOLOGY
5.1 General Considerations
5.1.1 Analysis Quality
The evaluations required by 10 CFR 50.67 are re-analyses of the design basis safety
analyses and evaluations required by 10 CFR 50.34; they are considered to be a significant input to
the evaluations required by 10 CFR 50.92 or 10 CFR 50.59. These analyses should be prepared,
reviewed, and maintained in accordance with quality assurance programs that comply with
Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,”
to 10 CFR Part 50.
These design basis analyses were structured to provide a conservative set of assumptions to
test the performance of one or more aspects of the facility design. Many physical processes and
phenomena are represented by conservative, bounding assumptions rather than being modeled
18 Note that for some parameters, the technical specification value may be adjusted for analysis purposes by factors provided in
other regulatory guidance. For example, ESF filter efficiencies are based on the guidance in Regulatory Guide 1.52 (Ref. 25) and
in Generic Letter 99-02 (Ref. 27) rather than the surveillance test criteria in the technical specifications. Generally, these
adjustments address potential changes in the parameter between scheduled surveillance tests.
1.183-21
directly. The staff has selected assumptions and models that provide an appropriate and prudent
safety margin against unpredicted events in the course of an accident and compensate for large
uncertainties in facility parameters, accident progression, radioactive material transport, and
atmospheric dispersion. Licensees should exercise caution in proposing deviations based upon
data from a specific accident sequence since the DBAs were never intended to represent any
specific accident sequence -- the proposed deviation may not be conservative for other accident
sequences.
5.1.2 Credit for Engineered Safeguard Features
Credit may be taken for accident mitigation features that are classified as safety-related, are
required to be operable by technical specifications, are powered by emergency power sources, and
are either automatically actuated or, in limited cases, have actuation requirements explicitly
addressed in emergency operating procedures. The single active component failure that results in
the most limiting radiological consequences should be assumed. Assumptions regarding the
occurrence and timing of a loss of offsite power should be selected with the objective of
maximizing the postulated radiological consequences.
5.1.3 Assignment of Numeric Input Values
The numeric values that are chosen as inputs to the analyses required by 10 CFR 50.67
should be selected with the objective of determining a conservative postulated dose. In some
instances, a particular parameter may be conservative in one portion of an analysis but be
nonconservative in another portion of the same analysis. For example, assuming minimum
containment system spray flow is usually conservative for estimating iodine scrubbing, but in
many cases may be nonconservative when determining sump pH. Sensitivity analyses may be
needed to determine the appropriate value to use. As a conservative alternative, the limiting value
applicable to each portion of the analysis may be used in the evaluation of that portion. A single
value may not be applicable for a parameter for the duration of the event, particularly for
parameters affected by changes in density. For parameters addressed by technical specifications,
the value used in the analysis should be that specified in the technical specifications.18 If a range of
values or a tolerance band is specified, the value that would result in a conservative postulated
dose should be used. If the parameter is based on the results of less frequent surveillance testing,
e.g., steam generator nondestructive testing (NDT), consideration should be given to the
degradation that may occur between periodic tests in establishing the analysis value.
5.1.4 Applicability of Prior Licensing Basis
The NRC staff considers the implementation of an AST to be a significant change to the
design basis of the facility that is voluntarily initiated by the licensee. In order to issue a license
amendment authorizing the use of an AST and the TEDE dose criteria, the NRC staff must make a
current finding of compliance with regulations applicable to the amendment. The characteristics
of the ASTs and the revised dose calculational methodology may be incompatible with many of the
analysis assumptions and methods currently reflected in the facility’s design basis analyses. The
NRC staff may find that new or unreviewed issues are created by a particular site-specific
1.183-22
implementation of the AST, warranting review of staff positions approved subsequent to the initial
issuance of the license. This is not considered a backfit as defined by 10 CFR 50.109,
“Backfitting.” However, prior design bases that are unrelated to the use of the AST, or are
unaffected by the AST, may continue as the facility’s design basis. Licensees should ensure that
analysis assumptions and methods are compatible with the ASTs and the TEDE criteria.
5.2 Accident-Specific Assumptions
The appendices to this regulatory guide provide accident-specific assumptions that are
acceptable to the staff for performing analyses that are required by 10 CFR 50.67. The DBAs
addressed in these attachments were selected from accidents that may involve damage to irradiated
fuel. This guide does not address DBAs with radiological consequences based on technical
specification reactor or secondary coolant-specific activities only. The inclusion or exclusion of a
particular DBA in this guide should not be interpreted as indicating that an analysis of that DBA is
required or not required. Licensees should analyze the DBAs that are affected by the specific
proposed applications of an AST.
The NRC staff has determined that the analysis assumptions in the appendices to this guide
provide an integrated approach to performing the individual analyses and generally expects
licensees to address each assumption or propose acceptable alternatives. Such alternatives may be
justifiable on the basis of plant-specific considerations, updated technical analyses, or, in some
cases, a previously approved licensing basis consideration. The assumptions in the appendices are
deemed consistent with the AST identified in Regulatory Position 3 and internally consistent with
each other. Although licensees are free to propose alternatives to these assumptions for
consideration by the NRC staff, licensees should avoid use of previously approved staff positions
that would adversely affect this consistency.
The NRC is committed to using probabilistic risk analysis (PRA) insights in its regulatory
activities and will consider licensee proposals for changes in analysis assumptions based upon risk
insights. The staff will not approve proposals that would reduce the defense in depth deemed
necessary to provide adequate protection for public health and safety. In some cases, this defense
in depth compensates for uncertainties in the PRA analyses and addresses accident considerations
not adequately addressed by the core damage frequency (CDF) and large early release frequency
(LERF) surrogate indicators of overall risk.
5.3 Meteorology Assumptions
Atmospheric dispersion values (χ/Q) for the EAB, the LPZ, and the control room that were
approved by the staff during initial facility licensing or in subsequent licensing proceedings may be
used in performing the radiological analyses identified by this guide. Methodologies that have
been used for determining χ/Q values are documented in Regulatory Guides 1.3 and 1.4,
Regulatory Guide 1.145, “Atmospheric Dispersion Models for Potential Accident Consequence
Assessments at Nuclear Power Plants,” and the paper, “Nuclear Power Plant Control Room
Ventilation System Design for Meeting General Criterion 19" (Refs. 6, 7, 22, and 28).
19 The ARCON96 computer code contains processing options that may yield χ/Q values that are not sufficiently conservative for
use in accident consequence assessments or may be incompatible with release point and ventilation intake configurations at
particular sites. The applicability of these options and associated input parameters should be evaluated on a case-by-case basis.
The assumptions made in the examples in the ARCON96 documentation are illustrative only and do not imply NRC staff
acceptance of the methods or data used in the example.
1.183-23
References 22 and 28 should be used if the FSAR χ/Q values are to be revised or if values
are to be determined for new release points or receptor distances. Fumigation should be
considered where applicable for the EAB and LPZ. For the EAB, the assumed fumigation period
should be timed to be included in the worst 2-hour exposure period. The NRC computer code
PAVAN (Ref. 29) implements Regulatory Guide 1.145 (Ref. 28) and its use is acceptable to the
NRC staff. The methodology of the NRC computer code ARCON9619 (Ref. 26) is generally
acceptable to the NRC staff for use in determining control room χ/Q values. Meteorological data
collected in accordance with the site-specific meteorological measurements program described in
the facility FSAR should be used in generating accident χ/ Q values. Additional guidance is
provided in Regulatory Guide 1.23, “Onsite Meteorological Programs” (Ref. 30). All changes in
ˇ/Q analysis methodology should be reviewed by the NRC staff.
6. ASSUMPTIONS FOR EVALUATING THE RADIATION DOSES FOR
EQUIPMENT QUALIFICATION
The assumptions in Appendix I to this guide are acceptable to the NRC staff for performing
radiological assessments associated with equipment qualification. The assumptions in Appendix I
will supersede Regulatory Positions 2.c(1) and 2.c(2) and Appendix D of Revision 1 of Regulatory
Guide 1.89, “Environmental Qualification of Certain Electric Equipment Important to Safety for
Nuclear Power Plants” (Ref. 11), for operating reactors that have amended their licensing basis to
use an alternative source term. Except as stated in Appendix I, all other assumptions, methods,
and provisions of Revision 1 of Regulatory Guide 1.89 remain effective.
The NRC staff is assessing the effect of increased cesium releases on EQ doses to
determine whether licensee action is warranted. Until such time as this generic issue is resolved,
licensees may use either the AST or the TID14844 assumptions for performing the required EQ
analyses. However, no plant modifications are required to address the impact of the difference in
source term characteristics (i.e., AST vs TID14844) on EQ doses pending the outcome of the
evaluation of the generic issue.
D. IMPLEMENTATION
The purpose of this section is to provide information to applicants and licensees regarding
the NRC staff’s plans for using this regulatory guide.
Except in those cases in which an applicant or licensee proposes an acceptable alternative
method for complying with the specified portions of the NRC’s regulations, the methods described
in this guide will be used in the evaluation of submittals related to the use of ASTs in radiological
consequence analyses at operating power reactors.
1.183-24
1.183-25
REFERENCES
{See the inside front cover of this guide for information on obtaining NRC documents.}
1. J.J. DiNunno et al., “Calculation of Distance Factors for Power and Test Reactor Sites,”
USAEC TID-14844, U.S. Atomic Energy Commission (now USNRC), 1962.
2. USNRC, “Clarification of TMI Action Plan Requirements,” NUREG-0737, November
1980.
3. USNRC, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear
Power Plants,” NUREG-0800, September 1981 (or updates of specific sections).
4. USNRC, “Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Final
Policy Statement,” Federal Register, Volume 60, page 42622 (60 FR 42622) August 16,
1995.
5. L. Soffer et al., “Accident Source Terms for Light-Water Nuclear Power Plants,”
NUREG-1465, USNRC, February 1995.
6. USNRC, “Assumptions Used for Evaluating the Potential Radiological Consequences of a
Loss of Coolant Accident for Boiling Water Reactors.” Regulatory Guide 1.3, Revision 2,
June 1974.
7. USNRC, “Assumptions Used for Evaluating the Potential Radiological Consequences of a
Loss of Coolant Accident for Pressurized Water Reactors,” Regulatory Guide 1.4, Revision
2, June 1974.
8. USNRC, “Assumptions Used for Evaluating the Potential Radiological Consequences of a
Steam Line Break Accident for Boiling Water Reactors,” Regulatory Guide 1.5, March
1971.
9. USNRC, “Assumptions Used for Evaluating the Potential Radiological Consequences of a
Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and
Pressurized Water Reactors,” Regulatory Guide 1.25, March 1972.
10. USNRC, “Assumptions Used for Evaluating a Control Rod Ejection Accident for
Pressurized Water Reactors,” Regulatory Guide 1.77, May 1974.
11. USNRC, “Environmental Qualification of Certain Electric Equipment Important to Safety
for Nuclear Power Plants,” Regulatory Guide 1.89, Revision 1, June 1984.
12. USNRC, “Planning Basis for the Development of State and Local Government
Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants,”
NUREG-0396, December 1978.
1.183-26
13. USNRC, “Criteria for Preparation and Evaluation of Radiological Emergency Response
Plans and Preparedness in Support of Nuclear Power Plants,” NUREG-0654, Revision 1
(FEMA-REP-1), November 1980.
14. USNRC, “Results of the Revised (NUREG-1465) Source Term Rebaselining for Operating
Reactors,” SECY-98-154, June 30, 1998.
15. USNRC, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed
Decisions on Plant-Specific Changes to the Licensing Basis,” Regulatory Guide 1.174, July
1998.
16. USNRC, “Standard Format and Content of Safety Analysis Reports for Nuclear Power
Plants (LWR Edition),” Regulatory Guide 1.70, Revision 3, November 1978.
17. A.G. Croff, “A User’s Manual for the ORIGEN 2 Computer Code,” ORNL/TM-7175, Oak
Ridge National Laboratory, July 1980.
18. S.M. Bowman and L.C. Leal, “The ORIGNARP Input Processor for ORIGEN-ARP,”
Appendix F7.A in SCALE: A Modular Code System for Performing Standardized Analyses
for Licensing Evaluation, NUREG/CR-0200, USNRC, March 1997.
19. ICRP, “Limits for Intakes of Radionuclides by Workers,” ICRP Publication 30, 1979.
20. K.F. Eckerman et al., “Limiting Values of Radionuclide Intake and Air Concentration and
Dose Conversion Factors for Inhalation, Submersion, and Ingestion,” Federal Guidance
Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988.
21. K.F. Eckerman and J.C. Ryman, “External Exposure to Radionuclides in Air, Water, and
Soil,” Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency,
1993.
22. K.G. Murphy and K.W. Campe, “Nuclear Power Plant Control Room Ventilation System
Design for Meeting General Criterion 19,” published in Proceedings of 13th AEC Air
Cleaning Conference, Atomic Energy Commission (now USNRC), August 1974.
23. USNRC, “Computer Codes for Evaluation of Control Room Habitability (HABIT V1.1),”
Supplement 1 to NUREG/CR-6210, November 1998.
24. S.L. Humphreys et al., “RADTRAD: A Simplified Model for Radionuclide Transport and
Removal and Dose Estimation,” NUREG/CR-6604, USNRC, April 1998.
25. USNRC, “Design, Testing, and Maintenance Criteria for Postaccident Engineered Safety
Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-WaterCooled Nuclear Power Plants,” Regulatory Guide 1.52, Revision 2, March 1978.
1.183-27
26. J.V. Ramsdell and C.A. Simonen, “Atmospheric Relative Concentrations in Building
Wakes, NUREG-6331, Revision 1, USNRC, May 1997.
27. USNRC, “Laboratory Testing of Nuclear-Grade Activated Charcoal,” NRC Generic Letter
99-02, June 3, 1999.
28. USNRC, “Atmospheric Dispersion Models for Potential Accident Consequence
Assessments at Nuclear Power Plants,” Regulatory Guide 1.145, Revision 1, November
1982.
29. T.J. Bander, “PAVAN: An Atmospheric Dispersion Program for Evaluating Design Basis
Accidental Releases of Radioactive Materials from Nuclear Power Stations,” NUREG2858, USNRC, November 1982.
30. USNRC, “Onsite Meteorological Programs,” Regulatory Guide 1.23, February 1972.
A-1
Appendix A
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES
OF A LWR LOSS-OF-COOLANT ACCIDENT
The assumptions in this appendix are acceptable to the NRC staff for evaluating the
radiological consequences of loss-of-coolant accidents (LOCAs) at light water reactors (LWRs).
These assumptions supplement the guidance provided in the main body of this guide.
Appendix A, “General Design Criteria for Nuclear Power Plants,” to 10 CFR Part 50
defines LOCAs as those postulated accidents that result from a loss of coolant inventory at rates
that exceed the capability of the reactor coolant makeup system. Leaks up to a double-ended
rupture of the largest pipe of the reactor coolant system are included. The LOCA, as with all
design basis accidents (DBAs), is a conservative surrogate accident that is intended to challenge
selective aspects of the facility design. Analyses are performed using a spectrum of break sizes to
evaluate fuel and ECCS performance. With regard to radiological consequences, a large-break
LOCA is assumed as the design basis case for evaluating the performance of release mitigation
systems and the containment and for evaluating the proposed siting of a facility.
SOURCE TERM ASSUMPTIONS
1. Acceptable assumptions regarding core inventory and the release of radionuclides from the
fuel are provided in Regulatory Position 3 of this guide.
2. If the sump or suppression pool pH is controlled at values of 7 or greater, the chemical
form of radioiodine released to the containment should be assumed to be 95% cesium iodide (CsI),
4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine species, including those
from iodine re-evolution, for sump or suppression pool pH values less than 7 will be evaluated on
a case-by-case basis. Evaluations of pH should consider the effect of acids and bases created
during the LOCA event, e.g., radiolysis products. With the exception of elemental and organic
iodine and noble gases, fission products should be assumed to be in particulate form.
ASSUMPTIONS ON TRANSPORT IN PRIMARY CONTAINMENT
3. Acceptable assumptions related to the transport, reduction, and release of radioactive
material in and from the primary containment in PWRs or the drywell in BWRs are as follows:
3.1 The radioactivity released from the fuel should be assumed to mix instantaneously and
homogeneously throughout the free air volume of the primary containment in PWRs or the
drywell in BWRs as it is released. This distribution should be adjusted if there are internal
compartments that have limited ventilation exchange. The suppression pool free air
volume may be included provided there is a mechanism to ensure mixing between the
drywell to the wetwell. The release into the containment or drywell should be assumed to
terminate at the end of the early in-vessel phase.
3.2 Reduction in airborne radioactivity in the containment by natural deposition within the
containment may be credited. Acceptable models for removal of iodine and aerosols are
1 This document describes statistical formulations with differing levels of uncertainty. The removal rate constants selected for
use in design basis calculations should be those that will maximize the dose consequences. For BWRs, the simplified model
should be used only if the release from the core is not directed through the suppression pool. Iodine removal in the suppression
pool affects the iodine species assumed by the model to be present initially.
A-2
described in Chapter 6.5.2, “Containment Spray as a Fission Product Cleanup System,” of
the Standard Review Plan (SRP), NUREG-0800 (Ref. A-1) and in NUREG/CR-6189, “A
Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments”
(Ref. A-2). The latter model is incorporated into the analysis code RADTRAD (Ref. A-3).
The prior practice of deterministically assuming that a 50% plateout of iodine is released
from the fuel is no longer acceptable to the NRC staff as it is inconsistent with the
characteristics of the revised source terms.
3.3 Reduction in airborne radioactivity in the containment by containment spray systems that
have been designed and are maintained in accordance with Chapter 6.5.2 of the SRP (Ref.
A-1) may be credited. Acceptable models for the removal of iodine and aerosols are
described in Chapter 6.5.2 of the SRP and NUREG/CR-5966, “A Simplified Model of
Aerosol Removal by Containment Sprays”1 (Ref. A-4). This simplified model is
incorporated into the analysis code RADTRAD (Refs. A-1 to A-3).
The evaluation of the containment sprays should address areas within the primary
containment that are not covered by the spray drops. The mixing rate attributed to natural
convection between sprayed and unsprayed regions of the containment building, provided
that adequate flow exists between these regions, is assumed to be two turnovers of the
unsprayed regions per hour, unless other rates are justified. The containment building
atmosphere may be considered a single, well-mixed volume if the spray covers at least 90%
of the volume and if adequate mixing of unsprayed compartments can be shown.
The SRP sets forth a maximum decontamination factor (DF) for elemental iodine based on
the maximum iodine activity in the primary containment atmosphere when the sprays
actuate, divided by the activity of iodine remaining at some time after decontamination.
The SRP also states that the particulate iodine removal rate should be reduced by a factor
of 10 when a DF of 50 is reached. The reduction in the removal rate is not required if the
removal rate is based on the calculated time-dependent airborne aerosol mass. There is no
specified maximum DF for aerosol removal by sprays. The maximum activity to be used in
determining the DF is defined as the iodine activity in the columns labeled “Total” in
Tables 1 and 2 of this guide multiplied by 0.05 for elemental iodine and by 0.95 for
particulate iodine (i.e., aerosol treated as particulate in SRP methodology).
3.4 Reduction in airborne radioactivity in the containment by in-containment recirculation filter
systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and
Generic Letter 99-02 (Refs. A-5 and A-6). The filter media loading caused by the
increased aerosol release associated with the revised source term should be addressed.
3.5 Reduction in airborne radioactivity in the containment by suppression pool scrubbing in
BWRs should generally not be credited. However, the staff may consider such reduction on
an individual case basis. The evaluation should consider the relative timing of the blowdown
and the fission product release from the fuel, the force driving the release through the pool,
A-3
and the potential for any bypass of the suppression pool (Ref. 7). Analyses should consider
iodine re-evolution if the suppression pool liquid pH is not maintained greater than 7.
3.6 Reduction in airborne radioactivity in the containment by retention in ice condensers, or other
engineering safety features not addressed above, should be evaluated on an individual case
basis. See Section 6.5.4 of the SRP (Ref. A-1).
3.7 The primary containment (i.e., drywell for Mark I and II containment designs) should be
assumed to leak at the peak pressure technical specification leak rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the technical
specification leak rate. For BWRs, leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if
supported by plant configuration and analyses, to a value not less than 50% of the technical
specification leak rate. Leakage from subatmospheric containments is assumed to terminate
when the containment is brought to and maintained at a subatmospheric condition as defined
by technical specifications.
For BWRs with Mark III containments, the leakage from the drywell into the primary
containment should be based on the steaming rate of the heated reactor core, with no credit
for core debris relocation. This leakage should be assumed during the two-hour period
between the initial blowdown and termination of the fuel radioactivity release (gap and early
in-vessel release phases). After two hours, the radioactivity is assumed to be uniformly
distributed throughout the drywell and the primary containment.
3.8 If the primary containment is routinely purged during power operations, releases via the
purge system prior to containment isolation should be analyzed and the resulting doses
summed with the postulated doses from other release paths. The purge release evaluation
should assume that 100% of the radionuclide inventory in the reactor coolant system liquid is
released to the containment at the initiation of the LOCA. This inventory should be based on
the technical specification reactor coolant system equilibrium activity. Iodine spikes need not
be considered. If the purge system is not isolated before the onset of the gap release phase,
the release fractions associated with the gap release and early in-vessel phases should be
considered as applicable.
ASSUMPTIONS ON DUAL CONTAINMENTS
4. For facilities with dual containment systems, the acceptable assumptions related to the
transport, reduction, and release of radioactive material in and from the secondary containment or
enclosure buildings are as follows.
4.1 Leakage from the primary containment should be considered to be collected, processed by
engineered safety feature (ESF) filters, if any, and released to the environment via the
secondary containment exhaust system during periods in which the secondary containment
has a negative pressure as defined in technical specifications. Credit for an elevated release
should be assumed only if the point of physical release is more than two and one-half times
the height of any adjacent structure.
A-4
4.2 Leakage from the primary containment is assumed to be released directly to the environment
as a ground-level release during any period in which the secondary containment does not
have a negative pressure as defined in technical specifications.
4.3 The effect of high wind speeds on the ability of the secondary containment to maintain a
negative pressure should be evaluated on an individual case basis. The wind speed to be
assumed is the 1-hour average value that is exceeded only 5% of the total number of hours in
the data set. Ambient temperatures used in these assessments should be the 1-hour average
value that is exceeded only 5% or 95% of the total numbers of hours in the data set,
whichever is conservative for the intended use (e.g., if high temperatures are limiting, use
those exceeded only 5%).
4.4 Credit for dilution in the secondary containment may be allowed when adequate means to
cause mixing can be demonstrated. Otherwise, the leakage from the primary containment
should be assumed to be transported directly to exhaust systems without mixing. Credit for
mixing, if found to be appropriate, should generally be limited to 50%. This evaluation
should consider the magnitude of the containment leakage in relation to contiguous building
volume or exhaust rate, the location of exhaust plenums relative to projected release
locations, the recirculation ventilation systems, and internal walls and floors that impede
stream flow between the release and the exhaust.
4.5 Primary containment leakage that bypasses the secondary containment should be evaluated at
the bypass leak rate incorporated in the technical specifications. If the bypass leakage is
through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine
and aerosols may be considered on a case-by-case basis. Similarly, deposition of aerosol
radioactivity in gas-filled lines may be considered on a case-by-case basis.
4.6 Reduction in the amount of radioactive material released from the secondary containment
because of ESF filter systems may be taken into account provided that these systems meet the
guidance of Regulatory Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).
ASSUMPTIONS ON ESF SYSTEM LEAKAGE
5. ESF systems that recirculate sump water outside of the primary containment are assumed to
leak during their intended operation. This release source includes leakage through valve packing
glands, pump shaft seals, flanged connections, and other similar components. This release source
may also include leakage through valves isolating interfacing systems (Ref. A-7). The radiological
consequences from the postulated leakage should be analyzed and combined with consequences
postulated for other fission product release paths to determine the total calculated radiological
consequences from the LOCA. The following assumptions are acceptable for evaluating the
consequences of leakage from ESF components outside the primary containment for BWRs and
PWRs.
5.1 With the exception of noble gases, all the fission products released from the fuel to the
containment (as defined in Tables 1 and 2 of this guide) should be assumed to
instantaneously and homogeneously mix in the primary containment sump water (in PWRs)
or suppression pool (in BWRs) at the time of release from the core. In lieu of this
A-5
deterministic approach, suitably conservative mechanistic models for the transport of
airborne activity in containment to the sump water may be used. Note that many of the
parameters that make spray and deposition models conservative with regard to containment
airborne leakage are nonconservative with regard to the buildup of sump activity.
5.2 The leakage should be taken as two times the sum of the simultaneous leakage from all
components in the ESF recirculation systems above which the technical specifications, or
licensee commitments to item III.D.1.1 of NUREG-0737 (Ref. A-8), would require declaring
such systems inoperable. The leakage should be assumed to start at the earliest time the
recirculation flow occurs in these systems and end at the latest time the releases from these
systems are terminated. Consideration should also be given to design leakage through valves
isolating ESF recirculation systems from tanks vented to atmosphere, e.g., emergency core
cooling system (ECCS) pump miniflow return to the refueling water storage tank.
5.3 With the exception of iodine, all radioactive materials in the recirculating liquid should be
assumed to be retained in the liquid phase.
5.4 If the temperature of the leakage exceeds 212°F, the fraction of total iodine in the liquid that
becomes airborne should be assumed equal to the fraction of the leakage that flashes to
vapor. This flash fraction, FF, should be determined using a constant enthalpy, h, process,
based on the maximum time-dependent temperature of the sump water circulating outside the
containment:
h h
h
f f
fg
= −1 2
Where: hf1 is the enthalpy of liquid at system design temperature and pressure; hf2 is the
enthalpy of liquid at saturation conditions (14.7 psia, 212ºF); and hfg is the heat of
vaporization at 212ºF.
5.5 If the temperature of the leakage is less than 212°F or the calculated flash fraction is less than
10%, the amount of iodine that becomes airborne should be assumed to be 10% of the total
iodine activity in the leaked fluid, unless a smaller amount can be justified based on the
actual sump pH history and area ventilation rates.
5.6 The radioiodine that is postulated to be available for release to the environment is assumed to
be 97% elemental and 3% organic. Reduction in release activity by dilution or holdup within
buildings, or by ESF ventilation filtration systems, may be credited where applicable. Filter
systems used in these applications should be evaluated against the guidance of Regulatory
Guide 1.52 (Ref. A-5) and Generic Letter 99-02 (Ref. A-6).
ASSUMPTIONS ON MAIN STEAM ISOLATION VALVE LEAKAGE IN BWRS
6. For BWRs, the main steam isolation valves (MSIVs) have design leakage that may result in a
radioactivity release. The radiological consequences from postulated MSIV leakage should be
analyzed and combined with consequences postulated for other fission product release paths to
A-6
determine the total calculated radiological consequences from the LOCA. The following
assumptions are acceptable for evaluating the consequences of MSIV leakage.
6.1 For the purpose of this analysis, the activity available for release via MSIV leakage
should be assumed to be that activity determined to be in the drywell for evaluating
containment leakage (see Regulatory Position 3). No credit should be assumed for
activity reduction by the steam separators or by iodine partitioning in the reactor vessel.
6.2 All the MSIVs should be assumed to leak at the maximum leak rate above which the
technical specifications would require declaring the MSIVs inoperable. The leakage
should be assumed to continue for the duration of the accident. Postulated leakage may
be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by site-specific analyses, to a value not
less than 50% of the maximum leak rate.
6.3 Reduction of the amount of released radioactivity by deposition and plateout on steam
system piping upstream of the outboard MSIVs may be credited, but the amount of
reduction in concentration allowed will be evaluated on an individual case basis.
Generally, the model should be based on the assumption of well-mixed volumes, but
other models such as slug flow may be used if justified.
6.4 In the absence of collection and treatment of releases by ESFs such as the MSIV leakage
control system, or as described in paragraph 6.5 below, the MSIV leakage should be
assumed to be released to the environment as an unprocessed, ground- level release.
Holdup and dilution in the turbine building should not be assumed.
6.5 A reduction in MSIV releases that is due to holdup and deposition in main steam piping
downstream of the MSIVs and in the main condenser, including the treatment of air
ejector effluent by offgas systems, may be credited if the components and piping systems
used in the release path are capable of performing their safety function during and
following a safe shutdown earthquake (SSE). The amount of reduction allowed will be
evaluated on an individual case basis. References A-9 and A-10 provide guidance on
acceptable models.
ASSUMPTION ON CONTAINMENT PURGING
7. The radiological consequences from post-LOCA primary containment purging as a
combustible gas or pressure control measure should be analyzed. If the installed containment
purging capabilities are maintained for purposes of severe accident management and are not
credited in any design basis analysis, radiological consequences need not be evaluated. If the
primary containment purging is required within 30 days of the LOCA, the results of this analysis
should be combined with consequences postulated for other fission product release paths to
determine the total calculated radiological consequences from the LOCA. Reduction in the
amount of radioactive material released via ESF filter systems may be taken into account
provided that these systems meet the guidance in Regulatory Guide 1.52 (Ref. A-5) and Generic
Letter 99-02 (Ref. A-6).
A-7
Appendix A REFERENCES
A-1 USNRC, “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear
Power Plants,” NUREG-0800.
A-2 D.A. Powers et al, “A Simplified Model of Aerosol Removal by Natural Processes in
Reactor Containments,” NUREG/CR-6189, USNRC, July 1996.
A-3 S.L. Humphreys et al., “RADTRAD: A Simplified Model for Radionuclide Transport and
Removal and Dose Estimation,” NUREG/CR-6604, USNRC, April 1998.
A-4 D.A. Powers and S.B. Burson, “A Simplified Model of Aerosol Removal by Containment
Sprays,” NUREG/CR-5966, USNRC, June 1993.
A-5 USNRC, “Design, Testing, and Maintenance Criteria for Postaccident EngineeredSafety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of LightWater-Cooled Nuclear Power Plants,” Regulatory Guide 1.52, Revision 2, March 1978.
A-6 USNRC, “Laboratory Testing of Nuclear Grade Activated Charcoal,” Generic Letter 99-
02, June 3, 1999.
A-7 USNRC, “Potential Radioactive Leakage to Tank Vented to Atmosphere,” Information
Notice 91-56, September 19, 1991.
A-8 USNRC, “Clarification of TMI Action Plan Requirements,” NUREG-0737, November
1980.
A-9 J.E. Cline, “MSIV Leakage Iodine Transport Analysis,” Letter Report dated March 26,
1991. (ADAMS Accession Number ML003683718)
A-10 USNRC, “Safety Evaluation of GE Topical Report, NEDC-31858P (Proprietary GE
report), Revision 2, BWROG Report for Increasing MSIV Leakage Limits and
Elimination of Leakage Control Systems, September 1993,” letter dated March 3, 1999,
ADAMS Accession Number 9903110303.
B-1
Appendix B
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES
OF A FUEL HANDLING ACCIDENT
This appendix provides assumptions acceptable to the staff for evaluating the radiological
consequences of a fuel handling accident at light water reactors. These assumptions supplement
the guidance provided in the main body of this guide.
1. SOURCE TERM
Acceptable assumptions regarding core inventory and the release of radionuclides from
the fuel are provided in Regulatory Position 3 of this guide. The following assumptions also
apply.
1.1 The number of fuel rods damaged during the accident should be based on a conservative
analysis that considers the most limiting case. This analysis should consider parameters
such as the weight of the dropped heavy load or the weight of a dropped fuel assembly
(plus any attached handling grapples), the height of the drop, and the compression,
torsion, and shear stresses on the irradiated fuel rods. Damage to adjacent fuel
assemblies, if applicable (e.g., events over the reactor vessel), should be considered.
1.2 The fission product release from the breached fuel is based on Regulatory Position 3.2 of
this guide and the estimate of the number of fuel rods breached. All the gap activity in
the damaged rods is assumed to be instantaneously released. Radionuclides that should
be considered include xenons, kryptons, halogens, cesiums, and rubidiums.
1.3 The chemical form of radioiodine released from the fuel to the spent fuel pool should be
assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent
organic iodide. The CsI released from the fuel is assumed to completely dissociate in the
pool water. Because of the low pH of the pool water, the iodine re-evolves as elemental
iodine. This is assumed to occur instantaneously. The NRC staff will consider, on a
case-by-case basis, justifiable mechanistic treatment of the iodine release from the pool.
2. WATER DEPTH
If the depth of water above the damaged fuel is 23 feet or greater, the decontamina-tion
factors for the elemental and organic species are 500 and 1, respectively, giving an overall
effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the
damaged rods is retained by the water). This difference in decontamination factors for elemental
(99.85%) and organic iodine (0.15%) species results in the iodine above the water being
composed of 57% elemental and 43% organic species. If the depth of water is not 23 feet, the
decontamination factor will have to be determined on a case-by-case method (Ref. B-1).
1 These analyses should consider the time for the radioactivity concentration to reach levels corresponding to the monitor
setpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, as
applicable.
2 Containment isolation does not imply containment integrity as defined by technical specifications for non-shutdown modes.
The term isolation is used here collectively to encompass both containment integrity and containment closure, typically in place
during shutdown periods. To be credited in the analysis, the appropriate form of isolation should be addressed in technical
specifications.
B-2
3. NOBLE GASES
The retention of noble gases in the water in the fuel pool or reactor cavity is negligible
(i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the
water in the fuel pool or reactor cavity (i.e., infinite decontamination factor).
4. FUEL HANDLING ACCIDENTS WITHIN THE FUEL BUILDING
For fuel handling accidents postulated to occur within the fuel building, the following
assumptions are acceptable to the NRC staff.
4.1 The radioactive material that escapes from the fuel pool to the fuel building is assumed to
be released to the environment over a 2-hour time period.
4.2 A reduction in the amount of radioactive material released from the fuel pool by
engineered safety feature (ESF) filter systems may be taken into account provided these
systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2,
B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of
ventilation flow to the ESF filtration system1 should be determined and accounted for in
the radioactivity release analyses.
4.3 The radioactivity release from the fuel pool should be assumed to be drawn into the ESF
filtration system without mixing or dilution in the fuel building. If mixing can be
demonstrated, credit for mixing and dilution may be considered on a case-by-case basis.
This evaluation should consider the magnitude of the building volume and exhaust rate,
the potential for bypass to the environment, the location of exhaust plenums relative to
the surface of the pool, recirculation ventilation systems, and internal walls and floors that
impede stream flow between the surface of the pool and the exhaust plenums.
5. FUEL HANDLING ACCIDENTS WITHIN CONTAINMENT
For fuel handling accidents postulated to occur within the containment, the following
assumptions are acceptable to the NRC staff.
5.1 If the containment is isolated2 during fuel handling operations, no radiological
consequences need to be analyzed.
5.2 If the containment is open during fuel handling operations, but designed to automatically
isolate in the event of a fuel handling accident, the release duration should be based on
3 The staff will generally require that technical specifications allowing such operations include administrative controls to close
the airlock, hatch, or open penetrations within 30 minutes. Such adminstrative controls will generally require that a dedicated
individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occur.
Radiological analyses should generally not credit this manual isolation.
B-3
delays in radiation detection and completion of containment isolation. If it can be shown
that containment isolation occurs before radioactivity is released to the environment,1 no
radiological consequences need to be analyzed.
5.3 If the containment is open during fuel handling operations (e.g., personnel air lock or
equipment hatch is open),3 the radioactive material that escapes from the reactor cavity
pool to the containment is released to the environment over a 2-hour time period.
5.4 A reduction in the amount of radioactive material released from the containment by ESF
filter systems may be taken into account provided that these systems meet the guidance of
Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation
detection, actuation of the ESF filtration system, or diversion of ventilation flow to the
ESF filtration system should be determined and accounted for in the radioactivity release
analyses.1
5.5 Credit for dilution or mixing of the activity released from the reactor cavity by natural or
forced convection inside the containment may be considered on a case-by-case basis.
Such credit is generally limited to 50% of the containment free volume. This evaluation
should consider the magnitude of the containment volume and exhaust rate, the potential
for bypass to the environment, the location of exhaust plenums relative to the surface of
the reactor cavity, recirculation ventilation systems, and internal walls and floors that
impede stream flow between the surface of the reactor cavity and the exhaust plenums.
B-4
Appendix B REFERENCES
B-1. G. Burley, “Evaluation of Fission Product Release and Transport,” Staff Technical Paper,
1971. (NRC Accession number 8402080322 in ADAMS or PARS)
B-2. USNRC, “Design, Testing, and Maintenance Criteria for Postaccident Engineered-SafetyFeature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-WaterCooled Nuclear Power Plants,” Regulatory Guide 1.52, Revision 2, March 1978.
B-3. USNRC, “Laboratory Testing of Nuclear Grade Activated Charcoal,” Generic Letter 99-
02, June 3, 1999.
1 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum
technical specification values, whichever maximizes the radiological consequences. In determining the dose equivalent I-131
(DE I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected
fuel damage should not be included.
2 If there are forced flow paths from the turbine or condenser, such as unisolated motor vacuum pumps or unprocessed air
ejectors, the leakage rate should be assumed to be the flow rate associated with the most limiting of these paths. Credit for
collection and processing of releases, such as by off gas or standby gas treatment, will be considered on a case-by-case basis.
C-1
Appendix C
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES
OF A BWR ROD DROP ACCIDENT
This appendix provides assumptions acceptable to the NRC staff for evaluating the
radiological consequences of a rod drop accident at BWR light-water reactors. These
assumptions supplement the guidance provided in the main body of this guide.
1. Assumptions acceptable to the NRC staff regarding core inventory are provided in
Regulatory Position 3 of this guide. For the rod drop accident, the release from the breached fuel
is based on the estimate of the number of fuel rods breached and the assumption that 10% of the
core inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel
melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for
fuel melting and on the assumption that 100% of the noble gases and 50% of the iodines
contained in that fraction are released to the reactor coolant.
2. If no or minimal1 fuel damage is postulated for the limiting event, the released activity
should be the maximum coolant activity (typically 4 µCi/gm DE I-131) allowed by the technical
specifications.
3. The assumptions acceptable to the NRC staff that are related to the transport, reduction,
and release of radioactive material from the fuel and the reactor coolant are as follows.
3.1 The activity released from the fuel from either the gap or from fuel pellets is assumed to
be instantaneously mixed in the reactor coolant within the pressure vessel.
3.2 Credit should not be assumed for partitioning in the pressure vessel or for removal by the
steam separators.
3.3 Of the activity released from the reactor coolant within the pressure vessel, 100% of the
noble gases, 10% of the iodine, and 1% of the remaining radionuclides are assumed to
reach the turbine and condensers.
3.4 Of the activity that reaches the turbine and condenser, 100% of the noble gases, 10% of
the iodine, and 1% of the particulate radionuclides are available for release to the
environment. The turbine and condensers leak to the atmosphere as a ground- level
release at a rate of 1% per day2 for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leakage is
assumed to terminate. No credit should be assumed for dilution or holdup within the
C-2
turbine building. Radioactive decay during holdup in the turbine and condenser may be
assumed.
3.5 In lieu of the transport assumptions provided in paragraphs 3.2 through 3.4 above, a more
mechanistic analysis may be used on a case-by-case basis. Such analyses account for the
quantity of contaminated steam carried from the pressure vessel to the turbine and
condensers based on a review of the minimum transport time from the pressure vessel to
the first main steam isolation (MSIV) and considers MSIV closure time.
3.6 The iodine species released from the reactor coolant within the pressure vessel should be
assumed to be 95% CsI as an aerosol, 4.85% elemental, and 0.15% organic. The release
from the turbine and condenser should be assumed to be 97% elemental and 3% organic.
1 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum
technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent I-131 (DE I131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel
damage should not be included.
D-1
Appendix D
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A
BWR MAIN STEAM LINE BREAK ACCIDENT
This appendix provides assumptions acceptable to the NRC staff for evaluating the
radiological consequences of a main steam line accident at BWR light water reactors. These
assumptions supplement the guidance provided in the main body of this guide.
SOURCE TERM
1. Assumptions acceptable to the NRC staff regarding core inventory and the release of
radionuclides from the fuel are provided in Regulatory Position 3 of this guide. The release from
the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the
number of fuel rods breached.
2. If no or minimal1 fuel damage is postulated for the limiting event, the released activity
should be the maximum coolant activity allowed by technical specification. The iodine
concentration in the primary coolant is assumed to correspond to the following two cases in the
nuclear steam supply system vendor’s standard technical specifications.
2.1 The concentration that is the maximum value (typically 4.0 µCi/gm DE I-131) permitted
and corresponds to the conditions of an assumed pre-accident spike, and
2.1 The concentration that is the maximum equilibrium value (typically 0.2 µCi/gm DE
I-131) permitted for continued full power operation.
3. The activity released from the fuel should be assumed to mix instantaneously and
homogeneously in the reactor coolant. Noble gases should be assumed to enter the steam phase
instantaneously.
TRANSPORT
4. Assumptions acceptable to the NRC staff related to the transport, reduction, and release
of radioactive material to the environment are as follows.
4.1 The main steam line isolation valves (MSIV) should be assumed to close in the maximum
time allowed by technical specifications.
4.2 The total mass of coolant released should be assumed to be that amount in the steam line
and connecting lines at the time of the break plus the amount that passes through the
valves prior to closure.
D-2
4.3 All the radioactivity in the released coolant should be assumed to be released to the
atmosphere instantaneously as a ground-level release. No credit should be assumed for
plateout, holdup, or dilution within facility buildings.
4.4 The iodine species released from the main steam line should be assumed to be 95% CsI as
an aerosol, 4.85% elemental, and 0.15% organic.
1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the
guidance that is being developed in Draft Regulatory Guide DG-1074, “Steam Generator Tube Integrity,” for acceptable
assumptions and methodologies for performing radiological analyses.
2 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum
technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent I-131 (DE I131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel
damage should not be included.
E-1
Appendix E
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A
PWR MAIN STEAM LINE BREAK ACCIDENT
This appendix provides assumptions acceptable to the NRC staff for evaluating the
radiological consequences of a main steam line break accident at PWR light water reactors.
These assumptions supplement the guidance provided in the main body of this guide.1
SOURCE TERMS
1. Assumptions acceptable to the NRC staff regarding core inventory and the release of
radionuclides from the fuel are provided in Regulatory Position 3 of this regulatory guide. The
release from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate
of the number of fuel rods breached. The fuel damage estimate should assume that the highest
worth control rod is stuck at its fully withdrawn position.
2. If no or minimal2 fuel damage is postulated for the limiting event, the activity released
should be the maximum coolant activity allowed by the technical specifications. Two cases of
iodine spiking should be assumed.
2.1 A reactor transient has occurred prior to the postulated main steam line break (MSLB)
and has raised the primary coolant iodine concentration to the maximum value (typically
60 µCi/gm DE I-131) permitted by the technical specifications (i.e., a preaccident iodine
spike case).
2.2 The primary system transient associated with the MSLB causes an iodine spike in the
primary system. The increase in primary coolant iodine concentration is estimated using a
spiking model that assumes that the iodine release rate from the fuel rods to the primary
coolant (expressed in curies per unit time) increases to a value 500 times greater than the
release rate corresponding to the iodine concentration at the equilibrium value (typically
1.0 µCi/gm DE I-131) specified in technical specifications (i.e., concurrent iodine spike
case). A concurrent iodine spike need not be considered if fuel damage is postulated.
The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be
considered on a case-by-case basis if it can be shown that the activity released by the 8-
hour spike exceeds that available for release from the fuel gap of all fuel pins.
3. The activity released from the fuel should be assumed to be released instantaneously and
homogeneously through the primary coolant.
3 In this appendix, ruptured refers to the state of the steam generator in which primary-to-secondary leakage rate has increased to
a value greater than technical specifications. Faulted refers to the state of the steam generator in which the secondary side has
been depressurized by a MSLB such that protective system response (main steam line isolation, reactor trip, safety injection, etc.)
has occurred. Partitioning Coefficient is defined as:
PC mass of I per unit mass of liquid
mass of I per unit mass of gas = 2
2
E-2
4. The chemical form of radioiodine released from the fuel should be assumed to be 95%
cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine
releases from the steam generators to the environment should be assumed to be 97% elemental
and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine
released during normal operations, including iodine spiking.
TRANSPORT 3
5. Assumptions acceptable to the NRC staff related to the transport, reduction, and release
of radioactive material to the environment are as follows.
5.1 For facilities that have not implemented alternative repair criteria (see Ref. E-1, DG1074), the primary-to-secondary leak rate in the steam generators should be assumed to
be the leak rate limiting condition for operation specified in the technical specifications.
For facilities with traditional generator specifications (both per generator and total of all
generators), the leakage should be apportioned between affected and unaffected steam
generators in such a manner that the calculated dose is maximized.
5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
lbm/hr) should be consistent with the basis of the parameter being converted. The ARC
leak rate correlations are generally based on the collection of cooled liquid. Surveillance
tests and facility instrumentation used to show compliance with leak rate technical
specifications are typically based on cooled liquid. In most cases, the density should be
assumed to be 1.0 gm/cc (62.4 lbm/ft3
).
5.3 The primary-to-secondary leakage should be assumed to continue until the primary
system pressure is less than the secondary system pressure, or until the temperature of the
leakage is less than 100°C (212°F). The release of radioactivity from unaffected steam
generators should be assumed to continue until shutdown cooling is in operation and
releases from the steam generators have been terminated.
5.4 All noble gas radionuclides released from the primary system are assumed to be released
to the environment without reduction or mitigation.
5.5 The transport model described in this section should be utilized for iodine and particulate
releases from the steam generators. This model is shown in Figure E-1 and summarized
below:
E-3
Steam Space
Bulk Water
Primary
Leakage
Scrubbing
Partitioning
Release
Figure E-1
Transport Model
5.5.1 A portion of the primary-to-secondary leakage will flash to vapor, based on the
thermodynamic conditions in the reactor and secondary coolant.
• During periods of steam generator dryout, all of the primary-to-secondary
leakage is assumed to flash to vapor and be released to the environment
with no mitigation.
• With regard to the unaffected steam generators used for plant cooldown,
the primary-to-secondary leakage can be assumed to mix with the
secondary water without flashing during periods of total tube
submergence.
5.5.2 The leakage that immediately flashes to vapor will rise through the bulk water of
the steam generator and enter the steam space. Credit may be taken for scrubbing
in the generator, using the models in NUREG-0409, “Iodine Behavior in a PWR
Cooling System Following a Postulated Steam Generator Tube Rupture Accident”
(Ref. E-2), during periods of total submergence of the tubes.
5.5.3 The leakage that does not immediately flash is assumed to mix with the bulk
water.
5.5.4 The radioactivity in the bulk water is assumed to become vapor at a rate that is the
function of the steaming rate and the partition coefficient. A partition coefficient
for iodine of 100 may be assumed. The retention of particulate radionuclides in
the steam generators is limited by the moisture carryover from the steam
generators.
5.6 Operating experience and analyses have shown that for some steam generator designs,
tube uncovery may occur for a short period following any reactor trip (Ref. E-3). The
potential impact of tube uncovery on the transport model parameters (e.g., flash fraction,
scrubbing credit) needs to be considered. The impact of emergency operating procedure
restoration strategies on steam generator water levels should be evaluated.
E-4
Appendix E REFERENCES
E-1 USNRC, “Steam Generator Tube Integrity,” Draft Regulatory Guide DG-1074, December
1998.
E-2. USNRC, “Iodine Behavior in a PWR Cooling System Following a Postulated Steam
Generator Tube Rupture Accident,” NUREG-0409, May 1985.
E-3 USNRC, “Steam Generator Tube Rupture Analysis Deficiency,” Information Notice 88-
31, May 25, 1988.
1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the
guidance that is being developed in Draft Regulatory Guide DG-1074, “Steam Generator Tube Integrity” (USNRC, December
1998), for acceptable assumptions and methodologies for performing radiological analyses.
2 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum
technical specification values, whichever maximizes the radiological consequences. In determining dose equivalent I-131 (DE I131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel
damage should not be included.
F-1
Appendix F
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A
PWR STEAM GENERATOR TUBE RUPTURE ACCIDENT
This appendix provides assumptions acceptable to the NRC staff for evaluating the
radiological consequences of a steam generator tube rupture accident at PWR light-water
reactors. These assumptions supplement the guidance provided in the main body of this guide.1
SOURCE TERM
1. Assumptions acceptable to the NRC staff regarding core inventory and the release of
radionuclides from the fuel are in Regulatory Position 3 of this guide. The release from the
breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of
fuel rods breached.
2. If no or minimal2 fuel damage is postulated for the limiting event, the activity released
should be the maximum coolant activity allowed by technical specification. Two cases of iodine
spiking should be assumed.
2.1 A reactor transient has occurred prior to the postulated steam generator tube rupture
(SGTR) and has raised the primary coolant iodine concentration to the maximum value
(typically 60 µCi/gm DE I-131) permitted by the technical specifications (i.e., a
preaccident iodine spike case).
2.2 The primary system transient associated with the SGTR causes an iodine spike in the
primary system. The increase in primary coolant iodine concentration is estimated using
a spiking model that assumes that the iodine release rate from the fuel rods to the primary
coolant (expressed in curies per unit time) increases to a value 335 times greater than the
release rate corresponding to the iodine concentration at the equilibrium value (typically
1.0 µCi/gm DE I-131) specified in technical specifications (i.e., concurrent iodine spike
case). A concurrent iodine spike need not be considered if fuel damage is postulated.
The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be
considered on a case-by-case basis if it can be shown that the activity released by the 8-
hour spike exceeds that available for release from the fuel gap of all fuel pins.
3. The activity released from the fuel, if any, should be assumed to be released
instantaneously and homogeneously through the primary coolant.
3 In this appendix, ruptured refers to the state of the steam generator in which primary-to-secondary leakage rate has increased to
a value greater than technical specifications.
F-2
4. Iodine releases from the steam generators to the environment should be assumed to be
97% elemental and 3% organic.
TRANSPORT 3
5. Assumptions acceptable to the NRC staff related to the transport, reduction, and release
of radioactive material to the environment are as follows:
5.1 The primary-to-secondary leak rate in the steam generators should be assumed to be the
leak rate limiting condition for operation specified in the technical specifications. The
leakage should be apportioned between affected and unaffected steam generators in such
a manner that the calculated dose is maximized.
5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
lbm/hr) should be consistent with the basis of surveillance tests used to show compliance
with leak rate technical specifications. These tests are typically based on cool liquid.
Facility instrumentation used to determine leakage is typically located on lines containing
cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3
).
5.3 The primary-to-secondary leakage should be assumed to continue until the primary
system pressure is less than the secondary system pressure, or until the temperature of the
leakage is less than 100°C (212° F). The release of radioactivity from the unaffected
steam generators should be assumed to continue until shutdown cooling is in operation
and releases from the steam generators have been terminated.
5.4 The release of fission products from the secondary system should be evaluated with the
assumption of a coincident loss of offsite power.
5.5 All noble gas radionuclides released from the primary system are assumed to be released
to the environment without reduction or mitigation.
5.6 The transport model described in Regulatory Positions 5.5 and 5.6 of Appendix E should
be utilized for iodine and particulates.
1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the
guidance that is being developed in Draft Regulatory Guide DG-1074, “Steam Generator Tube Integrity” (USNRC, December
1998), for acceptable assumptions and methodologies for performing radiological analyses.
G-1
Appendix G
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A
PWR LOCKED ROTOR ACCIDENT
This appendix provides assumptions acceptable to the NRC staff for evaluating the
radiological consequences of a locked rotor accident at PWR light water reactors.1 These
assumptions supplement the guidance provided in the main body of this guide.
SOURCE TERM
1. Assumptions acceptable to the NRC staff regarding core inventory and the release of
radionuclides from the fuel are in Regulatory Position 3 of this regulatory guide. The release
from the breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the
number of fuel rods breached.
2. If no fuel damage is postulated for the limiting event, a radiological analysis is not
required as the consequences of this event are bounded by the consequences projected for the
main steam line break outside containment.
3. The activity released from the fuel should be assumed to be released instantaneously and
homogeneously through the primary coolant.
4. The chemical form of radioiodine released from the fuel should be assumed to be 95%
cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. Iodine
releases from the steam generators to the environment should be assumed to be 97% elemental
and 3% organic. These fractions apply to iodine released as a result of fuel damage and to iodine
released during normal operations, including iodine spiking.
RELEASE TRANSPORT
5. Assumptions acceptable to the NRC staff related to the transport, reduction, and release
of radioactive material to the environment are as follows.
5.1 The primary-to-secondary leak rate in the steam generators should be assumed to be the
leak-rate-limiting condition for operation specified in the technical specifications. The
leakage should be apportioned between the steam generators in such a manner that the
calculated dose is maximized.
5.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
lbm/hr) should be consistent with the basis of surveillance tests used to show compliance
with leak rate technical specifications. These tests are typically based on cool liquid.
G-2
Facility instrumentation used to determine leakage is typically located on lines containing
cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3
).
5.3 The primary-to-secondary leakage should be assumed to continue until the primary
system pressure is less than the secondary system pressure, or until the temperature of the
leakage is less than 100°C (212° F). The release of radioactivity should be assumed to
continue until shutdown cooling is in operation and releases from the steam generators
have been terminated.
5.4 The release of fission products from the secondary system should be evaluated with the
assumption of a coincident loss of offsite power.
5.5 All noble gas radionuclides released from the primary system are assumed to be released
to the environment without reduction or mitigation.
5.6 The transport model described in assumptions 5.5 and 5.6 of Appendix E should be
utilized for iodine and particulates.
1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with the
guidance that is being developed in Draft Regulatory Guide DG-1074, “Steam Generator Tube Integrity” (USNRC, December
1998), for acceptable assumptions and methodologies for performing radiological analyses.
H-1
Appendix H
ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A
PWR ROD EJECTION ACCIDENT
This appendix provides assumptions acceptable to the NRC staff for evaluating the
radiological consequences of a rod ejection accident at PWR light water reactors.1 These
assumptions supplement the guidance provided in the main body of this guide.
SOURCE TERM
1. Assumptions acceptable to the NRC staff regarding core inventory are in Regulatory
Position 3 of this guide. For the rod ejection accident, the release from the breached fuel is based
on the estimate of the number of fuel rods breached and the assumption that 10% of the core
inventory of the noble gases and iodines is in the fuel gap. The release attributed to fuel melting
is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel
melting and the assumption that 100% of the noble gases and 25% of the iodines contained in
that fraction are available for release from containment. For the secondary system release
pathway, 100% of the noble gases and 50% of the iodines in that fraction are released to the
2. If no fuel damage is postulated for the limiting event, a radiological analysis is not
required as the consequences of this event are bounded by the consequences projected for the
loss-of-coolant accident (LOCA), main steam line break, and steam generator tube rupture.
3. Two release cases are to be considered. In the first, 100% of the activity released from
the fuel should be assumed to be released instantaneously and homogeneously through the
containment atmosphere. In the second, 100% of the activity released from the fuel should be
assumed to be completely dissolved in the primary coolant and available for release to the
secondary system.
4. The chemical form of radioiodine released to the containment atmosphere should be
assumed to be 95% cesium iodide (CsI), 4.85% elemental iodine, and 0.15% organic iodide. If
containment sprays do not actuate or are terminated prior to accumulating sump water, or if the
containment sump pH is not controlled at values of 7 or greater, the iodine species should be
evaluated on an individual case basis. Evaluations of pH should consider the effect of acids
created during the rod ejection accident event, e.g., pyrolysis and radiolysis products. With the
exception of elemental and organic iodine and noble gases, fission products should be assumed
to be in particulate form.
5. Iodine releases from the steam generators to the environment should be assumed to be
97% elemental and 3% organic.
H-2
TRANSPORT FROM CONTAINMENT
6. Assumptions acceptable to the NRC staff related to the transport, reduction, and release
of radioactive material in and from the containment are as follows.
6.1 A reduction in the amount of radioactive material available for leakage from the
containment that is due to natural deposition, containment sprays, recirculating filter
systems, dual containments, or other engineered safety features may be taken into
account. Refer to Appendix A to this guide for guidance on acceptable methods and
assumptions for evaluating these mechanisms.
6.2 The containment should be assumed to leak at the leak rate incorporated in the technical
specifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate
for the remaining duration of the accident. Peak accident pressure is the maximum
pressure defined in the technical specifications for containment leak testing. Leakage
from subatmospheric containments is assumed to be terminated when the containment is
brought to a subatmospheric condition as defined in technical specifications.
TRANSPORT FROM SECONDARY SYSTEM
7. Assumptions acceptable to the NRC staff related to the transport, reduction, and release
of radioactive material in and from the secondary system are as follows.
7.1 A leak rate equivalent to the primary-to-secondary leak rate limiting condition for
operation specified in the technical specifications should be assumed to exist until
shutdown cooling is in operation and releases from the steam generators have been
terminated.
7.2 The density used in converting volumetric leak rates (e.g., gpm) to mass leak rates (e.g.,
lbm/hr) should be consistent with the basis of surveillance tests used to show compliance
with leak rate technical specifications. These tests typically are based on cooled liquid.
The facility’s instrumentation used to determine leakage typically is located on lines
containing cool liquids. In most cases, the density should be assumed to be 1.0 gm/cc
(62.4 lbm/ft3
).
7.3 All noble gas radionuclides released to the secondary system are assumed to be released
to the environment without reduction or mitigation.
7.4 The transport model described in assumptions 5.5 and 5.6 of Appendix E should be
utilized for iodine and particulates.
I-1
Appendix I
ASSUMPTIONS FOR EVALUATING RADIATION DOSES FOR
EQUIPMENT QUALIFICATION
This appendix addresses assumptions associated with equipment qualification that are
acceptable to the NRC staff for performing radiological assessments. As stated in Regulatory
Position 6 of this guide, this appendix supersedes Regulatory Positions 2.c.(1) and 2.c.(2) and
Appendix D of Revision 1 of Regulatory Guide 1.89, “Environmental Qualification of Certain
Electric Equipment Important to Safety for Nuclear Power Plants” (USNRC, June 1984), for
operating reactors that have amended their licensing basis to use an alternative source term.
Except as stated in this appendix, other assumptions, methods, and provisions of Revision 1 of
Regulatory Guide 1.89 remain effective.
BASIC ASSUMPTIONS
1. Gamma and beta doses and dose rates should be determined for three types of radioactive
source distributions: (1) activity suspended in the containment atmosphere, (2) activity plated out
on containment surfaces, and (3) activity mixed in the containment sump water. A given piece of
equipment may receive a dose contribution from any or all of these sources. The amount of dose
contributed by each of these sources is determined by the location of the equipment, the timedependent and location-dependent distribution of the source, and the effects of shielding. For EQ
components located outside of the containment, additional radiation sources may include piping
and components in systems that circulate containment sump water outside of containment.
Activity deposited in ventilation and process filter media may be a source of post-accident dose.
2. The integrated dose should be determined from estimated dose rates using appropriate
integration factors determined for each of the major source terms (e.g., containment sump,
containment atmosphere, ECCS, normal operation). The period of exposure should be consistent
with the survivability period for the EQ equipment being evaluated. The survivability period is
the maximum duration, post-accident, that the particular EQ component is expected to operate
and perform its intended safety function. The period of exposure for normal operation dose is
generally the duration of the plant license, i.e., 40 years.
FISSION PRODUCT CONCENTRATIONS
3. The radiation environment resulting from normal operations should be based on the
conservative source term estimates reported in the facility's Safety Analysis Report or should be
consistent with the primary coolant specific activity limits contained in the facility's technical
specifications. The use of equilibrium primary coolant concentrations based on 1% fuel cladding
failures would be one acceptable method. In estimating the integrated dose from prior normal
operations, appropriate historical dose rate data may be used where available.
4. The radioactivity released from the core during a design basis loss-of-coolant accident
(LOCA) should be based on the assumptions provided in Regulatory Position 3 and Appendix A
of this regulatory guide. Although the design basis LOCA is generally limiting for radiological
I-2
environmental qualification (EQ) purposes, there may be components for which another design
basis accident may be limiting. In these cases, the assumptions provided in Appendices B
through H of this regulatory guide, as applicable, should be used. Applicable features and
mechanisms may be assumed in EQ calculations provided that any prerequisites and limitations
identified regarding their use are met. There are additional considerations:
• For PWR ice condenser containments, the source should be assumed to be initially
released to the lower containment compartment. The distribution of the activity should
be based on the forced recirculation fan flow rates and the transfer rates through the ice
beds as functions of time.
• For BWR Mark III designs, all the activity should be assumed initially released to the
drywell area and the transfer of activity from these regions via containment leakage to the
surrounding reactor building volume should be used to predict the qualification levels
within the reactor building (secondary containment).
DOSE MODEL FOR CONTAINMENT ATMOSPHERE
5. The beta and gamma dose rates and integrated doses from the airborne activity within the
containment atmosphere and from the plateout of aerosols on containment surfaces generally
should be calculated for the midpoint in the containment, and this dose rate should be used for all
exposed components. Radiation shielding afforded by internal structures may be neglected for
modeling simplicity. It is expected that the shielding afforded by these structures would reduce
the dose rates by factors of two or more depending on the specific location and geometry. More
detailed calculations may be warranted for selected components if acceptable dose rates cannot
be achieved using the simpler modeling assumptions.
6. Because of the short range of the betas in air, the airborne beta dose rates should be
calculated using an infinite medium model. Other models, such as finite cloud and semi-infinite
cloud, may be applicable to selected components with sufficient justification. The applicability
of the semi-infinite model would depend on the location of the component, available shielding,
and receptor geometry. For example, beta dose rates for equipment located on the containment
walls or on large internal structures might be adequately assessed using the semi-infinite model.
Use of a finite cloud model will be considered on a case-by-case method.
7. All gamma dose rates should be multiplied by a correction factor of 1.3 to account for the
omission of the contribution from the decay chains of the radionuclides. This correction is
particularly important for non-gamma-emitting radionuclides having gamma emitting progeny,
for example, Cs-137 decay to Ba-137m. This correction may be omitted if the calculational
method explicitly accounts for the emissions from buildup and decay of the radioactive progeny.
DOSE MODEL FOR CONTAINMENT SUMP WATER SOURCES
8. With the exception of noble gases, all the activity released from the fuel should be
assumed to be transported to the containment sump as it is released. This activity should be
assumed to mix instantaneously and uniformly with other liquids that drain to the sump. This
I-3
transport can also be modeled mechanistically as the time-dependent washout of airborne
aerosols by the action of containment sprays. Radionuclides that do not become airborne on
release from the reactor coolant system, e.g., they are entrained in non-flashed reactor coolant,
should be assumed to be instantaneously transported to the sump and be uniformly distributed in
the sump water.
9. The gamma and beta dose rates and the integrated doses should be calculated for a point
located on the surface of the water at the centerline of the large pool of sump water. The effects
of buildup should be considered. More detailed modeling with shielding analysis codes may be
performed.
DOSE MODEL FOR EQUIPMENT LOCATED OUTSIDE CONTAINMENT
10. EQ equipment located outside of containment may be exposed to (1) radiation from
sources within the containment building, (2) radiation from activity contained in piping and
components in systems that re-circulate containment sump water outside of containment (e.g.,
ECCS, RHR, sampling systems), (3) radiation from activity contained in piping and components
in systems that process containment atmosphere (e.g., hydrogen recombiners, purge systems), (4)
radiation from activity deposited in ventilation and process filter media, and (5) radiation from
airborne activity in plant areas outside of the containment (i.e., leakage from recirculation
systems). The amount of dose contributed by each of these sources is determined by the location
of the equipment, the time-dependent and location-dependent distribution of the source, and the
effects of shielding.
11. Because of the large amount of EQ equipment and the complexity of system and
component layout in plant buildings, it is generally not reasonable to model each EQ component.
A reasonable approach is to determine the limiting dose rate from all sources in a particular plant
area (e.g., cubicle, floor, building) to a real or hypothetical receptor and to base the integrated
doses for all components in that area on this postulated dose rate. Individual detailed modeling
of selected equipment may be performed.
12. The integrated doses from components and piping in systems recirculating sump water
should assume a source term based on the time-dependent containment sump source term
described above. Similarly, the doses from components that contain air from the containment
atmosphere should assume a source term based on the time-dependent containment atmosphere
source term described above.
13. Analyses of integrated doses caused by radiation from the buildup of activity on
ventilation and process filter media in systems containing containment sump water or atmosphere
or both should assume that the ventilation or process flow is at its nominal design value and that
the filter media is 100% efficient for iodine and particulates. The duration of flow through the
filter media should be consistent with the plant design and operating procedures. Radioactive
decay in the filter media should be considered. Shielding by structures and components between
the filter and the EQ equipment may be considered.
K-1
Appendix K
BWR Boiling water reactor
CDF Core damage frequency
CEDE Committed effective dose equivalent
COLR Core operating limits report
DBA Design basis accident
DNBR Departure from nucleate boiling ratio
EAB Exclusion area boundary
EDE Effective dose equivalent
EPA Environmental Protection Agency
EQ Environmental qualification
ESF Engineered safety feature
FHA Fuel handling accident
FSAR Final safety analysis report
IPF Iodine protection factor
LERF Large early release fraction
LOCA Loss-of-coolant accident
LPZ Low population zone
MOX Mixed oxide
MSLB Main steam line break
NDT Non-destructive testing
NSSS Nuclear supply system supplier
PRA Probabilistic risk assessment
PWR Pressurized water reactor
RMS Radiation monitoring system
SGTR Steam generator tube rupture
TEDE Total effective dose equivalent
TID Technical information document
TMI Three Mile Island
VALUE / IMPACT STATEMENT
A separate value/impact analysis has not been prepared for this Regulatory Guide 1.183.
A value/impact analysis was included in the regulatory analysis for the proposed amendments to
10 CFR Parts 21, 50, and 54 published on March 11, 1999 (64 FR 12117). This regulatory
analysis was updated as part of the final amendments to 10 CFR Parts 21, 50, and 54, published
in December 1999 (64 FR 71998). Copies of both regulatory analyses are available for
inspection or copying for a fee in the Commission’s Public Document Room at 2120 L Street
NW, Washington, DC, under RGIN AG12.
ADAMS Accession
Number ML003716792