Regulatory guides are issued to describe and make available to the public such information as methods acceptable to the NRC staff for implementing specific

parts of the NRC’s regulations, techniques used by the staff in evaluating specific problems or postulated accidents, and data needed by the NRC staff in its

review of applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and

solutions different from those set out in the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permit

or license by the Commission.

This guide was issued after consideration of comments received from the public. Comments and suggestions for improvements in these guides are encouraged

at all times, and guides will be revised, as appropriate, to accommodate comments and to reflect new information or experience. Written comments may be

submitted to the Rules and Directives Branch, ADM, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

Regulatory guides are issued in ten broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, Environmental

and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review; and 10, General.

Single copies of regulatory guides (which may be reproduced) may be obtained free of charge by writing the Distribution Services Section, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, or by fax to (301)415-2289, or by email to DISTRIBUTION@NRC.GOV. Electronic copies of this guide

are available on the internet at NRC’s home page at <WWW.NRC.GOV> in the Reference Library under Regulatory Guides and through the Electronic Reading

Room, as Accession Number ML003716792, along with other recently issued guides, at the same web site.

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

Single copies of regulatory guides, both active and draft, and draft NUREG documents may

be obtained free of charge by writing the Reproduction and Distribution Services Section, OCIO,

USNRC, Washington, DC 20555-0001, or by email to <DISTRIBUTION@NRC.GOV>, or by fax

to (301)415-2289. Active guides may also be purchased from the National Technical Information

Service on a standing order basis. Details on this service may be obtained by writing NTIS, 5285

Port Royal Road, Springfield, VA 22161.

Many NRC documents are available electronically in our Reference Library on our web

site, <WWW.NRC.GOV>, and through our Electronic Reading Room (ADAMS, or PARS,

document system) at the same site. Copies of active and draft guides and many other NRC

documents are available for inspection or copying for a fee from the NRC Public Document Room

at 2120 L Street NW., Washington, DC; the PDR’s mailing address is Mail Stop LL-6,

Washington, DC 20555; telephone (202)634-3273 or (800)397-4209; fax (202)634-3343; email is

<PDR@NRC.GOV>.

Copies of NUREG-series reports are available at current rates from the U.S. Government

Printing Office, P.O. Box 37082, Washington, DC 20402-9328 (telephone (202)512-1800); or

from the National Technical Information Service by writing NTIS at 5285 Port Royal Road,

Springfield, VA 22161; telephone (703)487-4650; or on the internet at

<http://www.ntis.gov/ordernow>. Copies are available for inspection or copying for a fee from the

NRC Public Document Room at 2120 L Street NW., Washington, DC; the PDR’s mailing address

is Mail Stop LL-6, Washington, DC 20555; telephone (202)634-3273 or (800)397-4209; fax

(202)634-3343; email is <PDR@NRC.GOV>.
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

50.34.

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
1.183-2

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
1.183-4

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

CFR Part 50)

 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.

1.183-8

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.

1.183-9

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
1.183-10

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.
1.183-15

Table 4

LOCA Release Phases

PWRs BWRs

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 hours, the breathing rate of persons offsite should be assumed to be

3.5 x 10-4 cubic meters per second. From 8 to 24 hours 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 hours 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).

DDE Equation 1 DDE V

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

CFR 50.67(b)(2)(iii).
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

MSIV (BWR) leakage

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 hours

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 hours

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 hours.

For PWRs, the leak rate may be reduced after the first 24 hours to 50% of the technical

specification leak rate. For BWRs, leakage may be reduced after the first 24 hours, 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:

FF

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 hours, 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 hours, 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 hours. 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 hours. 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

reactor coolant.

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 hours, 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

Acronyms

AST Alternative source term

BWR Boiling water reactor

CDF Core damage frequency

CEDE Committed effective dose equivalent

COLR Core operating limits report

DBA Design basis accident

DDE Deep dose equivalent

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

SG Steam generator

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