ML020160023

Draft Regulatory Guide DG-1113, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors

ML020160023
Person / Time
Issue date:	01/31/2002
From:	Office of Nuclear Regulatory Research
To:
Blumberg W M (301)415-1083
References
-nr, DG-1113
Download: ML020160023 (57)

Text

This regulatory guide is being issued in draft form to involve the public in the early stages of the development of a regulatory position in thisarea. It has not received complete staff review or approval and does not represent an official NRC staff position.Public comments are being solicited on this draft guide (including any implementation schedule) and its associated regulatory analysis orvalue/impact statement. Comments should be accompanied by appropriate supporting data. Written comments may be submitted to the Rulesand Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. Comments may besubmitted electronically or downloaded through the NRC's interactive web site at <WWW.NRC.GOV> through Rulemaking. Copies ofcomments received may be examined at the NRC Public Document Room, 11555 Rockville Pike, Rockville, MD. Comments will be most helpfulif received by April 30, 2002.Requests for single copies of draft or active regulatory guides (which may be reproduced) or for placement on an automatic distribution list forsingle copies of future draft guides in specific divisions should be made to the U.S. Nuclear Regulatory Commission, Washington, DC 20555,Attention: Reproduction and Distribution Services Section, or by fax to (301)415-2289; or by email to DISTRIBUTION@NRC.GOV. Electroniccopies of this draft guide are available through NRC's interactive web site (see above) and in NRC's Electronic Reading Room at the same website, under Accession Number ML020160023.

U.S. NUCLEAR REGULATORY COMMISSIONJanuary 2002 OFFICE OF NUCLEAR REGULATORY RESEARCHDivision 1Draft DG-1113 DRAFT REGULATORY GUIDEContact: W. M. Blumberg (301)415-1083DRAFT REGULATORY GUIDE DG-1113METHODS AND ASSUMPTIONS FOR EVALUATING RADIOLOGICAL CONSEQUENCES OF DESIGN BASIS ACCIDENTS AT LIGHT-WATER NUCLEAR POWER REACTORS GENERAL INFORMATIONWritten comments on this draft regulatory guide may be submitted to the Rules andDirectives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Comments may be accompanied by relevant information or supporting data. Copies of comments received may be examined at the NRC Public Document Room, 11555 Rockville Pike, Rockville, Maryland. You may also provide comments electronically to the Rules and Directives Branch at <http://ruleforum.llnl.gov/cgi-bin/library?source=*&library=rg_lib&file=*>. This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Ms. Carol Gallagher, (301) 415-5905; e-mail CAG@nrc.gov. For information about this draft guide andthe related documents, contact Mr. M. Blumberg, (301)415-1083; e-mail WMB1@nrc.gov.Regulatory guides are issued to describe to the public methods acceptable to the NRCstaff for implementing specific parts of the NRC's regulations, to explain techniques used by the staff in evaluating specific problems or postulated accidents, and to provide guidance to applicants. Regulatory guides are not substitutes for regulations, and compliance with regulatory guides is not required. Regulatory guides are issued first in draft form for public comment to involve the public in developing the regulatory positions. Draft regulatory guides have not received complete staff approval; they therefore do not represent official NRC staff positions. AVAILABILITIESSingle copies of regulatory guides, both active and draft, and draft NUREGdocuments may be obtained free of charge by writing the Reproduction and Distribution Services Section, OCIO, USNRC, Washington, DC 20555-0001, or by fax to (301)415-2289, or by email to <DISTRIBUTION@NRC.GOV>. 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 through our Electronic ReadingRoom at <WWW.NRC.GOV>. Copies of active and draft guides and many other NRCdocuments are available for inspection or copying for a fee from the NRC Public Document Room at 11555 Rockville Pike (first floor), Rockville, MD; the PDR's mailing address is USNRC PDR, Washington, DC 20555; telephone (301)415-4737 or 1-(800)397-4209; fax (301)415-3548; e-mail <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 11555 Rockville Pike (first floor), Rockville, MD; the PDR's mailing address is USNRC PDR, Washington, DC 20555; telephone (301)415-4737 or 1-(800)397-4209; fax (301)415-3548; e-mail <PDR@NRC.GOV>.

iiiTABLE OF CONTENTSA. INTRODUCTION..........................................................1B. DISCUSSION.............................................................2 C. REGULATORY POSITION...................................................4 1.GENERIC CONSIDERATIONS ............................................41.1Safety Margins....................................................4 1.2Defense in Depth..................................................4 1.3Scope of Required Analyses.........................................41.4Risk Implications..................................................6 1.5Submittal Requirements.............................................6 1.6Final Safety Analysis Report Requirements..............................72.DOSE ANALYSIS MODELS...............................................72.1 Radiological Consequences..........................................72.2 Activities in Compartments Without Inflow from the Environment.............82.3 Activities in the Environment.........................................8 2.4 Activities in Compartments That Intake Only Outside Contaminated Air........9 2.5 Integrated Activity Released into the Environment ........................9 2.6 Integrated Activity in a Compartment..................................10 2.7 Offsite Doses....................................................10 2.8 Compartment Doses..............................................103.ACCIDENT SOURCE TERM..............................................113.1Fission Product Inventory...........................................113.2Release Fractions................................................12 3.3Timing of Release Phases..........................................13 3.4Radionuclide Composition..........................................133.5Chemical Form..................................................14 3.6Fuel Damage in Non-LOCA DBAs....................................144.DOSE CALCULATIONAL METHODOLOGY..................................144.1Offsite Dose Consequences........................................144.2Control Room Dose Consequences...................................154.3Other Dose Consequences.........................................184.4Offsite Acceptance Criteria.........................................18 4.5Control Room Acceptance Criteria....................................18 4.6Other Acceptance Criteria..........................................185.ANALYSIS ASSUMPTIONS AND METHODOLOGY............................195.1General Considerations............................................195.2Accident-Specific Assumptions......................................205.3Meteorology Assumptions..........................................20D. IMPLEMENTATION........................................................21 REFERENCES..............................................................21 APPENDICES ivA.Assumptions for Evaluating the Radiological Consequences of a LWRLoss-of-Coolant Accident ...............................................A-1B.Assumptions for Evaluating the Radiological Consequences of a FuelHandling Accident.....................................................B-1C.Assumptions for Evaluating the Radiological Consequences of a BWRRod Drop Accident....................................................C-1D.Assumptions for Evaluating the Radiological Consequences of a BWR Main Steam Line Break Accident..........................................D-1E.Assumptions for Evaluating the Radiological Consequences of a PWR Steam Generator Tube Rupture Accident...................................E-1F.Assumptions for Evaluating the Radiological Consequences of a PWR Main Steam Line Break Accident..........................................F-1G.Assumptions for Evaluating the Radiological Consequences of a PWR Locked Rotor Accident.................................................G-1H.Assumptions for Evaluating the Radiological Consequences of a PWR Rod Ejection Accident..................................................H-1I.Acronyms............................................................I-1 REGULATORY ANALYSIS...................................................RA-1 1A. INTRODUCTIONThis guide provides guidance to licensees of operating power reactors on acceptablemethods and assumptions for performing evaluations of fission product releases and radiological consequences of several postulated light-water reactor design basis accidents.

The NRC is proposing this new guide as a means to provide guidance to licensees for license amendment requests that, in whole or part, seek to modify the licensing basis methodology and assumptions for performing evaluations of fission product releases and radiological consequences of several postulated light-water reactor design basis accidents. It describes the sources; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and the content of submittals acceptable to the NRC staff. 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, "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. General Design Criterion (GDC-19), "Control Room," of Appendix A, "General DesignCriteria for Nuclear Power Plants," to 10 CFR Part 50 establishes criteria for a control room and requires means for remote plant shutdown. GDC-19 also requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures more than 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. 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, 1 As defined in 10 CFR 50.2, design bases means information that identifies the specific functions to beperformed by a structure, system, or component of a facility and the specific values or ranges of valueschosen 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.

2including the source term, are to be included in the FSAR and become part of the facilitydesign basis.

1 Since the publication of TID-14844, significant advances have been made inunderstanding 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 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. This guide is not applicable to facilities that use the alternative source term as described in 10 CFR 50.67, "Accident Source Term." Guidance for the alternative source term is provided in Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (Ref. 3). This regulatory guide does not apply to applicants for a construction permit, a designcertification, or a combined license who do not reference a standard design certification and who applied after January 10, 1997, nor to licensees authorized to use an alternative source term (AST) under 10 CFR 50.67. These applicants and licensees are required by regulation to calculate offsite dose in units of total effective dose equivalent (TEDE). TEDE criteria are expected to be used with the AST and not with results calculated according to TID-14844 (Ref. 1). Therefore, because this guide pertains to the TID-14844 source terms and the corresponding whole body and thyroid criteria, it does not apply to applicants and licensees who are required to use the TEDE criteria.The information collections contained in this draft regulatory guide are covered by therequirements of 10 CFR Part 50, which were approved by the Office of Management and Budget, approval number 3150-0011. If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection. B. DISCUSSIONAn accident source term is intended to be representative of a major accident involvingsignificant 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 defense in depth is achieved, defense in depth continues to be an effective way to account for uncertainties in 3equipment 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. The NRC's traditional methods for calculating the radiological consequences of designbasis accidents were described in a series of regulatory guides and Standard Review Plan (SRP) chapters. That guidance was developed to be consistent with the release fractions and timing from the TID-14844 source term and the whole body and thyroid doses stated in 10 CFR 100.11. The guidance contained in this regulatory guide will supersede corresponding radiological analysis assumptions provided in other regulatory guides when used in conjunction with guidance that is being developed in a draft guide, Draft Regulatory Guide DG-1114, "Control Room Habitability at Nuclear Power Reactors," which will be published soon.

The affected guides will not be withdrawn as they may still be used at the option of licensees.

Specifically, the guidance in the two draft guides, when final, could be used instead of the following regulatory guides:Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Loss of Coolant Accident for Boiling Water Reactors" (Ref. 5)Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Loss of Coolant Accident for Pressurized Water Reactors" (Ref. 6) Regulatory Guide 1.5, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Steam Line Break Accident for Boiling Water Reactors" (Ref. 7)Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors" (Ref. 8)Regulatory Guide 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident forPressurized Water Reactors" (Ref. 9)This guide primarily addresses design basis accidents, such as those addressed inChapter 15 of typical final safety analysis reports. 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 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 POSITION1.GENERIC CONSIDERATIONS 1.1 Safety MarginsThe proposed uses of this guide and the associated proposed facility modifications andchanges 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. Changes, or the net effects of multiple changes, that result in a reduction in safety margins may require prior NRC approval. Licensees may use 410 CFR 50.59 and its supporting guidance to assess safety margins related to facilitymodifications and changes to procedures that are described in the Updated Final Safety Analysis Report.1.2 Defense in DepthThe proposed uses of this guide and the associated proposed facility modifications andchanges 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 for facilities to which GDC criteria apply.

Modifications proposed for the facility generally should not create a need for compensatory programmatic activities, such as reliance on manual operator actions, use of potassium iodide as a prophylactic drug, or self-contained breathing apparatus.Proposed modifications that seek to downgrade or remove required engineeredsafeguards 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. The radiological analyses provide a fundamental basis upon which a significant portionof the facility design is based. Additionally, many aspects of facility operation derive from radiological design analyses. 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. 1.3Scope of Required Analyses1.3.1 Design Basis Radiological AnalysesA fundamental commitment required for application of the methodology in this guide isto perform an assessment of each applicable accident. The analyses should include accidents mentioned in this guide, supplemented by those in the FSAR and other licensee documents, as appropriate. The performance of these assessments will determine the limiting event with respect to offsite and control room dose. Some licensees have evaluated the control room dose only for the DBA LOCA, which is typically the limiting event for offsite radiological releases. The DBA LOCA is generally the large break (LB) LOCA event analysis.

Other events may be analyzed as part of the design basis accident evaluation for the facility.

Although these events may have been shown to be nonlimiting with respect to offsite dose, control room dose assessments for these events are required to identify the limiting event for the GDC-19 control room dose design criterion. There are several regulatory requirements for which compliance is demonstrated, inpart, 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) 5Emergency Response Facility Habitability (Paragraph IV.E.8 of Appendix E to 10CFR Part 50)Environmental Reports (10 CFR Part 51)Facility Siting (10 CFR 100.11)There may be other areas in which the technical specification bases and variouslicensee commitments refer to specific evaluations. 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 GuidanceFacility modification should be supported by evaluations of all significant radiologicaland 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 methodology utilized. The scope and extent of the re-evaluation will be a function of the specific proposed facility modification or the change in methodology. The NRC staff expects 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 recalculated, all affected assumptions and inputs should be updated. A 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.1.3.3 Use of Sensitivity or Scoping AnalysesIt may be possible to demonstrate by sensitivity or scoping evaluations that existinganalyses 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 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. 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 6(EAB), low population zone (LPZ), or control room dose unless there is a clear and defensiblebasis for doing so.1.4Risk ImplicationsThis guide provides guidance only on the regulatory assumptions that licenseesshould make in their calculation of the radiological consequences of design basis accidents.

These assumptions have no direct affect on the probability of the design basis accident initiator. These analyses assumptions cannot increase the core damage frequency (CDF) or the large early release frequency (LERF). However, facility modifications made possible by the use of this guide could have an impact on risk. If the proposed implementation of this guide 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 thatseek 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 thedesign 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. 10).1.5Submittal RequirementsAccording to 10 CFR 50.90, an application for an amendment must fully describe thechanges desired and 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. 11), provides additional guidance. The NRC staff's finding as to whether an amendment is to be approved or rejected is based in part on the licensee's analyses, 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 computercode, 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.1.6Final Safety Analysis Report RequirementsRequirements for updating the facility's FSAR are in 10 CFR 50.71, "Maintenance ofRecords, Making of Reports." The regulations in 10 CFR 50.71(e) require that the FSAR be updated to include the effects of all changes made in the facility or procedures described in the FSAR and all safety analyses and evaluations performed by the licensee in support of approved license amendments or in support of conclusions that changes did not require a 7license amendment in accordance with 10 CFR 50.59. The affected radiological analysisdescriptions in the FSAR should be updated to reflect the replacement of the design basis changes to the methodology and inputs. The analysis descriptions should contain sufficient detail to identify the methodologies used, significant assumptions and inputs, and numeric results. Regulatory Guide 1.70 (Ref. 11) provides additional guidance. The descriptions of superseded analyses should be removed from the FSAR in the interest of maintaining a clear design basis. The NRC staff reviews licensee amendment requests to ensure the proposed changewill maintain an adequate level of protection of public health and safety. The NRC staff accomplishes these reviews by evaluating the information submitted in the amendment request against the current plant design basis as documented in the FSAR, staff safety evaluation reports (SERs), regulatory guidance, other licensee commitment, and staff experience gained in approving similar requests for other plants. The NRC staff bases its finding that the amendment is acceptable on its assessment of the licensee's analysis, since it is the licensee's analysis that becomes part of the facility's design basis. Licensees should ensure that adequate information, including analysis assumptions, inputs, and methods, are presented in the submittal to support the staff's assessment. The NRC staff's assessment may include performance of independent analyses to confirm the licensee's conclusion.

Licensees should expect an NRC staff effort to resolve critical differences in analysis assumptions, inputs, and methods used by the licensee and those deemed acceptable to the NRC staff. 2.DOSE ANALYSIS MODELS2.1 Radiological ConsequencesThe radiological consequences of an accident in a nuclear reactor depend on thequantity of the radioactive material that escapes to the environment or enters the control room. As the radioactivity is transported through the containment and other buildings, credit is given for several natural and engineered removal mechanisms. Within compartments, these removal mechanisms include sprays, natural deposition, leakage, natural and forced convection, filters, and suppression pools. This section describes the general equations used to model the transport and removal of fission products between compartments, the calculation of activities in the environment, and the calculation of offsite and compartment doses. 2.2 Activities in Compartments Without Inflow From the EnvironmentThe following balance equation models the rate of change of activity of a nuclide in acompartment r. An example of a compartment that is typically without inflow from the environment is a reactor containment building. (1)M r k 1 k N 1 j r rj k tkr r a a dt dawhere: a r =activity of a nuclide in compartment r at time t, Ci a k =activity of a nuclide in compartment k at time t, CiN =the number of removal processes 8M =the number of compartments modeled

!rj =removal constant of the j th removal process internal to compartment r, i.e., decay,plateout, filtration, spray in containment, flow out of compartment r, sec

-1!tkr =transfer constant from region k to compartment r, i.e., flow rate from compartmentk to compartment r divided by the volume of compartment k, sec

-1For halogens, a more specific form of Equation 1 may be written to account forremoval mechanisms that are chemically species-specific, e.g., filter efficiencies for particulate, elemental, and organic iodine. For these situations, Equation 1 can be rewritten to define the activity and removal constants on a per nuclide and species basis. 2.3 Activities in the EnvironmentEquation 1 is solved for the time-dependent activity in each compartment. Therelease rate from M compartments to the environment is given by Equation 2. The activity in the environment from each compartment is given by Equation 3.(2)M 1 k ke R R(3)k keU ke keF k ke a Q f 1 Q V 1 Rwhere: a k=activity of compartment k, Ci f ke=filter removal efficiency fraction for a filter between compartment k and theenvironment, dimensionless Q keF=filtered flow from compartment k to the environment, m³/sec Q keU=unfiltered flow from compartment k to the environment, m³/secR=release rate of activity from M compartments to the environment, Ci/sec

R ke=release rate of activity from compartment k to the environment, Ci/sec V k=free volume of compartment k, m 32.4 Activities in Compartments That Intake Only Outside Contaminated Air Equation 4 models compartments that intake radioactivity transported to thecompartment via only atmospheric dispersion. Control rooms or technical support centers that do not intake radioactivity directly from other buildings are examples of these compartments.(4)M r k 1 k N 1 j r rj ke ker erU erF r a R Q Q FQ dt dawhere: F= filter nonremoval fraction of the intake from the environment to compartment r(i.e., 1-filter removal efficiency fraction), dimensionless 9 QerF=filtered intake flow rate from the environment to compartment r, m³/sec QerU= unfiltered intake flow rate from the environment to the compartment r, m³/sec

[/Q]ker=atmospheric dispersion factor from compartment k to intake of compartment r, sec/m³

Examples of removal process (!rj) typically modeled for control rooms are given below:

!r1=exhaust rate from the control room to the environment, sec

-1 = Q E/V r, where Q E is the exhaust flow rate from the control room to the environment (Q E = QerF +QerU), m 3/sec, and V r is the free volume of the control room, m 3 !r2=nuclide decay constant, sec

-1 !r3=recirculation removal rate, sec

-1 = (Q R/V r) x f R where Q R is the recirculation flowrate in the control room, m 3/sec, V r is the free volume of the control room, m 3 , and f R is the recirculation filter removal efficiency fraction, dimensionless 2.5 Integrated Activity Released Into the Environment The integrated activity (Curies) released into the environment over the time interval jfrom time t0, to t 1, IAR j, is given by the following equation. In calculating IAR j, no credit istaken for cloud depletion by ground deposition or radioactive decay during transit to the exclusion area boundary or the LPZ outer boundary.(5)1 0 t t j Rdt IAR2.6 Integrated Activity in a CompartmentThe integrated activity (Ci-sec) in a compartment k over the time interval j from time t 0to t 1 , IA kj, is determined by the expression:(6)1 0 t t k kj dt a IA2.7 Offsite Doses The following equations give the models used to calculate offsite doses. Equations forcalculating thyroid and whole body doses are given. Offsite thyroid doses are calculated using the equation: (7)N 1 i T 1 j j j ij i TH TH)Q ()BR ()IAR ()DCF (DAssuming a semi-infinite cloud of photon emitters, offsite whole body doses arecalculated using the equation: (8)N 1 i T 1 j j ij i B B)Q ()IAR ()DCF (D!""where:

2 Control room envelopes may be composed of more than one room or subcompartment. If those roomscontain shielding that blocks the majority (99% or greater) of whole body dose outside the room, the geometry factor is calculated using the free volume of the largest subcompartment.

10(BR)j= breathing rate during time interval j, m³/sec D TH= offsite thyroid dose via inhalation during time interval j, rem D" B=offsite whole body dose during time interval j, rem (DCF TH)i= thyroid dose conversion factor via inhalation for nuclide i, rem/Ci (DCF" B)i=photon body dose conversion factor for nuclide i, rem-m³/Ci-sec(IAR)ij= integrated activity of nuclide i released during the time interval j, CiN= number of nuclides T=number of time intervals over which (IAR) is calculated

(/Q)j= offsite atmospheric dispersion factor during time interval j, sec/m³ 2.8 Compartment Doses Compartment thyroid doses via inhalation pathway are calculated using the followingequation: (9)N 1 i T 1 j j j ij k i TH k k TH)BR (O)IA ()DCF (V 1)D (Because of the finite size of a compartment, the whole body photon doses in acompartment caused by the radioactive cloud will be substantially less than the doses caused by immersion in an infinite cloud of photon emitters. The finite cloud photon doses are calculated using Murphy's method, which models the compartment as a hemisphere. The following equation is used: (10)N 1 i T 1 j j ij k i B k k k B O)IA ()DCF (V GF 1)D (The beta skin doses in a compartment are calculated using the following equation:(11)N 1 i T 1 j j ij k i S k k S O)IA ()DCF (V 1)D (where: BR j=breathing rate in time interval j, m 3/sec (D TH)k=compartment k thyroid dose via inhalation, rem (D" B)k=compartment k whole body dose, rem (D#S)k=compartment k beta skin dose, rem (DCF#S)i=beta skin dose conversion factor for nuclide i, rem-m³/Ci-sec GF k=dose reduction due to the compartment geometry correction factor352/V k0.338, dimensionless 2 (see Regulatory Position 4.2.7) (IA k) ij=integrated activity concentration in compartment k, for nuclide i duringtime interval j, Ci-sec O j=compartment occupancy fraction during time interval jT=number of time intervals over which (IA) is calculated

V k=compartment k free volume, m 3

3 The uncertainty factor used in determining the core inventory should be that value provided in Appendix Kto 10 CFR Part 50, typically 1.02. A value lower than 1.02, but not less than 1.00 (correlates to the licensed power level) may be used provided the proposed alternative value has been demonstrated to account for uncertainties caused by power level instrumentation error.

4 Some plants evaluate the radiological consequences of a reactor head drop accident. For these analysesit is appropriate to use the core average inventory to assess the consequences of this accident.

5 The release fractions listed here have been determined to be acceptable for use with currently approvedLWR fuel with a peak rod burnup up to 62,000 MWD/MTU. The data in this section may not be applicable to cores containing mixed oxide (MOX) fuel.

113.ACCIDENT SOURCE TERMThis Regulatory Position provides a source term that is acceptable to the NRC staff. It provides guidance on the fission product inventory, release fractions, timing of the release, radionuclide composition, chemical form, and the fuel damage for DBAs. 3.1Fission Product InventoryThe inventory of fission products in the reactor core and available for release to thecontainment should be based on the maximum full-power operation of the core with, as a minimum, currently licensed values for fuel enrichment, fuel burnup, and an assumed core power equal to the current licensed rated thermal power times the emergency core cooling system (ECCS) evaluation uncertainty.

3 These parameters should be examined to maximizefission product inventory. 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.

The core inventory should be determined using an appropriate isotope generation and depletion computer code such as ORIGEN2 (Ref. 12) or ORIGEN-ARP (Ref. 13). Core inventory factors (Ci/MWt) provided in TID-14844 (Ref. 1) 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.For the DBA LOCA, all fuel assemblies in the core are assumed to be affected andthe core average inventory should be used.

4 Further assumptions are in Appendix A to thisguide. 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 postulatedto 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 shut down, e.g.,

a fuel handling accident, radioactive decay from the time of shutdown may be modeled.3.2Release FractionsThe core inventory release fractions, 5 by radionuclide groups, for DBA LOCAs arelisted in Table 1 for boiling-water reactors (BWRs) and pressurized-water reactors (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 gapfor the various radionuclides are given in Table 2. The release fractions from Table 2 are 6 If containment sprays are not modeled mechanistically, such as in Revision 2 of Standard Review Plan(Ref. 14) Section 6.5.2, one-half of the equilibrium radioactive iodine inventory released into the containment atmosphere may be assumed to be deposited on the walls of the containment. The net value of core inventory available for release from containment would, therefore, be 0.25 for a nonmechanistic spray representation. Please note that Revision 2 of SRP Section 6.5.2 erroneously states that 25% of the equilibrium radioactive iodine inventory developed from maximum full-power operation of the core should be assumed to be immediately available for the leakage from the primary reactor system. This value should be 50% of the equilibrium radioactive iodine inventory. Revision 2 erroneously accounted twice for the iodine deposited on the wall of the containment.

7 The release fractions listed here have been determined to be acceptable for use with currently approvedLWR 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 rods with burnups that exceed 54 GWD/MTU.

As an alternative, fission gas release calculations performed using NRC-approved 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.

12used in conjunction with the calculated fission product inventory and the maximum core radialpeaking factor. For non-LOCA DBAs when fuel melt is postulated, the core inventory release fractions, by radionuclide groups, are listed in Table 2 for BWRs and PWRs. Table 1BWR And PWR Core Inventory Fraction Released Into Containment AtmosphereGroupRelease FractionNoble Gases1.0Halogens 60.5Table 2 7Non-LOCA Fraction of Fission Product Inventory in Gap GroupFractionI-1310.08Kr-850.10 Other Noble Gases0.05 Other Halogens0.053.3Timing of Release PhasesFor LOCA DBAs, the core activity released is assumed to be immediately available forrelease from containment. For non-LOCA DBAs in which fuel damage is projected, the activity available for release from the fuel is assumed to be immediately available for release from the containment or the building where the fuel is damaged.3.4Radionuclide CompositionTable 3 lists the elements in each radionuclide group that should be considered indesign basis analyses.

13Table 3Radionuclide GroupsGroupElementsNoble GasesXe, KrHalogensI3.5Chemical FormOf the radioiodine released from the reactor coolant system (RCS) to the containmentin a postulated accident, 5% of the iodine released should be assumed to be particulate iodine, 91% elemental iodine, and 4% organic iodide. This includes releases from the gap and the fuel pellets. The same chemical form is assumed in releases from fuel pins in fuel handling accidents (FHAs) and from releases from the fuel pins through the reactor coolant system 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 A through H to this regulatory guide provide additional details.3.6 Fuel Damage in Non-LOCA DBAsThe amount of fuel damage caused by non-LOCA design basis events should beanalyzed 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.For the postulated main steam line break, steam generator tube rupture, and lockedrotor accidents, the amount of fuel damage should be evaluated assuming that the highest worth control rod is stuck at its fully withdrawn position. The amount of fuel damage caused by a FHA is addressed in Appendix B to thisguide.4. DOSE CALCULATIONAL METHODOLOGYThis Regulatory Position provides a dose calculational methodology that is acceptableto the NRC staff. It provides guidance on the calculation of offsite and onsite consequences and on offsite and control room acceptance criteria.4.1 Offsite Dose ConsequencesThe following assumptions should be used in determining the doses for personslocated at or beyond the boundary of the exclusion area (EAB):4.1.1The dose calculations should determine the thyroid and whole body doses.4.1.2The exposure-to-thyroid factors for inhalation of radioactive material should bederived from the data provided in ICRP Publication 30, "Limits for Intakes of Radionuclides byWorkers" (Ref. 15). Table 2.1 of Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, 8 Licensees who use these dose conversion factors in accident calculations should determine whether thiswill impact the facility's technical specification definition for dose equivalent I-131.

14Submersion, and Ingestion" (Ref. 16), provides tables of conversion factors acceptable to theNRC staff. The factors in the column headed "thyroid" should be used.

8 4.1.3For the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the breathing rate of persons offsite should be assumed tobe 3.5 x 10

-4 cubic meters per second. From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, thebreathing rate should be assumed to be 1.8 x 10

-4 cubic meters per second. After that anduntil the end of the accident, the rate should be assumed to be 2.3 x 10

-4 cubic meters per second.4.1.4The whole body doses should be calculated assuming submergence in semi-infinite cloud assumptions with appropriate credit for attenuation by body tissue. Table III.1 ofFederal Guidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 17), provides external conversion factors acceptable to the NRC staff. The factors in the column headed "effective" yield doses correspond to the whole body dose. The use of effective dose-conversion factors (DCFs) as a surrogate for whole body DCFs is appropriate because of the uniform body exposure associated with semi-infinite cloud dose modeling. 4.1.5The whole body and thyroid doses should be determined for an individual atthe most limiting EAB location. The maximum EAB dose for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following thestart of the radioactivity release should be determined and used in determining compliance with the dose criteria given in Table 4. 4.1.6The whole body and thyroid doses should be determined for the most limitingreceptor at the outer boundary of the LPZ and should be used in determining compliance withthe dose criteria in Table 4.4.1.7No correction should be made for depletion of the effluent plume by depositionon the ground.4.2 Control Room Dose ConsequencesThe following guidance should be used in determining the whole body, thyroid, andskin doses for persons located in the control room envelope.4.2.1The whole body, thyroid, and skin dose analyses should consider all sources ofradiation that will cause exposure to control room personnel. The applicable sources willvary from facility to facility, but typically will include:Contamination of the control room envelope atmosphere by the intake or infiltration of the radioactive material contained in the radioactive plume released from the facility,Contamination of the control room envelope atmosphere by the intake or infiltration of airborne radioactive material from areas and structures adjacent to the control room envelope, 15Table 4EAB and LPZ Accident Dose CriteriaDose Criteria (rem)Accident or CaseWhole BodyThyroid Analysis Release DurationLOCA25 rem300 rem30 days for containment, ECCS,and MSIV (BWR) leakageBWR Main Steam Line BreakInstantaneous puffFuel Damage or Pre-incidentSpike25 rem300 remEquilibrium Iodine Activity2.5 rem30 remBWR Rod Drop Accident6.3 rem75 rem24 hours PWR Steam Generator TubeRuptureAffected SG: time to isolate;Unaffected SG(s): until cold shutdown is establishedFuel Damage or Pre-incidentSpike25 rem300 remCoincident Iodine Spike2.5 rem30 remPWR Main Steam Line BreakUntil cold shutdown isestablishedFuel Damage or Pre-incidentSpike25 rem300 remCoincident Iodine Spike2.5 rem30 remPWR Locked Rotor Accident2.5 rem30 remUntil cold shutdown isestablishedPWR Rod Ejection Accident6.3 rem75 rem30 days for containmentpathway; until cold shutdown is established for secondary pathwayFuel Handling Accident6.3 rem75 rem2 hoursThe column labeled "Analysis Release Duration" is a summary of the assumed radioactivity releasedurations identified in the individual appendices to this guide. Refer to these appendices for complete descriptions of the release pathways and durations.Radiation shine from the external radioactive plume released from the facility,Radiation shine from radioactive material in the reactor containment, Radiation shine from radioactive material in systems and components inside orexternal to the control room envelope, e.g., radioactive material buildup in recirculation filters.4.2.2The radioactive material releases and radiation levels used in the control roomenvelope dose analysis should be determined using the same source term, in-plant transport,and release assumptions used for determining the EAB and the LPZ dose values, unless these assumptions would result in nonconservative results for the control room envelope.

9 The iodine protection factor (IPF) methodology of Reference 18 may not be adequately conservative forall 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. 19) and RADTRAD (Ref. 20) incorporate suitable

methodologies.

1. 0 This occupancy is modeled in the/Q values determined in Reference 18 and should not be credited twice. TheARCON96 Code (Ref. 23) does not incorporate these occupancy assumptions, making it necessary to apply this correction in the dose calculations.

164.2.3The models used to transport radioactive material into and through the controlroom envelope, 9 and the shielding models used to determine radiation dose rates fromexternal sources, should be structured to provide suitably conservative estimates of the exposure to control room personnel.4.2.4Credit for engineered safety features that mitigate airborne radioactive materialwithin the control room envelope may be assumed. Such features may include control roomisolation or pressurization, or intake or recirculation filtration. Refer to Section 6.5.1, "ESF Atmospheric Cleanup System," of the SRP (Ref. 14); Regulatory Guide 1.52, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants" (Ref. 21); and Generic Letter 99-02 (Ref. 22) for guidance. The control room envelope 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. 4.2.5Credit should generally not be taken for the use of personal protectiveequipment or use of potassium iodide (KI) as a thyroid prophylactic drug. 4.2.6The dose receptor for these analyses is the hypothetical maximum exposed individual who is present in the control room envelope for 100% of the time during the first 24hours after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days to 30 days.

10 For the duration of the event, the breathing rate of this individual shouldbe assumed to be 3.5 x 10

-4 cubic meters per second.4.2.7Control room envelope doses should be calculated using dose conversionfactors identified in Regulatory Position 4.1 above for use in offsite dose analyses. Thecalculation should consider all radionuclides that are significant with regard to dose consequences and the release of radioactivity. The whole body dose from photons may be corrected for the difference between finite cloud geometry in the control room envelope and the semi-infinite cloud assumption used in calculating the dose conversion factors using a compartment geometry correction factor. This factor is incorporated in Equation 10 of Regulatory Position 2.8. This correction is not applied to the beta skin dose estimates, as the range of beta particles in air is less than the typical control room dimensions. The skin dose DCFs presented in Federal Guidance Report 12 (Ref. 17) are based on both photon and beta emissions. Without the geometry correction, the photon dose component will be overestimated. If the geometry correction is included, the beta component will be underestimated. DOE/EH-0070, "External-Dose Rate Conversion Factors for Calculation of Dose to the Public" (Ref. 24), tabulates the beta and photon skin dose DCFs separately.

174.2.8Skin doses should be calculated using the factors in the column headed "Skin"in Table III.1 of Federal Guidance Report 12 (Ref. 17).4.3Other Dose ConsequencesThe guidance provided in Regulatory Positions 4.1 and 4.2 should be used, asapplicable, in re-assessing the radiological analyses identified in Regulatory Position 1.3.1, such as those in NUREG-0737 (Ref. 2). 4.4Offsite Acceptance CriteriaThe radiological criteria for the EAB and the outer boundary of the LPZ are given in10 CFR 100.11. 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. For events with a higher probability of occurrence, postulated EAB and LPZ doses should not exceed the criteria tabulated in Table 4. The criteria provided in Table 4 are the same criteria provided in the Standard Review Plan (Ref. 14). For PWRs with steam generator alternative repair criteria, different dose criteria may apply to steam generator tube rupture and main steam line break analyses as suggested by guidance being developed in Draft Regulatory Guide DG-1074, "Steam Generator Integrity" (Ref. 25). 4.5Control Room Acceptance CriteriaThe following guidelines may be used in lieu of those provided in SRP 6.4 (Ref. 14)when showing compliance with the dose guidelines in GDC-19 of Appendix A to 10 CFR Part

50. The following guidelines relax the thyroid acceptance criteria from that given in SRP 6.4.

This relaxation from 30 to 50 rem-thyroid is based on a change to 0.03 in the thyroid organ dose weighting factor given in 10 CFR 20.1003. Although this change gives an equivalent thyroid dose of 167 rem-thyroid, 10 CFR 20.1201 limits organ dose to 50 rem annually. The release duration is specified in Table 4. The exposure period is 30 days for all accidents.

The criterion in GDC-19 applies to all accidents.Whole body5 remThyroid50 rem Beta or skin30 rem4.6 Other Acceptance CriteriaThe acceptance criteria for the various NUREG-0737 (Ref. 2) items generallyreference GDC-19 from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC-19. These criteria remain unchanged except for the thyroid dose limit as stated in Regulatory Position 4.5. Before the General Design Criteria were established in 10 CFR Part 50, these criteriaexisted in draft form. Some of the facilities that were licensed during this time period committed to various draft criterion for control room habitability. These commitments may be different from GDC-19.

185.ANALYSIS ASSUMPTIONS AND METHODOLOGY5.1General Considerations5.1.1 Analysis QualityThe analyses required by 10 CFR 100.11 and GDC-19 in Appendix A to 10 CFRPart 50 and any re-analyses of these analyses required by 10 CFR 50.34 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 ofassumptions 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 modeled 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 FeaturesCredit may be taken for accident mitigation features that are classified as safetyrelated, 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.

Design basis delays in actuation of these features should be considered, especially for features that rely on manual operator intervention.5.1.3 Assignment of Numeric Input ValuesThe numeric values that are chosen as inputs to the dose analyses required byregulations and described in Regulatory Position 5.1.1 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 11 Note that for some parameters, the technical specification value may be adjusted for analysis purposesby factors provided in other regulatory guidance. For example, ESF filter efficiencies are based on the guidance in Regulatory Guide 1.52 (Ref. 21) and in Generic Letter 99-02 (Ref. 22) rather than the surveillance test criteria in the technical specifications. Generally, these adjustments address possible changes in the parameter between scheduled surveillance tests.

19specifications.

11 If a range of values or a tolerance band is specified, the value that wouldresult 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. 5.2Accident-Specific AssumptionsThe appendices to this regulatory guide provide accident-specific assumptions thatare acceptable to the staff for performing analyses that are described in Regulatory Position 5.1.1. The DBAs addressed in these appendices 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 changes to the facility or to the radiological analyses. The NRC staff has determined that the analysis assumptions in the appendices to thisguide 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 or updated technical analyses. 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 the synergy among the assumptions in this guide.The NRC is committed to using probabilistic risk analysis (PRA) insights in itsregulatory activities and will consider licensee proposals for changes in analysis assumptions that reflect 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.5.3Meteorology AssumptionsAtmospheric dispersion values (/Q) for the EAB, the LPZ, and the control room that were approved by the staff during initial facility licensing or in subsequent licensingproceedings may be used in performing the radiological analyses identified by this guide provided such values remain relevant to the particular accident, its release points, andreceptor location. Methodologies that have been used for determining /Q values aredocumented in Regulatory Guides 1.3, 1.4, and 1.145, "Atmospheric Dispersion Models forPotential Accident Consequence Assessments at Nuclear Power Plants" (Refs. 5, 6, and 26),

and in the Murphy-Campe paper, "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19" (Ref. 18).

20References 18 and 26 should be used if the FSAR /Q values are to be revised or ifvalues are to be determined for new release points or receptor distances. For stackreleases, 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. 27) implements Regulatory Guide 1.145 (Ref. 26), and its use is acceptable to the NRC staff. Draft Regulatory Guide DG-1111, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants" (Ref. 28) is being developed to provide guidance ondetermining control room /Q values. The NRC computer code ARCON96 (Ref. 23) may beused in determining control room /Q values. Meteorological data collected in accordancewith the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident Q values. Additional guidance is provided inRegulatory Guide 1.23, "Onsite Meteorological Programs" (Ref. 29). All changes in

$/Qanalysis methodology should be reviewed by the NRC staff.D. IMPLEMENTATIONThe purpose of this section is to provide information to applicants and licenseesregarding the NRC staff's plans for using this draft regulatory guide. No backfitting is intended or approved in connection with the issuance of this guide. This draft guide has been released to encourage public participation in itsdevelopment. 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 to be described in the active guide reflecting public comments will be used in the evaluation of submittals in connection with radiological consequences at nuclear power reactors.

21 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 ReactorSites," 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, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents at Nuclear Power Reactors," Regulatory Guide 1.183, July 2000.4.USNRC, "Use of Probabilistic Risk Assessment Methods in Nuclear Activities: FinalPolicy Statement," Federal Register, Volume 60, page 42622 (60 FR 42622) August16, 1995.5.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Loss of Coolant Accident for Boiling Water Reactors." Regulatory Guide 1.3, Revision 2, June 1974.6.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Loss of Coolant Accident for Pressurized Water Reactors," Regulatory Guide 1.4, Revision 2, June 1974.7.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Steam Line Break Accident for Boiling Water Reactors," Regulatory Guide 1.5, March 1971.8.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," Regulatory Guide 1.25, March 1972.9.USNRC, "Assumptions Used for Evaluating a Control Rod Ejection Accident forPressurized Water Reactors," Regulatory Guide 1.77, May 1974.10.USNRC, "An Approach for Using Probabilistic Risk Assessment in Risk-InformedDecisions on Plant-Specific Changes to the Licensing Basis," Regulatory Guide 1.174, July 1998.11.USNRC, "Standard Format and Content of Safety Analysis Reports for Nuclear PowerPlants (LWR Edition)," Regulatory Guide 1.70, Revision 3, November 1978.12.A.G. Croff, "A User's Manual for the ORIGEN2 Computer Code," ORNL/TM-7175, OakRidge National Laboratory, July 1980.13.S.M. Bowman and L.C. Leal, "The ORIGENARP Input Processor for ORIGEN-ARP,"Appendix F7.A in SCALE: A Modular Code System for Performing Standardized Analysesfor Licensing Evaluation, NUREG/CR-0200, USNRC, March 1997.14.USNRC, "Standard Review Plan for the Review of Safety Analysis Reports for NuclearPower Plants," NUREG-0800, September 1981 (or updates of specific sections).

2215.ICRP, "Limits for Intakes of Radionuclides by Workers," ICRP Publication 30, 1979.16.K.F. Eckerman et al., "Limiting Values of Radionuclide Intake and Air Concentrationand Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency, 1988. 17.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. 18.K.G. Murphy and K.W. Campe, "Nuclear Power Plant Control Room VentilationSystem Design for Meeting General Criterion 19," published in Proceedings of 13thAEC Air Cleaning Conference, Atomic Energy Commission (now USNRC), August 1974.19.USNRC, "Computer Codes for Evaluation of Control Room Habitability (HABIT V1.1),"Supplement 1 to NUREG/CR-6210, November 1998.20.S.L. Humphreys et al., "RADTRAD: A Simplified Model for Radionuclide Transport andRemoval and Dose Estimation," NUREG/CR-6604, USNRC, April 1998.21.USNRC, "Design, Inspection, and Testing Criteria for Air Filtration and AdsorptionUnits of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," Regulatory Guide 1.52, Revision 3, June

2001.22.USNRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal," NRC GenericLetter 99-02, June 3, 1999.23.J.V. Ramsdell and C.A. Simonen, "Atmospheric Relative Concentrations in BuildingWakes," NUREG-6331, Revision 1, USNRC, May 1997.24. USDOE, "External-Dose Rate Conversion Factors for Calculation of Dose to thePublic," DOE/EH-0070, U.S. Department of Energy, July 1988.25. USNRC, "Steam Generator Tube Integrity," Draft Regulatory Guide DG-1074,December 1998.26.USNRC, "Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants," Regulatory Guide 1.145, Revision 1, November 1982.27.T.J. Bander, "PAVAN: An Atmospheric Dispersion Program for Evaluating DesignBasis Accidental Releases of Radioactive Materials from Nuclear Power Stations,"

NUREG-2858, USNRC, November 1982.28.USNRC, "Atmospheric Relative Concentrations for Control Room RadiologicalHabitability Assessments at Nuclear Power Plants," Draft Regulatory Guide DG-1111, December 2001. 29.USNRC, "Onsite Meteorological Programs," Regulatory Guide 1.23, February 1972.

23 1 Note that Revision 2 of Standard Review Plan Chapter 6.5.2 erroneously states that 25% of theequilibrium radioactive iodine inventory developed from maximum full-power operation of the core should be assumed to be immediately available for the leakage from the primary reactor system. This value should be 50% of the equilibrium radioactive iodine inventory. Revision 2 erroneously accounted twice for the iodine deposited on the wall of the containment. A-1Appendix AASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCESOF A LWR LOSS-OF-COOLANT ACCIDENT The assumptions in this appendix are acceptable to the NRC staff for evaluating theradiological 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 50defines 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 emergency core cooling system 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. 1. SOURCE TERM ASSUMPTIONSAcceptable assumptions regarding core inventory and the release of radionuclidesfrom the fuel are provided in Regulatory Position 3 of this guide.2. ASSUMPTIONS ON TRANSPORT IN PRIMARY CONTAINMENTAcceptable assumptions related to the transport, reduction, and release of radioactivematerial in and from the primary containment in PWRs or the drywell in BWRs are as follows.2.1At the start of the accident, the radioactivity released from the fuel should be assumedto mix instantaneously and homogeneously throughout the free air volume of theprimary containment in PWRs or the drywell in BWRs. 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. 2.2Reduction in airborne radioactivity in the containment by natural deposition within thecontainment may be credited. An acceptable model for removal of iodine andparticulates is 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). 2.3Reduction in airborne radioactivity in the containment by containment spray systemsthat have been designed and are maintained in accordance with Chapter 6.5.2 of the SRP 1 (Ref. A-1) may be credited. An acceptable model for the removal of iodine andparticulates is described in Chapter 6.5.2 of the SRP.

A-2The evaluation of the containment sprays should address areas within the primarycontainment 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 region volume per hour, unless other rates are justified. On a case-by-case basis, containment mixing rates determined by the cooldown rate in the sprayed region and the buoyancy-driven flow that results may be considered. 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 maximum decontamination factor (DF) for elemental iodine is based on themaximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine in the containment atmosphere remaining in equilibrium with the dissolved iodine in the containment water. This equilibrium is determined by the effective iodine partition coefficient. 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. 2.4Reduction in airborne radioactivity in the containment by in-containment recirculationfilter systems may be credited if these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. A-2 and A-3). 2.5Guidance for reduction in airborne radioactivity in the containment by suppressionpool scrubbing in BWRs is given in Section 6.5.5 of the SRP (Ref. A-1). Forsuppression pool solutions having pH less than 7, molecular iodine vapor should be conservatively assumed to evolve into the containment atmosphere.2.6Reduction in airborne radioactivity in the containment by retention in ice condensers,or other engineered safety features not addressed above, should be evaluated on anindividual case basis. See Section 6.5.4 of the SRP (Ref. A-1).2.7The primary containment (e.g., drywell and wetwell for Mark I and II containmentdesigns) should be assumed to leak at the peak pressure technical specification leakrate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 50% of the technical specification leak rate. For BWRs, leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by plant configuration and analyses, to a value not less than 50% of the technical specification leak rate. Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specifications. 2.8If the primary containment is routinely purged during power operations, releases viathe purge system prior to containment isolation should be analyzed and the resultingdoses 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. 3. ASSUMPTIONS ON DUAL CONTAINMENTS A-3For facilities with dual containment systems, the acceptable assumptions related tothe transport, reduction, and release of radioactive material in and from the secondary containment or enclosure buildings are as follows.3.1Leakage from the primary containment should be considered to be collected,processed by engineered safety feature (ESF) filters, if any, and released to theenvironment 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 2-1/2 times the height of any adjacent structure.3.2Leakage from the primary containment is assumed to be released directly to theenvironment as a ground-level release during any period in which the secondarycontainment does not have a negative pressure as defined in technical specifications. 3.3The effect of high wind speeds on the ability of the secondary containment to maintaina negative pressure should be evaluated on a case-by-case basis. The wind speed tobe 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 either 5% or 95% of the total number 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% of the time).3.4Credit for dilution in the secondary containment may be allowed when adequatemeans to cause mixing can be demonstrated. Otherwise, the leakage from theprimary 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. 3.5Primary containment leakage that bypasses the secondary containment should beevaluated at the bypass leak rate incorporated in the technical specifications. If thebypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of iodine and particulates may be considered on a case-by-case basis. Similarly, deposition of particulate radioactivity in gas-filled lines may be considered on a case-by-case basis.3.6Reduction in the amount of radioactive material released from the secondarycontainment because of ESF filter systems may be taken into account provided thatthese systems meet the guidance of Regulatory Guide 1.52 (Ref. A-2) and Generic Letter 99-02 (Ref. A-3).4. ASSUMPTIONS ON ESF SYSTEM LEAKAGEESF systems that recirculate sump water outside of the primary containment areassumed 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 A-4interfacing systems (Ref. A-4). The radiological consequences from the postulated leakageshould 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.4.1It is assumed that 50% of the core iodine inventory, based on the maximum reactorpower level, is mixed instantaneously and homogeneously in the primary containmentsump water (in PWRs) or the suppression pool (in BWRs) at the start of the accident.

In lieu of this 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.4.2The leakage should be taken as two times the sum of the simultaneous leakage fromall components in the ESF recirculation systems above which the technicalspecifications, or licensee commitments to item III.D.1.1 of NUREG-0737 (Ref. A-5),would require declaring such systems out of service. The factor of two multiplier is used to account for increased leakage in these systems over the duration of the accident and between surveillances or leakage checks. 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.4.3If the temperature of the leakage exceeds 212°F, the fraction of total iodine in theliquid that becomes airborne should be assumed equal to the fraction of the leakagethat 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:

fg f f h h h FF 2 1Where: h f1 is the enthalpy of liquid at system design temperature and pressure; h f2 isthe enthalpy of liquid at saturation conditions (14.7 psia, 212ºF); and h fg is the heat ofvaporization at 212ºF.4.4If the temperature of the leakage is less than 212°F or the calculated FF is less than10%, the amount of iodine that becomes airborne should be assumed to be 10% ofthe 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.4.5The radioiodine that is postulated to be available for release to the environment isassumed to be 97% elemental and 3% organic. Reduction in release activity bydilution or holdup within buildings, or by ESF ventilation filtration systems, may be credited where applicable. Filter systems used in these applications should be A-5evaluated against the guidance of Regulatory Guide 1.52 (Ref. A-2) and GenericLetter 99-02 (Ref. A-3).5. ASSUMPTIONS ON MAIN STEAM ISOLATION VALVE LEAKAGE IN BWRSFor BWRs, the main steam isolation valves (MSIVs) have design leakage that mayresult 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 determine the total calculated radiological consequences from the LOCA. The following assumptions are acceptable for evaluating the consequences of MSIV leakage.5.1For the purpose of this analysis, the activity available for release via MSIV leakageshould be assumed to be that activity determined to be in the drywell for evaluatingcontainment leakage (see Assumption 2 of this appendix). No credit should be assumed for activity reduction by the steam separators or by iodine partitioning in the reactor vessel.5.2All the MSIVs should be assumed to leak at the maximum leak rate above which thetechnical specifications would require declaring the MSIVs inoperable. The leakageshould be assumed to continue for the duration of the accident. Postulated leakage may be reduced after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if supported by site-specific analyses, to a value not less than 50% of the maximum leak rate. 5.3Reduction of the amount of released radioactivity by deposition and plateout on steamsystem piping upstream of the outboard MSIVs may be credited, but the amount ofreduction 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. 5.4In the absence of collection and treatment of releases by ESFs such as the MSIVleakage control system, or as described in Section 5.5 below, the MSIV leakageshould 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.5.5A reduction in MSIV releases that is due to holdup and deposition in main steampiping downstream of the MSIVs and in the main condenser, including the treatment ofair 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-6 and A-7 provide guidance on acceptable models.6. ASSUMPTION ON CONTAINMENT PURGINGThe radiological consequences from post-LOCA primary containment purging as acombustible 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 A-6other fission product release paths to determine the total calculated radiologicalconsequences 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-2) and Generic Letter 99-02 (Ref. A-3).

A-7Appendix A REFERENCESA-1USNRC, "Standard Review Plan for the Review of Safety Analysis Reports for NuclearPower Plants," NUREG-0800, July 1981 (or updates of specific sections).A-2USNRC, "Design, Inspection, and Testing Criteria for Air Filtration and AdsorptionUnits of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," Regulatory Guide 1.52, Revision 3, June

2001.A-3USNRC, "Laboratory Testing of Nuclear Grade Activated Charcoal," Generic Letter99-02, June 3, 1999.A-4USNRC, "Potential Radioactive Leakage to Tank Vented to Atmosphere," InformationNotice 91-56, September 19, 1991.A-5USNRC, "Clarification of TMI Action Plan Requirements," NUREG-0737, November 1980.A-6J.E. Cline, "MSIV Leakage Iodine Transport Analysis," Report, March 26, 1991. (ADAMS Accession Number ML003683718)A-7 USNRC, "Safety Evaluation of GE Topical Report, NEDC-31858P (Proprietary GEreport), Revision 2, BWROG Report for Increasing MSIV Leakage Limits and Eliminationof Leakage Control Systems , September 1993," letter dated March 3, 1999. (ADAMSAccession Number ML003683734, NUDOCS 9903110303)

B-1Appendix BASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENTThis appendix provides assumptions acceptable to the staff for evaluating theradiological 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 TERMAcceptable assumptions regarding core inventory and the release of radionuclidesfrom the fuel are provided in Regulatory Position 3 of this guide. The following assumptions also apply.1.1The number of fuel rods damaged during the accident should be based on aconservative analysis that considers the most limiting case. This analysis shouldconsider 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. 1.2The fission product release from the breached fuel is based on Regulatory Position3.2 of this guide and the estimate of the number of fuel rods breached. All the gapactivity in the damaged rods is assumed to be instantaneously released.

Radionuclides that should be considered include xenons, kryptons, and halogens.1.3The iodine gap inventory is composed of inorganic species (99.75%) and organicspecies (0.25%).1.4The radioactive material available for release is assumed to be from the assemblies with the peak inventory. The fission product inventory for the peak assemblyrepresents an upper limit value. The inventory should be calculated assuming the maximum achievable operational power history at the end of core life immediately preceding shutdown. This inventory calculation should include appropriate assembly peaking factors.2. WATER DEPTHIf the depth of water above the damaged fuel is 23 feet or greater, an effectivedecontamination factor (DF) of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water) for the elemental and organic species may be assumed. The difference in DFs for elemental (99.75%) and organic iodine (0.25%) species results in the iodine above the water being composed of 44% elemental and 56% organic species. If the depth of water is not at least 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref. B-1). Proposed increases in the pool DF above 200 will need to address re-evolution of the scrubbed iodine species over the accident duration and should be supported by empirical data.

1 These analyses should consider the time for the radioactivity concentration to reach levels correspondingto 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 andcontainment 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-2For release pressures greater than 1,200 psig, the iodine DFs will be less than thoseassumed in this guide and must be calculated on a case-by-case basis using assumptions comparable in conservatism to those of this guide.3. NOBLE GASESThe 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 DF).4. FUEL HANDLING ACCIDENTS WITHIN THE FUEL BUILDINGFor fuel handling accidents postulated to occur within the fuel building, the followingassumptions are acceptable to the NRC staff.4.1The radioactive material that escapes from the fuel pool to the fuel building isassumed to be released to the environment over a 2-hour time period. 4.2A reduction in the amount of radioactive material released from the fuel pool by engineered safety feature (ESF) filter systems may be taken into account providedthese 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 system 1 should be determined andaccounted for in the radioactivity release analyses. 4.3The radioactivity release from the fuel pool should be assumed to be drawn into theESF filtration system without mixing or dilution in the fuel building. If mixing can bedemonstrated, 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 CONTAINMENTFor fuel handling accidents postulated to occur within the containment, the followingassumptions are acceptable to the NRC staff.5.1If the containment is isolated 2 during fuel handling operations, no radiologicalconsequences need to be analyzed.5.2If the containment is open during fuel handling operations, but designed to automatically isolate in the event of a fuel handling accident, the release durationshould be based on delays in radiation detection and completion of containment 3 Technical specifications that allow such operations usually include administrative controls to close theairlock, hatch, or open penetrations within 30 minutes. Such adminstrative controls 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-3isolation. If it can be shown that containment isolation occurs before radioactivity isreleased to the environment, 1 no radiological consequences need to be analyzed.5.3If the containment is open during fuel handling operations (e.g., personnel air lock orequipment hatch is open), 3 the radioactive material that escapes from the reactorcavity pool to the containment is released to the environment over a 2-hour time period.5.4A reduction in the amount of radioactive material released from the containment byESF filter systems may be taken into account provided that these systems meet theguidance 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 5.5Credit for dilution or mixing of the activity released from the reactor cavity by natural orforced 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-4Appendix B REFERENCESB-1.G. Burley, "Evaluation of Fission Product Release and Transport," Staff TechnicalPaper, 1971. (NRC Accession number 8402080322 in NUDOCS in NRC's Public Document Room)B-2.USNRC, "Design, Inspection, and Testing Criteria for Air Filtration and AdsorptionUnits of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants," Regulatory Guide 1.52, Revision 3, June

2001.B-3.USNRC, "Laboratory Testing of Nuclear Grade Activated Charcoal," Generic Letter99-02, June 3, 1999.

1 The activity assumed in the analysis should be based on the activity associated with the projected fueldamage 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 mechanical vacuumpumps 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 offgas or standby gas treatment, will be considered on a case-by-case basis.C-1Appendix CASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR ROD DROP ACCIDENTThis appendix provides assumptions acceptable to the NRC staff for evaluating theradiological 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 inRegulatory Position 3 of this guide. For the rod drop accident, the release from the breachedfuel 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 minimal 1 fuel damage is postulated for the limiting event, the released activityshould be the maximum coolant activity (typically a pre-existing spike of 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 asfollows.3.1The activity released from the fuel from either the gap and/or from the fuel pellets isassumed to be instantaneously mixed in the reactor coolant within the pressurevessel.3.2Credit should not be assumed for partitioning in the pressure vessel or for removal bythe steam separators.3.3Of the activity released from the reactor coolant within the pressure vessel, 100% of the noble gases and 10% of the iodine are assumed to reach the turbine andcondensers. 3.4Of the activity that reaches the turbine and condensers, 100% of the noble gases and10% of the iodine are available for release to the environment. The turbine andcondensers leak to the environment as a ground-level release at a rate of 1% per day 2for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at which time the leakage is assumed to terminate. No credit should be assumed for dilution or holdup within the turbine building. Radioactive decay during holdup in the turbine and condenser may be assumed.

C-23.5In lieu of the transport assumptions provided in Sections 3.2 through 3.4 above, amore mechanistic analysis may be used on a case-by-case basis. Such analysesaccount 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 valve (MSIV) and considers MSIV closure time. 3.6The iodine species released from the reactor coolant within the pressure vesselshould be assumed to be 5% particulate, 91% elemental, and 4% organic. Therelease from the turbine and condenser should be assumed to be 97% elemental and 3% organic.

1 Minimal fuel damage is defined as an amount of damage that will yield reactor coolant system activityconcentration levels less than the maximum technical specification limits. 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 I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.D-1Appendix DASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A BWR MAINSTEAM LINE BREAK ACCIDENTThis appendix provides assumptions acceptable to the NRC staff for evaluating theradiological 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.1. SOURCE TERMAssumptions acceptable to the NRC staff regarding core inventory and the release ofradionuclides 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. 1.1If no or minimal 1 fuel damage is postulated for the limiting event, the released activityshould be the maximum coolant activity allowed by technical specification. The iodineconcentration in the primary coolant is assumed to correspond to the following two cases in the nuclear steam supply system vendor's standard technical specifications.1.1.1The 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, and1.1.2The concentration that is the maximum equilibrium value (typically 0.2 µCi/gmDE I-131) permitted for continued full power operation.1.2The activity released from the fuel should be assumed to mix instantaneously and homogeneously in the reactor coolant. Noble gases should be assumed to enter thesteam phase instantaneously.2. TRANSPORTAssumptions acceptable to the NRC staff related to the transport, reduction, andrelease of radioactive material to the environment are as follows.2.1The main steam line isolation valves (MSIV) should be assumed to close in themaximum time allowed by technical specifications.2.2The 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 throughthe valves prior to closure. 2.3All the radioactivity in the released coolant should be assumed to be released to theenvironment instantaneously as a ground-level release. No credit should be assumedfor plateout, holdup, or dilution within facility buildings.

D-22.4The iodine species released from the main steam line should be assumed to be 5%particulate, 91% elemental, and 4% organic.

1 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section inconjunction 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 Minimal fuel damage is defined as an amount of damage that will yield reactor coolant system activityconcentration levels less than the maximum technical specification limits. 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 I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.E-1Appendix EASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWRSTEAM GENERATOR TUBE RUPTURE ACCIDENTThis appendix provides assumptions acceptable to the NRC staff for evaluating theradiological 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.11. SOURCE TERMAssumptions acceptable to the NRC staff regarding core inventory and the release ofradionuclides from the fuel are in Regulatory Position 3 of this guide. 1.1 If no or minimal 2 fuel damage is postulated for the limiting event, the activity releasedshould be the maximum coolant activity allowed by technical specification. Two casesof iodine spiking should be assumed.1.1.1A reactor transient has occurred prior to the postulated steam generator tuberupture (SGTR) and has raised the primary coolant iodine concentration to themaximum value (typically 60 µCi/gm DE I-131) permitted by the technical specifications (i.e., a pre-accident iodine spike case). 1.1.2The primary system transient associated with the SGTR causes an iodine spikein the primary system. The increase in primary coolant iodine concentration isestimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in Curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated.

The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel pins assumed to have defects.1.2The activity released from the fuel, if any, should be assumed to be releasedinstantaneously and homogeneously through the primary coolant. The release fromthe breached fuel is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached.

E-21.3The specific activity in the steam generator liquid at the onset of the SGTR should beassumed to be at the maximum value permitted by secondary activity technicalspecifications (typically 0.1 µCi/gm DE I-131).1.4Iodine releases from the steam generators to the environment should be assumed tobe 97% elemental and 3% organic.2. TRANSPORTAssumptions acceptable to the NRC staff related to the transport, reduction, andrelease of radioactive material to the environment are as follows:2.1The primary-to-secondary leak rate in the steam generators should be assumed to bethe 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.2.2The 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 showcompliance 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 room temperature liquids. In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft 3).2.3The primary-to-secondary leakage should be assumed to continue until the primarysystem pressure is less than the secondary system pressure, or until the temperatureof the leakage is less than 100C (212F). The release of radioactivity from theunaffected steam generators should be assumed to continue until shutdown cooling isin operation and releases from the steam generators have been terminated. The release of radioactivity from the affected steam generator should be assumed to continue until shutdown cooling is operating and releases from the steam generator have been terminated, or the steam generator is isolated from the environment such that no release is possible, whichever occurs first. 2.4The release of fission products from the secondary system should be evaluated withthe assumption of a coincident loss of offsite power. The loss of offsite power shouldbe assumed to occur coincident with the start of the accident. 2.5The transport model described in this section should be utilized for iodine andparticulate releases from the steam generators. This model is shown in Figure E-1and summarized below.

3 Partition Coefficient is defined as:

PCmass of I per unit mass of liquidmass of I per unit mass of gas2 2E-3 Steam Space Bulk Water Primary Leakage Scrubbing Partitioning ReleaseFigure E-1Transport Model2.5.1A portion of the primary-to-secondary leakage will flash to vapor, based on thethermodynamic conditions in the reactor and secondary coolant.

With regard to the unaffected steam generators used for plant cooldown, theprimary-to-secondary leakage can be assumed to mix with the secondary waterwithout flashing during periods of total tube submergence.

During periods when the tubes in an affected or unaffected steam generatorare not covered with secondary water, a portion of the primary to secondaryleakage will flash to vapor. The amount of vapor released to the environment is based on the thermodynamic conditions in the reactor and secondary coolant.2.5.2The primary to secondary leakage that immediately flashes to vapor will risethrough 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-1), during periods of total submergence of the tubes.2.5.3The primary to secondary leakage that does not immediately flash is assumedto mix with the bulk water.2.5.4The 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.

3 A partitioncoefficient 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. 2.6Operating experience and analyses have shown that for some steam generatordesigns, tube uncovery may occur for a short period following any reactor trip (Ref.E-2). 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 E-4operating procedure restoration strategies on steam generator water levels should beevaluated.2.7All noble gas radionuclides released from the primary system are assumed to bereleased to the environment without reduction or mitigation.Appendix E REFERENCESE-1USNRC, "Iodine Behavior in a PWR Cooling System Following a Postulated SteamGenerator Tube Rupture Accident," NUREG-0409, May 1985.E-2USNRC, "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 inconjunction 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 Minimal fuel damage is defined as an amount of damage that will yield reactor coolant system activityconcentration levels less than the maximum technical specification limits. 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 I-131), only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.F-1Appendix FASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A PWR MAIN STEAM LINE BREAK ACCIDENTThis appendix provides assumptions acceptable to the NRC staff for evaluating theradiological 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.

11. SOURCE TERMS 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. 1.1If no or minimal 2 fuel damage is postulated for the limiting event, the activity releasedshould be the maximum coolant activity allowed by the technical specifications. Twocases of iodine spiking should be assumed.1.1.1A reactor transient has occurred prior to the postulated main steam line break(MSLB) and has raised the primary coolant iodine concentration to themaximum value (typically 60 µCi/gm DE I-131) permitted by the technical specifications (i.e., a pre-accident iodine spike case). 1.1.2The primary system transient associated with the MSLB causes an iodine spikein the primary system. The increase in primary coolant iodine concentration isestimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in Curies per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulated.

The assumed iodine spike duration should be 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel pins assumed to have defects.1.2The activity released from the fuel should be assumed to be released instantaneouslyand homogeneously through the primary coolant. The release from the breached fuelis 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.

F-21.3The specific activity in the steam generator liquid at the onset of the MSLB should beassumed to be at the maximum value permitted by secondary activity technicalspecifications (typically 0.1 µCi/gm DE I-131).1.4The chemical form of radioiodine released from the fuel should be assumed to be 5%particulate iodine, 91% element iodine, and 4% organic iodide. Iodine releases fromthe 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.2. TRANSPORTAssumptions acceptable to the NRC staff related to the transport, reduction, andrelease of radioactive material to the environment are as follows.2.1The bulk water in the faulted steam generator is assumed to rapidly blow down to theenvironment. The duration of the blowdown is obtained from thermal-hydraulicanalysis codes. The activity in the faulted steam generator bulk water is assumed released to the environment without mitigation.2.2For facilities that have not implemented alternative repair criteria (ARC), 1 the primary-to-secondary leak rate in the steam generators should be assumed to be the leak-ratelimiting condition for operation specified in the technical specifications. For facilities with traditional steam 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. For example, for a four-loop facility with a limiting condition for operation of 500 gpd for any one generator not to exceed 1 gpm from all generators, it would be appropriate to assign 500 gpd to the faulted generator and 313 gpd to each of the unaffected generators. For facilities that have implemented ARC, the primary-to-secondary leak rate in thefaulted steam generator should be assumed to be the maximum accident-induced leakage derived from the repair criteria and burst correlations. For the unaffected steam generators, the leak rate limiting condition for operation specified in the technical specifications is equally apportioned between the unaffected steam generators. 2.3The 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/ft 3).2.4The primary-to-secondary leakage should be assumed to continue until the primarysystem pressure is less than the secondary system pressure, or until the temperatureof 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.

3 Faulted refers to the state of the steam generator in which the secondary side has been depressurized bya MSLB such that protective system response (main steam line isolation, reactor trip, safety injection, etc.)

has occurred. F-32.5All noble gas radionuclides released from the primary system are assumed to bereleased to the environment without reduction or mitigation. A loss of offsite powershould be assumed to occur coincident with the start of the accident. 2.6The transport model described in assumptions 2.5 and 2.6 of Appendix E should beused for iodine. During dryout in the faulted 3 steam generator, all the primary-to-secondary leakage is assumed to flash to vapor and be released to the environment with no mitigation.

1 Facilities licensed with, or applying for, alternate repair criteria (ARC) should use this section inconjunction 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-1Appendix GASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCESOF A PWR LOCKED ROTOR ACCIDENTThis appendix provides assumptions acceptable to the NRC staff for evaluating theradiological consequences of a locked rotor accident at PWR light-water reactors.

1 Theseassumptions supplement the guidance provided in the main body of this guide. 1. SOURCE TERMAssumptions acceptable to the NRC staff regarding core inventory and the release ofradionuclides 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. 1.1If no fuel damage is postulated for the limiting event, a radiological analysis is notrequired as the consequences of this event are bounded by the consequencesprojected for the main steam line break outside containment.1.2The activity released from the fuel should be assumed to be released instantaneouslyand homogeneously through the primary coolant. 1.3The chemical form of radioiodine released from the fuel should be assumed to be 5%

particulate iodine, 91% element iodine, and 4% organic iodide. Iodine releases fromthe 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.2. RELEASE TRANSPORTAssumptions acceptable to the NRC staff related to the transport, reduction, andrelease of radioactive material to the environment are as follows.2.1The primary-to-secondary leak rate in the steam generators should be assumed to bethe 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.2.2The 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 showcompliance with leak-rate technical specifications. These tests are typically based on room temperature 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/ft 3).

G-22.3The primary-to-secondary leakage should be assumed to continue until the primarysystem pressure is less than the secondary system pressure, or until the temperatureof the leakage is less than 100C (212F). The release of radioactivity should beassumed to continue until shutdown cooling is in operation and releases from thesteam generators have been terminated.2.4The release of fission products from the secondary system should be evaluated withthe assumption of a coincident loss of offsite power.2.5All noble gas radionuclides released from the primary system are assumed to be released to the environment without reduction or mitigation.2.6The transport model described in assumptions 2.5 and 2.6 of Appendix E should be used for iodine.

1 Facilities licensed with, or applying for, alternate repair criteria (ARC) should use this section inconjunction 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-1Appendix HASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCESOF A PWR ROD EJECTION ACCIDENTThis appendix provides assumptions acceptable to the NRC staff for evaluating theradiological consequences of a rod ejection accident at PWR light-water reactors.

1 Theseassumptions supplement the guidance provided in the main body of this guide. Two release paths are considered: (1) release via containment leakage and (2) release via the secondary plant. Each release path is evaluated independently as if it were the only pathway available.

The consequences of this event are acceptable if the dose from each path considered separately is less than the acceptance criterion in Table 4. 1. SOURCE TERMAssumptions acceptable to the NRC staff regarding core inventory are in RegulatoryPosition 3 of this guide. For a 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. 1.1If no fuel damage is postulated for the limiting event, a radiological analysis is notrequired as the consequences of this event are bounded by the consequencesprojected for the loss-of-coolant accident (LOCA) and the main steam line break.1.2In the first release case, 100% of the activity released from the fuel should beassumed to be released instantaneously and homogeneously through thecontainment 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.1.3The chemical form of radioiodine released to the containment atmosphere should beassumed to be 5% particulate iodine, 91% elemental iodine, and 4% organic iodide. Evaluations of pH should consider the effect of acids created during the rod ejection accident event, e.g., pyrolysis and radiolysis products. 1.4Iodine releases from the steam generators to the environment should be assumed tobe 97% elemental and 3% organic.2. TRANSPORT FROM CONTAINMENTAssumptions acceptable to the NRC staff related to the transport, reduction, andrelease of radioactive material in and from the containment are as follows.

H-22.1A reduction in the amount of radioactive material available for leakage from thecontainment that is due to natural deposition, containment sprays, recirculating filtersystems, 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.2.2The containment should be assumed to leak at the leak rate incorporated in thetechnical specifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% ofthis 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.3. TRANSPORT FROM SECONDARY SYSTEMAssumptions acceptable to the NRC staff related to the transport, reduction, andrelease of radioactive material in and from the secondary system are as follows.3.1A leak rate equivalent to the primary-to-secondary leak rate limiting condition foroperation specified in the technical specifications should be assumed to exist untilshutdown cooling is in operation and releases from the steam generators have been terminated.3.2The 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 showcompliance 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/ft 3).3.3All noble gas radionuclides released to the secondary system are assumed to bereleased to the environment without reduction or mitigation.3.4The transport model described in assumptions 2.5 and 2.6 of Appendix E should be used for iodine.

I-1APPENDIX IACRONYMSASTAlternative source term ARCAlternative repair criteria BWRBoiling water reactor CDFCore damage frequency CEDECommitted effective dose equivalent COLRCore operating limits report DBADesign basis accident DCFDose conversion factor DEDose equivalent DFDecontamination factor DNBRDeparture from nucleate boiling ratio EABExclusion area boundary ECCSEmergency core cooling system EPAEnvironmental Protection Agency ESFEngineered safety feature FFFlash fraction FHAFuel handling accident FSARFinal safety analysis report GDCGeneral Design Criteria (in Appendix A to 10 CFR Part 50) gpmGallon per minute gpdGallon per day IPFIodine protection factor LBLOCALarge break loss-of-coolant accident LERFLarge early release fraction LOCALoss-of-coolant accident LPZLow population zone LWRLight-water reactor MOXMixed oxide MSIVMain steam isolation valve MSLBMain steam line break NDTNondestructive testing PRAProbabilistic risk assessment PWRPressurized water reactor RCSReactor cooling system RMRadiation monitor SERSafety evaluation report SGTRSteam generator tube rupture SRPStandard review plan TEDETotal effective dose equivalent TIDTechnical information document TMIThree Mile Island RA-1REGULATORY ANALYSISI.STATEMENT OF PROBLEMThe NRC staff is proposing to develop and issue a new regulatory guide, "Methods andAssumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-water Nuclear Power Reactors." The NRC is proposing this new guide as a means to provide guidance to licensees for license amendment requests that, in whole or part, seek to modify the licensing basis methodology and assumptions for performing evaluations of fission product releases and radiological consequences of several postulated light-water-reactor design basis accidents. The staff proposes to issue a draft guide for public review and comment, and upon resolution of public comments, to finalize and implement the guide.In the early 1970s, the staff issued guides for evaluating radiological consequences usingthe source term described in TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites" (Ref. RA-1). These accidents include loss-of-coolant, fuel handling, main steamline break, and rod ejection accidents. Since the 1980s, the staff also issued several standard review plans (SRPs) for evaluating other accidents such as the boiling water reactor (BWR) rod drop, pressurized water reactor (PWR) main steamline break, PWR steam generator tube rupture, and PWR locked rotor accidents (Ref. RA-2). Since no guidance existed for these accidents, the industry used the staff's SRP guidance to determine what acceptance criteria and methodologies were acceptable to the staff. The proposed guide would provide the first comprehensive guidance that includes all these accidents and guidance for performing radiological consequences analyses using the TID-14844 source term. The staff is currently addressing deficiencies in the control room habitability systems atcurrently licensed plants. This task, which has a long history, has received more attention of late because of recent industry experience in performing tracer gas measurements of unfiltered inleakage. The Nuclear Energy Institute has prepared an industry guideline, NEI-99-03, "Control Room Habitability Assessment Guidance" (Ref. RA-3). In meetings related to NEI-99-03, industry representatives expressed a strong desire to update regulatory guidance used to perform radiological dose assessments and requested that the staff provide this guidance. The staff is developing a generic letter and regulatory guides that will provide guidance on demonstrating compliance with General Design Criteria (GDC-19). II.EXISTING REGULATORY FRAMEWORKAccording to 10 CFR 50.34, "Contents of Applications; Technical Information," eachapplicant for a construction permit or operating license must 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 the operation of the facility. Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 establishes minimum requirements for the design criteria for water-cooled nuclear power plants. GDC-19, "Control Room," establishes minimum requirements for most facilities' control rooms, including:

RA-2Adequate radiation protection shall be provided to permit access and occupancyof the control room under accident conditions without personnel receiving 2 For licensees who have implemented an alternative source term pursuant to 10 CFR 50.67, GDC-19 and 10 CFR50.67(b)(2)(iii) restate the numeric criterion as 5 rem TEDE.

3 For licensees who have implemented an alternative source term pursuant to 10 CFR 50.67, 10 CFR 50.67(b)(2)(i)and 50.67(b)(2)(ii) restate the numeric criterion as 25 rem TEDE for the exclusion area and low population zone. RA-3radiation exposures in excess of 5 rem whole body, or its equivalent to any partof the body, for the duration of the accident.

2 Criteria for evaluating the radiological aspects of the proposed site are in 10 CFR 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

This evaluation is based on limiting the total radiation dose to the whole body and thyroid to less than 25 and 300 Rem, respectively, at the exclusion area and low population zone boundaries.

3After plant licensing, licensees may seek license amendments or plant modifications pursuant to either 10 CFR 50.59, "Changes, Tests, and Experiments," or 10 CFR 50.92, "Issuance of Amendment." The proposed guide would provide significant guidance for performing assessments of control room habitability and offsite doses when such assessments are used to support the license amendment or plant modification. Several regulatory guides for the calculation of offsite and control room radiologicalconsequences use the source term described in TID-14844. These regulatory guides are: Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Loss of Coolant Accident for Boiling Water Reactors" (Ref. RA-4)Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Loss of Coolant Accident for Pressurized Water Reactors" (Ref. RA-5)Regulatory Guide 1.5, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Steam Line Break Accident for Boiling Water Reactors" (Ref. RA-6)Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential RadiologicalConsequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors" (Ref. RA-7)Regulatory Guide 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accidentfor Pressurized Water Reactors" (Ref. RA-8).The NRC developed these guides in the early 1970s and has not updated them since then.The proposed regulatory guide that is the subject of this regulatory analysis would update these regulatory guides.III.OBJECTIVE OF THE REGULATORY ACTIONThe objective of the proposed regulatory guide is to provide guidance to licensees of operating nuclear reactors on acceptable methods and assumptions for performing evaluations of fission product releases and radiological consequences of several postulated light-water reactor design basis accidents. The guide would describe the source and the scope, nature, and documentation of associated analyses and evaluations. It would RA-4consider impacts on analyzed risk and describe the content of submittals acceptable to theNRC staff. The staff has determined that holders of operating licenses may continue to use methodsand assumptions previously approved by the NRC. The staff expects that licensees could use the information in the guide if they voluntarily decide to replace these methods and assumptions with those specified in this guide.IV.ALTERNATIVE APPROACHES1.Alternative 1 - Do Not Provide GuidanceUnder this alternative, the staff would not issue the proposed regulatory guidance. This isthe no action alternative. Not providing the needed guidance would result in an increased unnecessary burden for the licensee and the staff. This burden would be in the preparation and response to requests for additional information (RAIs), re-analyses, and supplementation of license amendment applications. This option is not supportive of any of the four nuclear reactor safety performance goals.2.Alternative 2 - Endorse an Industry Initiative that Addresses Evaluation ofRadiological Consequences of Design Basis AccidentsUnder this alternative, the staff would not develop its own regulatory guidance, but insteadwould endorse an acceptable industry document. The Nuclear Energy Institute (NEI) has prepared an industry guideline, NEI 99-03, "Control Room Habitability Assessment Guidance" (Ref. RA-3). This document was first submitted in August 1999 and was found to not adequately address the staff's concerns. NEI restructured and re-submitted NEI 99-03 in October 2000. An appendix to the NEI document addressed evaluating radiological consequences of design basis accidents. This appendix was based, in part, on a preliminary staff talking paper on this subject. There were still areas of disagreement between the staff and industry regarding the overall document. The staff and NEI agreed that the staff should prepare formal guidance and resolve these issues in the public comment process. The proposed regulatory guidance on evaluating radiological consequences of design basis accidents is part of this formal guidance. Therefore, this alternative is no longer viable.3.Alternative 3 - Endorse a National Consensus StandardThe staff was not able to identify any national consensus standard that addresses evaluating radiological consequences of design basis accidents, or other comparable methodology. Therefore, this alternative is not viable. 4. Alternative 4 - Revise Current Regulatory Guides to Address the Proposed Changes Revision of the current regulatory guides to address the proposed changes would notaddress all the proposed changes. The proposed changes not only affect current guides, but also affect accidents that regulatory guides do not cover. These analyses include the BWR rod drop accident, the PWR main steamline break, the PWR steam generator tube rupture, and the PWR locked rotor accidents. Currently, the only guidance for these RA-5accidents is in the staff's Standard Review Plan (Ref. RA-2). This information is in need ofupdating and made into a regulatory guide format for use by the industry. The NRC could revise the current regulatory guides, but not in an efficient and effectivemanner. Regulatory guides affected by the proposed changes were issued in the early 1970s. They contain information that is common among many of the guides. Because of their age and commonality, combining these guides into one new guide is more efficient than updating the guides individually. This is the same method that the NRC used when the regulatory guidance for determining the consequences of an accident using the alternative source term was created. The staff updated and incorporated changes into one regulatory guide rather than updating each of the affected regulatory guides and creating separate new guides where there currently was no regulatory guidance. The staff has concluded that it would not be an efficient use of resources to revise the current regulatory guides.5.Alternative 5 - Issue New Regulatory GuideIn this alternative, the staff would prepare a new regulatory guide that addresses the methods and assumptions for evaluating radiological consequences of light-water reactor design basis accidents. This alternative is supportive of all four performance goals, namely, issuing a new regulatory guide would (1) maintain public safety by ensuring that safety analyses use appropriate analysis assumptions and methods, (2 ) reduce unnecessary regulatory burden, (3) improve efficiency and effectiveness as the guidance would provide licensees with the staff position, thereby minimizing RAIs and re-submittals, and (4) maintain public confidence by providing guidance that ensures that safety analyses are adequate.The staff has determined that this alternative-issuing a new regulatory guide-is the mostadvantageous approach to addressing the need for additional regulatory guidance on performing assessments of control room atmospheric dispersion.IV.EVALUATION OF VALUES AND IMPACTSNew regulatory guidance would be voluntary for currently licensed operating reactors. A licensee may propose alternative approaches to demonstrate compliance with the NRC's regulations. For operating reactors, it is assumed the licensees would revise their current methods and assumptions for evaluating radiological consequences only if they perceive it to be in their interest to do so. A qualitative analysis follows:

!Completion of the proposed action is estimated to require from 0.2 to 0.5 FTE.

Associated costs include publication costs. The draft and final guides would be prepared internally.

!Regulatory efficiency would be improved by reducing uncertainty as to what is acceptable and by encouraging consistency in the assessment of control room habitability and offsite consequences. The availability of this guidance should benefit licensees by reducing the likelihood for follow-up questions and possible revisions in licensees' analyses and plant modifications. The proposed regulatory guide would simplify NRC reviews because license amendments should be more predictable and consistent analytically.

!A new regulatory guide would result in cost savings to both the NRC and industry.

The NRC will incur one-time incremental costs to develop the draft regulatory guide RA-6for comment and to finalize the regulatory guide. However, the NRC should alsorealize cost savings associated with more efficient review of licensee submittals. The staff believes that the continuous and on-going cost savings associated with these reviews should offset the one-time development costs.

!It is also expected that the industry would realize a net savings, as their one-time incremental cost to review and comment on a new regulatory guide would be compensated for by the efficiencies to be gained in minimizing follow-up questions and revisions associated with each licensee submittal.

!Assumptions and inputs provided by the proposed guide would be used in the assessment of the habitability of the control room and for the calculation of offsite doses during and after certain postulated accidents. Habitability requirements are established in the interest of providing an environment in which control room personnel can take actions to mitigate the consequences of these accidents, thereby assuring the health and safety of the public. Although the acceptance criteria for control room radiological habitability are comparable to the routine occupational exposure limits, the primary concern is protection of the public, rather than limiting occupational radiation exposure. Use of the proposed guidance may reduce the magnitude of the projected dose and increase the apparent margin to the acceptance criteria. Licensees may propose modifications to the control room habitability envelope and other systems that use a portion of this increased margin. The proposed guidance is expected to ensure that inputs and assumptions are calculated appropriately and that sufficient margins will continue to be present.

!The inputs and assumptions are used to assess the ability of the control room habitability envelope and systems to maintain an acceptable environment after an accident has occurred. Thus, the proposed changes cannot, of themselves, affect the actual accident sequence or progression or the core damage frequency (CDF) and large early release frequency (LERF). While changes to the envelope or systems may be enabled by the reduced calculated doses and the radiation exposure of control room personnel following an accident, the potential impact on CDF and LERF is likely to be negligible. The staff bases this conclusion on (1) radiation doses that could impact the ability of the operator to take necessary actions are substantially greater than the habitability acceptance criteria, (2) the existence of elevated doses of this magnitude are associated with core damage and containment releases that are on-going, and (3) if an event has progressed to core damage, the ability of the operator to take effective actions is reduced.

!With the possible exception of applicant agencies, such as TVA or municipal licensees, no other governmental agencies would be affected by the proposed regulatory guide. Pursuant to the categorical exclusion in 10 CFR 51.22 (c)(16), the issuance of the proposedregulatory guide does not require an environmental review. Under the provisions of the National Technology Transfer Act of 1995, Pub. L. 104-113, no voluntary consensus standard has been identified that could be used instead of the proposed regulatory guide (government-unique standard).V.CONCLUSION RA-7Experience with licensee amendment reviews has demonstrated the need for up-to-date and new guidance for performing radiological dose calculations. Based on this regulatory analysis, it is recommended that the NRC prepare a new regulatory guide on calculating the radiological consequences of design basis accidents, issue the guidance as a draft regulatory guide for public comment, and upon resolution of public comments, finalize the regulatory guide.

RA-8REFERENCES FOR REGULATORY ANALYSISRA-1J.J. DiNunno et al., "Calculation of Distance Factors for Power and Test ReactorSites," USAEC TID-14844, U.S. Atomic Energy Commission (now USNRC), 1962.RA-2USNRC, "Standard Review Plan for the Review of Safety Analysis Reports forNuclear Power Plants," NUREG-0800, July 1981 (or updates of specific sections).RA-3Nuclear Energy Institute, "Control Room Habitability Assessment Guidance," NEI 99-03, August 1999, October 2000. (Available from the NRC Public Document Room.)RA-4USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Loss of Coolant Accident for Boiling Water Reactors," Regulatory Guide 1.3, Revision 2, June 1974.RA-5USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Loss of Coolant Accident for Pressurized Water Reactors," Regulatory Guide 1.4, Revision 2, June 1974.RA-6USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Steam Line Break Accident for Boiling Water Reactors," Regulatory Guide 1.5, March 1971.RA-7USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequencesof a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," Regulatory Guide 1.25, March 1972.RA-8USNRC, "Assumptions Used for Evaluating a Control Rod Ejection Accident forPressurized Water Reactors," Regulatory Guide 1.77, May 1974.

RA-9BACKFIT ANALYSISThe proposed regulatory guide does not require a backfit analysis as described in 10CFR 50.109(c) because it does not impose a new or amended provision in the NRC's regulations. It does not impose a regulatory staff position interpreting the NRC's regulations different from a previous applicable staff position. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Although the guidance in the proposed regulatory guide is a significant departure from earlier staff guidance, this guide does not require the modification or addition to systems, structures, components, or design of a facility, or the procedures or organization required to design, construct, or operate a facility. Methods and solutions different from those set out in the regulatory guide will be acceptable if they provide a basis for the regulatory findings needed to support issuance or continuance of a permit or license by the Commission. A licensee can select a preferred method of achieving compliance with a license condition, the rules, or orders of the Commission as described in 10 CFR 50.109(a)(7). This regulatory guide provides an opportunity to use an updated method fordetermining control room and offsite radiological assessments, if that is the method the licensee prefers. The guide will be used by the NRC staff to evaluate licensee-initiated changes if there is a clear nexus between the proposed change and the guidance contained in the guide. It will also be used to review changes when the licensees have committed to using this guide.