Regulatory Guide 1.195

Methods and Assumptions for Evaluating Radiological Consequences of Design-Basis Accidents at Light-Water Nuclear Power Reactors

ML031490640
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Issue date:	05/31/2003
From:	Office of Nuclear Regulatory Research
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DG-1113 RG-1.195
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Regulatory guides are issued to describe and make available to the public such information as methods acceptable to the NRC staff for implementing specificparts of the NRC's regulations, techniques used by the staff in evaluating specific problems or postulated accidents, and data needed by the NRC staff in itsreview of applications for permits and license Regulatory guides are not substitutes for regulations, and compliance with them is not require Methods andsolutions different from those set out in the guides will be acceptable if they provide a basis for the findings requisite to the issuance or continuance of a permitor license by the Commission.This guide was issued after consideration of comments received from the publi Comments and suggestions for improvements in these guides are encouragedat all times, and guides will be revised, as appropriate, to accommodate comments and to reflect new information or experienc Written comments may besubmitted to the Rules and Directives Branch, ADM, U.S. Nuclear Regulatory Commission, Washington, DC 20555-000 Regulatory guides are issued in ten broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, Environmentaland Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review; and 10, General. Single copies of regulatory guides (which may be reproduced) may be obtained free of charge by writing the Distribution Services Section, U.S. NuclearRegulatory Commission, Washington, DC 20555-0001, or by fax to (301)415-2289, or by email to DISTRIBUTION@NRC.GO Electronic copies of this guideand other recently issued guides are available at NRC's home page at <WWW.NRC.GOV> through the Electronic Reading Room, Accession NumberML031490640.U.S. NUCLEAR REGULATORY COMMISSION May 2003 REGULATORY GUIDEOFFICE OF NUCLEAR REGULATORY RESEARCHREGULATORY GUIDE 1.195(Draft guide was issued as DG-1113)METHODS AND ASSUMPTIONS FOREVALUATING RADIOLOGICAL CONSEQUENCES OFDESIGN BASIS ACCIDENTS ATLIGHT-WATER NUCLEAR POWER REACTORS NRC DOCUMENTSRequests for single copies of draft or active regulatory guides (which may be reproduced)or for placement on an automatic distribution list for single copies of future draft guides in specific divisions should be made in writing to the U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301)415-2289; email <DISTRIBUTION@NRC.GOV>. 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>.Copies are available at current rates from the U.S. Government Printing Office, P.O. Box37082, 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;

<http://www.ntis.gov/ordernow>; telephone (703)487-465 Copies are available for inspectionor copying for a fee from the NRC Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR's mailing address is USNRC PDR, Washington, DC 20555; telephone (301)415-

4737 or (800)397-4209; fax (301)415-3548; email is PDR@NRC.GOV.Comments and suggestions for improvements in regulatory guides are encouraged at alltimes, and guides will be revised, as appropriate, to accommodate comments and to reflect new information or experienc Comments may be mailed to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. You may e-mail comments to: NRCREP@nrc.go You may also submit comments viathe NRC's rulemaking web site at http://ruleforum.llnl.go Address questions about ourrulemaking website to Carol Gallagher (301) 415-5905; email CAG@nrc.go You may faxcomments to the Rules and Directives Branch at (301) 415-5144.Publicly available documents related to this regulatory guide may be examined andcopied for a fee at the NRC's Public Document Room (PDR), Public File Area O1F21, 11555 Rockville Pike, Rockville, Marylan The regulatory guide and related documents, including comments, can be viewed and downloaded electronically via the NRC's rulemaking web site at http://ruleforum.llnl.gov.Publicly available documents created or received at the NRC are available electronicallyat the NRC's Electronic Reading Room at http://www.nrc.gov/NRC/reading-rm/adams.html. From this site, the public can gain entry into the NRC's Agencywide Document Access and Management System (ADAMS), which provides text and image files of NRC's public document If you do not have access to ADAMS or if there are problems in accessing the documents located in ADAMS, contact the NRC Public Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or by email to PDR@nrc.go TABLE OF CONTENTS INTRODUCTION..........................................................1 DISCUSSION.............................................................2 REGULATORY POSITION..................................................3 1.GENERIC CONSIDERATIONS ...........................................31.1Safety Margins....................................................3 1.2Defense in Depth..................................................4 1.3Scope of Required Analyses.........................................4 1.4Risk Implications..................................................6 1.5Submittal Requirements.............................................61.6Final Safety Analysis Report Requirements.............................72.DOSE ANALYSIS MODELS..............................................7 Radiological Consequences..........................................7 Activities in Compartments Without Inflow from the Environment...........8 Activities in the Environment........................................8 Activities in Compartments That Intake Only Outside Contaminated Air......9 Integrated Activity Released into the Environment ......................10 Integrated Activity in a Compartment.................................10 Offsite Doses....................................................10 Compartment Doses...............................................113.ACCIDENT SOURCE TERM............................................123.1Fission Product Inventory..........................................123.2Release Fractions.................................................12 3.3Timing of Release Phases..........................................143.4Radionuclide Composition.........................................143.5Chemical Form...................................................14 3.6Fuel Damage in Non-LOCA DBAs...................................144.DOSE CALCULATIONAL

METHODOLOGY

...............................154.1Offsite Dose Consequences.........................................15 4.2Control Room Dose Consequences...................................17 4.3Other Dose Consequences..........................................19 4.4Offsite Acceptance Criteria.........................................194.5Control Room Acceptance Criteria...................................194.6Other Dose Consequences Acceptance Criteria..........................205.ANALYSIS ASSUMPTIONS AND

METHODOLOGY

........................205.1General Considerations............................................20 5.2Accident-Specific Assumptions......................................21 5.3Meteorology Assumptions..........................................22 IMPLEMENTATION......................................................22 REFERENCES..............................................................23 APPENDICESA.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 1 As defined in 10 CFR 50.2, design bases means information that identifies the specific functions to be performed by astructure, system, or component of a facility and the specific values or ranges of values chosen for controlling parameters as reference bounds for desig These values may be (1) restraints derived from generally accepted "state of the art" practices forachieving functional goals or (2) requirements derived from analysis (based on calculation or experiments or both) of the effectsof a postulated accident for which a structure, system, or component must meet its functional goal The NRC considers theaccident source term to be an integral part of the design basis because it sets forth specific values (or a range of values) forcontrolling parameters that constitute reference bounds for design.1.195-1

PURPOSE

AND SCOPEA.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 accident It describes the radiological 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," Section50.34, "Contents of Applications; Technical Information," requires that each applicant for aconstruction 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 facilit Applicants are also required by 10 CFR 50.34 to provide an analysis of the proposed sit In 10 CFR Part 100, "ReactorSite Criteria," Section 100.11, "Determination of Exclusion Area, Low Population Zone, andPopulation Center Distance," provides criteria for evaluating the radiological aspects of theproposed sit A footnote to 10 CFR 100.11 states that the fission product release assumed in these evaluations should be based on a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission product 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 andrequires means for remote plant shutdow 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 acciden TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites" (Ref. 1), iscited in 10 CFR Part 100 as a source of further guidance on these analyse 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 requiredby 10 CFR 50.34 for an operating license are documented in the facility's final safety analysis report(FSAR). Fundamental assumptions that are design inputs, including the source term, are to be included in the FSAR and become part of the facility design basi .195-2Since 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 accident 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 voluntarilyrevise the accident source term used in design basis radiological consequence analyse 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 inRegulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design BasisAccidents 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 50.6 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 regulatory guide are covered by therequirements of 10 CFR Part 50, which were approved by the Office of Management and Budget (OMB), approval number 3150-301 The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control numbe 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 occurrenc 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'sengineered safety feature These accident analyses are intentionally conservative in order to compensate for known uncertainties in accident progression, fission product transport, and atmospheric dispersio 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 equipment and human performanc 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-depthphilosoph .195-3The 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)

chapter 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 100.1 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 in Regulatory Guide 1.196, "Control Room Habitability at Light-Water Nuclear PowerReactors." The affected guides will not be withdrawn as they may still be used at the option oflicensees. Specifically, the guidance in Regulatory Guides 1.195 and 1.196 can 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 report This guide does not address all areas of potentially significant ris Although this guide addresses fuel handling accidents, other events that could occur during shutdown operations are not currently addresse The NRC staff has several ongoing initiatives involving risks of shutdown operations, extended burnup fuels, and risk-informing current regulation The information in this guide may be revised in the future as NRC staff evaluations are completed and regulatory decisions on these issues are mad REGULATORY POSITION1.GENERIC CONSIDERATIONS 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 uncertaintie Changes, or the net effects of multiple changes, that result in a 1.195-4reduction in safety margins may require prior NRC approva Licensees may use 10 CFR 50.59and its supporting guidance to assess safety margins related to facility modifications 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 dat 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 uncertaintie 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 appl 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 engineered safeguardsequipment 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 managementprogram. The radiological analyses provide a fundamental basis upon which a significant portion ofthe facility design is base Additionally, many aspects of facility operation derive from radiological design analyse 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 guid .3Scope of Required Analyses1. Design Basis Radiological AnalysesA fundamental commitment required for application of the methodology in this guide isto perform an analysis of each applicable acciden The scope of accidents considered should include accidents mentioned in this guide applicable to a specific design (i.e., boiling-water reactors (BWRs) or pressurized-water reactors (PWRs)), supplemented by those in the FSAR and other licensee documents, as appropriat Regulatory Positions 1.3.2 and 1.3.3 may be used to further define the scope and type of analyses performed. The performance of these analyses will determine the limiting event with respect to offsiteand control room dos Some licensees have evaluated the control room dose only for the DBA LOCA, which is typically the limiting event for offsite radiological release The DBA LOCA is generally the large break (LB) LOCA event analysi Other events may be analyzed as part of the design basis accident evaluation for the facilit Although these events may have been shown to 1.195-5be nonlimiting with respect to offsite dose, control room dose analyses for these events areneeded to identify the limiting event for the GDC-19 control room dose design criterio There are several regulatory requirements for which compliance is demonstrated, in part,by the evaluation of the radiological consequences of design basis accident A plant's licensingbases may include, but are not limited to, the following. Environmental Qualification of Equipment (10 CFR 50.49)Control Room Habitability (GDC-19 of Appendix A to 10 CFR Part 50)Emergency Response Facility Habitability (Paragraph IV.E.8 of Appendix E to 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 various licenseecommitments refer to specific evaluation A plant's licensing bases may include, but are notlimited 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. Re-Analysis GuidanceFacility modification should be supported by evaluations of all significant radiological andnonradiological impacts of the proposed action This evaluation should consider the impact of the proposed changes on the facility's compliance with the regulations and commitments listed aboveas well as any other facility-specific requirement These impacts may be due to (1) the associated facility modifications or (2) the differences in the methodology utilize The scope and extent of the re-evaluation will be a function of the specific proposed facility modification or the change in methodolog The NRC staff expects licensees to evaluate all impacts of the proposed changes and to update the affected analyses and the design bases appropriatel 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 vali 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 implementatio Sensitivity analyses, discussed below, may also be an optio If affected design basis analyses are to be recalculated, all affected assumptions and inputs should be update Any license amendment request should describe the licensee's re-analysis effort and providestatements 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 guid .195-61. 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 recalculate As used in this guide, a sensitivityanalysis 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 theresults of the analysis bound those obtainable from a more complete treatmen 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 input 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 analyse If sensitivity or scoping analyses are used, the license amendment request should include a discussion of the analyses performed and the conclusions draw Scoping or sensitivity analyses should not constitute a significant part of the evaluations for the design basis exclusion area boundary (EAB), low population zone (LPZ), or control room dose unless there is a clear and defensible basis for doing so.1.4Risk ImplicationsThis guide provides guidance only on the regulatory assumptions that licensees shouldmake in their calculation of the radiological consequences of design basis accident These assumptions have no direct influence on the probability of the design basis accident initiato 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 ris Consideration should be given to the risk impact of proposed implementations that seek toremove or downgrade the performance of previously required engineered safeguards equipment on the basis of the reduced postulated dose The NRC staff may request risk information if there is a reason to question adequate protection of public health and safet The licensee may elect to use risk insights in support of proposed changes to the designbasis that are not addressed in currently approved NRC staff position For guidance, refer to Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-InformedDecisions 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 application Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for NuclearPower Plants (LWR Edition)" (Ref. 11), provides additional guidanc The NRC staff's finding asto 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 and licensing basis of the facilit The amendment request should describe the licensee's analyses of the radiological and nonradiologicalimpacts of the proposed modification in sufficient detail to support review by the NRC staf The 1.195-7staff recommends that licensees submit affected FSAR pages annotated with changes that reflectthe revised analyses or submit the actual calculation documentatio If the licensee has used a current, approved version of an NRC-sponsored computer code,the NRC staff review can be made more efficient if the licensee identifies the code used and submits the inputs that the licensee used in the calculations made with that cod In many cases, this will reduce the need for NRC staff confirmatory analyse 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 beupdated 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 license amendment in accordance with 10 CFR 50.5 The affected radiological analysis descriptions in the FSAR should be updated to reflect the replacement of the design basis changes to the methodology and input The analysis descriptions should contain sufficient detail to identify the methodologies used, significant assumptions and inputs, and numeric result Regulatory Guide 1.70 (Ref. 11)

provides additional guidanc 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 change willmaintain an adequate level of protection of public health and safet 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 commitments, and staff experience gained in approving similar requests for other plant 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 thefacility's design basi Licensees should ensure that adequate information, including analysisassumptions, 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 thelicensee's conclusio Licensees should expect an NRC staff effort to resolve critical differencesin analysis assumptions, inputs, and methods used by the licensee and those deemed acceptable to the NRC staff. 2.DOSE ANALYSIS MODELS Radiological ConsequencesThe radiological consequences of an accident in a nuclear reactor depend on the quantity ofthe radioactive material that escapes to the environment or enters the control roo As the radioactivity is transported through the containment and other buildings, credit is given for several natural and engineered removal mechanism Within compartments, these removal mechanisms include sprays, natural deposition, leakage, natural and forced convection, filters, and suppression 1.195-8pool This Regulatory Position describes an acceptable set of general equations used to model thetransport and removal of fission products between compartments, the calculation of activities in the environment, and the calculation of offsite and compartment dose The equations contained within this Regulatory Position do not model the impact ofdaughter products (e.g., I-135 to Xe-135) that are due to the decay of the parent isotope Daughter products do not typically contribute significantly to the dos If it is determined that they contribute significantly to the whole body, thyroid, or beta doses, daughter products should be considered and the equations provided in Regulatory Position 2 will need to be modifie .2 Activities in Compartments Without Inflow From the EnvironmentThe following balance equation models the rate of change of activity of a nuclide in acompartment An example of a compartment that is typically without inflow from theenvironment is a reactor containment buildin (1)Mrk1kN1jrrjktkrraadtdawhere:ar=activity of a nuclide in compartment r at time t, Ciak=activity of a nuclide in compartment k at time t, CiN =the number of removal processes M =the number of compartments modeledrj =removal constant of the jth removal process internal to compartment r, i.e., decay,plateout, filtration, spray in containment, flow out of compartment r, sec-1tkr =transfer constant from region k to compartment r, i.e., flow rate from compartment k tocompartment r divided by the volume of compartment k, sec-1For halogens, a more specific form of Equation 1 may be written to account for removalmechanisms that are chemically species-specific, e.g., filter efficiencies for particulate, elemental, and organic iodin For these situations, Equation 1 can be rewritten to define the activity and removal constants on a per nuclide and species basi .3 Activities in the EnvironmentEquation 1 is solved for the time-dependent activity in each compartmen The release ratefrom M compartments to the environment is given by Equation The activity in the environmentfrom each compartment is given by Equation 3.(2)M1kkeRR 1.195-9(3)kkeUkekeFkkeaQf1QV1Rwhere:ak=activity of compartment k, Cifke=filter removal efficiency fraction for a filter between compartment k and theenvironment, dimensionless QkeF=filtered flow from compartment k to the environment, m³/secQkeU=unfiltered flow from compartment k to the environment, m³/secR=release rate of activity from M compartments to the environment, Ci/secRke=release rate of activity from compartment k to the environment, Ci/secVk=free volume of compartment k, m32.4 Activities in Compartments That Intake Only Outside Contaminated Air Equation 4 models compartments that intake radioactivity transported to the compartmentvia only atmospheric dispersio Control rooms or technical support centers that do not intake radioactivity directly from other buildings are examples of these compartments.(4)N1jrrjkeMrk1kkererUererFraRQ/Q)f1(Qdtdawhere: fer= filter removal fraction for a filter between the environment and compartment r,dimensionlessQerF=filtered intake flow rate from the environment to compartment r, m³/secQerU= 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 = QE/Vr, where QE isthe exhaust flow rate from the control room to the environment (QE = QerF + QerU), m3/sec, and Vr is the free volume of the control room, m3 r2=nuclide decay constant, sec-1 r3=recirculation removal rate, sec-1 = (QR/Vr) x fR where QR is the recirculation flowrate in the control room, m3/sec; Vr is the free volume of the control room, m3; andfR is the recirculation filter removal efficiency fraction, dimensionles .195-10 Integrated Activity Released Into the Environment The integrated activity (Curies) released into the environment over the time interval j fromtime t0, to t1, IARj, is given by the following equatio In calculating IARj, no credit is taken forcloud depletion by ground deposition or radioactive decay during transit to the exclusion area boundary or the LPZ outer boundary.(5)10ttjRdtIAR2.6 Integrated Activity in a CompartmentThe integrated activity (Ci-sec) in a compartment k over the time interval j from time t0 tot1, IAkj, is determined by the expression:(6)10ttkkjdtaIA2.7 Offsite Doses The following equations give the models used to calculate offsite dose Equations forcalculating thyroid and whole body doses are given. Offsite thyroid doses are calculated using the equation: (7)N1iT1jjjijiTHTH)Q()BR()IAR()DCF(DAssuming a semi-infinite cloud of photon emitters, offsite whole body doses are calculatedusing the equation: (8)N1iT1jjijiBB)Q()IAR()DCF(Dwhere: (BR)j= breathing rate during time interval j, m³/secDTH= offsite thyroid dose via inhalation during time interval j, rem DB=offsite whole body dose during time interval j, rem (DCFTH)i= thyroid dose conversion factor via inhalation for nuclide i, rem/Ci (DCFB)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 nuclidesT=number of time intervals over which (IAR) is calculated(/Q)j= offsite atmospheric dispersion factor during time interval j, sec/m³ 2 Control room envelopes may be composed of more than one room or subcompartmen If those rooms contain shielding thatblocks 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.1.195-112.8 Compartment Doses Compartment thyroid doses via inhalation pathway are calculated using the followingequation: N1iT1jjjijkiTHkkTH)BR(O)IA()DCF(V1)D((9)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 emitter The finite cloud photon doses are calculated using Murphy's method, which models the compartment as a hemispher The following equationis used: (10)N1iT1jjijkiBkkkBO)IA()DCF(VGF1)D(The beta skin doses in a compartment are calculated using the following equation:(11)N1iT1jjijkiSkkSO)IA()DCF(V1)D(where: BRj=breathing rate in time interval j, m3/sec (DTH)k=compartment k thyroid dose via inhalation, rem (DB)k=compartment k whole body dose, rem(DS)k=compartment k beta skin dose, rem (DCFS)i=beta skin dose conversion factor for nuclide i, rem-m³/Ci-secGFk=dose reduction due to the compartment geometry correction factor352/Vk0.338, or if Vk is defined in units of cubic feet the geometry factor is1173/Vk0.338, dimensionless2 (see Regulatory Position 4.2.7) (IAk) ij=integrated activity concentration in compartment k, for nuclide i during timeinterval j, Ci-secOj=compartment occupancy fraction during time interval jT=number of time intervals over which (IA) is calculatedVk=compartment k free volume, m3 In addition to the dose contribution from the compartment airborne activity describedabove, the whole body dose contribution from external sources as listed in Regulatory Position 4.2.1 should be considere The uncertainty factor used in determining the core inventory should be that value provided in Appendix K to 10 CFR Part 50,typically 1.0 A value lower than 1.02, but not less than 1.00 (correlates to the licensed power level) may be used provided theproposed 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 acciden For these analyses it is appropriate to usethe core average inventory to assess the consequences of this accident. 1.195-123.ACCIDENT SOURCE TERMThis Regulatory Position provides a source term that is acceptable to the NRC staf Itprovides 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 uncertaint These parameters should be examined to maximize fission productinventor The period of irradiation should be of sufficient duration to allow the activity of dose- significant radionuclides to reach equilibrium or to reach maximum value 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 and thecore average inventory should be use Further assumptions are in Appendix A to this guid ForDBA 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 cor 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 indetermining the inventory of the damaged rods.No adjustment to the fission product inventory should be made for events postulated tooccur during power operations at less than full rated power or those postulated to occur at the beginning of core lif 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, by radionuclide groups, for DBA LOCAs and non-LOCA DBAs where the fuel is melted and the cladding is breached are listed in Table 1 for BWRs and PWR These fractions are applied to the equilibrium core inventory described in Regulatory Position 3.1. For non-LOCA events, where only the cladding is breached, the fractions of the coreinventory assumed to be in the gap for the various radionuclides are given in Table The release 5 The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peakrod burnup up to 62,000 MWD/MT The data in this section may not be applicable to cores containing mixed oxide (MOX)

fue If wall deposition by containment sprays is modeled as a time-dependent process, such as in Revision 2 of Standard ReviewPlan Section 6.5.2 (Ref. 14), 50% of the equilibrium radioactive iodine inventory is assumed released into the containment atmosphere and is available to be deposited on the walls on the containmen Elemental plateout using the time-dependent models in Revision 2 of Standard Review Plan Section 6.5.2 may be assumed.If wall deposition by containment sprays is modeled as a time-independent process (instantaneous removal), such as inRevision 0 of Standard Review Plan Section 6.5.2, 50% of the equilibrium, radioactive iodine inventory is assumed released into the containment atmospher One-half of this iodine is assumed to be instantaneously deposited on the walls of the containment. The net value of core inventory available for release from containment would, therefore, be 25% of the equilibrium radioactive iodine inventor Further iodine removal by wall deposition should not be assume For these assumptions a limitation of 10 hr-1 should be imposed on the elemental iodine spray removal lambda.Please note that Revision 2 of SRP Section 6.5.2 erroneously implied that 25% of the equilibrium radioactive iodine inventorydeveloped from maximum full-power operation of the core should be assumed to be immediately available for the leakage from the primary reactor syste This value should be 50% of the equilibrium radioactive iodine inventory when time-dependent walldeposition by spray is assume Using 50% prevents accounting twice for the iodine deposited on the wall of the containmen The release fractions listed here have been determined to be acceptable for use with currently approved LWR fuel with a peakburnup up to 62,000 MWD/MTU provided that the maximum linear heat generation rate does not exceed 6.3 kw/ft peak rodaverage power for rods with burnups that exceed 54 GWD/MT As an alternative, fission gas release calculations performedusing NRC-approved methodologies may be considered on a case-by-case basi To be acceptable, these calculations must use aprojected power history that will bound the limiting projected plant-specific power history for the specific fuel loa For theBWR rod drop accident and the PWR rod ejection accident, the gap fractions are assumed to be 10% for iodines and noble gases.1.195-13fractions from Table 2 are used in conjunction with the calculated fission product inventory andthe maximum core radial peaking factor. Table 15BWR and PWR Core Inventory Fraction Released Into Containment AtmosphereGroupRelease FractionNoble Gases1.0Iodines60.5Table 27Non-LOCA Fraction of Fission Product Inventory in Gap GroupFractionI-1310.08 Kr-850.10 Other Noble Gases0.05 Other Iodines0.053.3Timing of Release PhasesFor LOCA DBAs, the core activity released is assumed to be immediately available forrelease from containmen 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 damage .195-143.4Radionuclide CompositionTable 3 lists the elements in each radionuclide group that should be considered in designbasis analyses.Table 3Radionuclide GroupsGroupElementsNoble GasesXe, KrIodinesI3.5Chemical FormOf the radioiodine released from the reactor coolant system (RCS) to the containment in apostulated accident, 5% of the iodine released should be assumed to be particulate iodine, 91%

elemental iodine, and 4% organic iodid This includes releases from the gap and the fuel pellet 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 LOCA However, the transport of these iodine species following release from the fuel may affect these assumed fraction 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 be analyzedto 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 breache 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 locked rotoraccidents, the amount of fuel damage should be evaluated assuming that the highest worth control rod is stuck at its fully withdrawn positio The amount of fuel damage caused by a FHA is addressed in Appendix B to this guide.4. DOSE CALCULATIONAL

METHODOLOGY

This Regulatory Position provides a dose calculational methodology that is acceptable tothe NRC staf It provides guidance on the calculation of offsite and onsite consequences and on offsite and control room acceptance criteri Licensees who use these dose conversion factors in accident calculations should determine whether this will impact thefacility's technical specification definition for dose equivalent I-13 .195-154.1 Offsite Dose ConsequencesThe following assumptions should be used in determining the doses for persons located ator 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 RadionuclideIntake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion" (Ref. 16), provides tables of conversion factors acceptable to the NRC staf Thefactors in the column headed "thyroid" should be use .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 to be3.5 x 10-4 cubic meters per secon From 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident, the breathing rateshould be assumed to be 1.8 x 10-4 cubic meters per secon After that and until the end of theaccident, 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-infinitecloud assumptions with appropriate credit for attenuation by body tissu Table III.1 of FederalGuidance Report 12, "External Exposure to Radionuclides in Air, Water, and Soil" (Ref. 17),provides external conversion factors acceptable to the NRC staf The factors in the column headed "effective" yield doses correspond to the whole body dos 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 at themost limiting EAB locatio 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 the start 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 with the dose criteria in Table 4.4.1.7No correction should be made for depletion of the effluent plume by deposition onthe groun For PWRs with steam generator alternative repair criteria, different dose criteria may apply to steam generator tube rupture andmain steam line break analyses as suggested by guidance being developed in Draft Regulatory Guide DG-1074, "SteamGenerator Integrity" (Ref. 18). 1.195-16Table 4EAB and LPZ Accident Dose CriteriaDose Criteria (rem)Accident or CaseWholeBodyThyroid Analysis Release DurationLOCA25 rem300 rem30 days for containment,ECCS, and MSIV (BWR)

leakageBWR Main Steam Line BreakInstantaneous puffFuel Damage or Pre-Accident Spike25 rem300 remEquilibrium Iodine Activity2.5 rem30 remBWR Rod Drop Accident6.3 rem75 rem24 hours PWR Steam Generator Tube9RuptureAffected SG: time to isolate;Unaffected SG(s): until cold shutdown is establishedFuel Damage or Pre-Accident Spike25 rem300 remCoincident Iodine Spike2.5 rem30 remPWR Main Steam Line Break9Until cold shutdown isestablishedFuel Damage or Pre-Accident Spike25 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 hours4.2 Control Room Dose ConsequencesThe following guidance should be used in determining the whole body, thyroid, and skindoses 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 personne The applicable sources will vary from facility to facility, but typically will include:

10 The iodine protection factor (IPF) methodology of Reference 19 may not be adequately conservative for all DBAs and controlroom arrangements since it models a steady-state control room conditio Since many analysis parameters change over the duration of the event, the IPF methodology should only be used with cautio The NRC computer codes HABIT (Ref. 20) and RADTRAD (Ref. 21) incorporate suitable methodologies.1.195-17Contamination of the control room envelope atmosphere by the intake orinfiltration of the radioactive material contained in the radioactive plume released from the facility,Contamination of the control room envelope atmosphere by the intake orinfiltration of airborne radioactive material from areas and structures adjacent to the control room envelope.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.4.2.3The models used to transport radioactive material into and through the control roomenvelope,10 and the shielding models used to determine radiation dose rates from external sources,should be structured to provide suitably conservative estimates of the exposure to control room personnel.4.2.4Credit for engineered safety features that mitigate airborne radioactive materialwithin the control room envelope may be assume Such features may include control room isolation or pressurization, or intake or recirculation filtratio Refer to Section 6.5.1, "ESFAtmospheric 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. 22);and Generic Letter 99-02 (Ref. 23) for guidanc The control room envelope design is often optimized for the DBA LOCA and the protection afforded for other accident sequences may not be as advantageou 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 accident 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 protective equipment oruse 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the event, 60% of the time between 1 and 4 days, and 40% of the time from 4 days 11 These occupancy factors are already included in the determination of the /Q values using the Murphy and Campemethodology (Ref. 19) and should not be credited twic The ARCON96 Code (Ref. 24) does not incorporate these occupancy factors in the determination of the /Q value Therefore, when using ARCON96 /Q values, occupancy factors should beincluded in the dose calculations.1.195-18to 30 days.11 For the duration of the event, the breathing rate of this individual should be assumedto be 3.5 x 10-4 cubic meters per second (Ref. 25).4.2.7Control room envelope doses should be calculated using dose conversion factorsidentified in Regulatory Position 4.1 above for use in offsite dose analyse The calculation should consider all radionuclides that are significant with regard to dose consequences and the release of radioactivit 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 facto This factor is incorporated in Equation 10 of Regulatory Position 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 dimension The skin dose DCFs presented in Federal Guidance Report 12 (Ref. 17) are based on both photon and beta emission Doses should be calculated using the factors in the column headed "Skin" in Table III.1 of Federal Guidance Report 12.4.3Other Dose ConsequencesThe guidance provided in Regulatory Positions 4.1 and 4.2 should be used, as applicable, inre-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 100.1 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 LOC 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). 4.5Control Room Acceptance CriteriaThe following guidelines may be used in lieu of those provided in SRP 6.4 (Ref. 14) whenshowing compliance with the dose guidelines in GDC-19 of Appendix A to 10 CFR Part 5 The following guidelines relax the thyroid and skin acceptance criteria from that given in SRP Currently, 10 CFR 20.1201 limits organ dose to 50 rem annuall The release duration is specified in Table The exposure period is 30 days for all accident The criterion in GDC-19 applies to all accident Credit for the beta radiation shielding afforded by special protective clothing and eye protedtion is allowed if the applicantcommits to their use during severe radiation release However, even though protective clothing is used, the calculated unprotected skin dose is not to exceed 75 re These limits are design criteria and are not to be interpreted as acceptableoccupational doses.1.195-19Whole body5 remThyroid50 rem Skin50 rem124.6 Other Dose Consequences Acceptance CriteriaThe acceptance criteria for the various NUREG-0737 (Ref. 2) items generally referenceGDC-19 from Appendix A to 10 CFR Part 50 or specify criteria derived from GDC-1 These criteria remain unchanged except for the thyroid and beta dose limits as stated in Regulatory Position 4.5. Before the General Design Criteria were established in 10 CFR Part 50, these criteriaexisted in draft for Some of the facilities that were licensed during this time period committed to various draft criterion for control room habitabilit These commitments may be different from GDC-1 Application of this regulatory guide to those plants will be considered on a case by case basi .ANALYSIS ASSUMPTIONS AND

METHODOLOGY

5.1General Considerations5. Analysis QualityThe analyses required by 10 CFR 100.11 and GDC-19 in Appendix A to 10 CFR Part 50and 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 50.5 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 ReprocessingPlants," to 10 CFR Part 5 These design basis analyses were structured to provide a conservative set of assumptions totest the performance of one or more aspects of the facility desig Many physical processes and phenomena are represented by conservative, bounding assumptions rather than modeled directl 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 dispersio Licensees should exercise caution in proposing deviations based on 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. Credit for Engineered Safeguard FeaturesCredit may be taken for accident mitigation features that are classified as safety related, arerequired 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 13 Note that for some parameters, the technical specification value may be adjusted for analysis purposes by factors provided inother regulatory guidanc For example, ESF filter efficiencies are based on the guidance in Regulatory Guide 1.52 (Ref. 22) andin Generic Letter 99-02 (Ref. 23) rather than the surveillance test criteria in the technical specification Generally, these adjustments address possible changes in the parameter between scheduled surveillance tests.1.195-20addressed in emergency operating procedure The single active component failure that results in the most limiting radiological consequences should be assume Assumptions regarding the occurrence and timing of a loss of offsite power should be selected with the objective of maximizing the postulated radiological consequence Design basis delays in actuation of these features should be considered, especially for features that rely on manual operator intervention.5. Assignment of Numeric Input ValuesThe numeric values that are chosen as inputs to the dose analyses required by regulationsand described in Regulatory Position 5.1.1 should be selected with the objective of determining a conservative postulated dos In some instances, a particular parameter may be conservative in one portion of an analysis but be nonconservative in another portion of the same analysi 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 p Sensitivity analyses may be needed to determine the appropriate value to us As a conservative alternative, the limiting value applicable to each portion of the analysis may be used in the evaluation of that portio A single value may not be applicable for a parameter for the duration of the event, particularly for parameters affected by changes in densit For parameters addressed by technical specifications, the value used in the analysis should be that specified in the technical specifications.13 If a range of values or a tolerance band is specified, the value that would result ina conservative postulated dose should be use 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 valu .2Accident-Specific AssumptionsThe appendices to this regulatory guide provide accident-specific assumptions that areacceptable to the staff for performing analyses that are described in Regulatory Position 5. Licensees should review their license basis documents for guidance pertaining to the analysis of radiological design basis accidents other than those provided in this guid Licensees should analyze the DBAs that are affected by the specific proposed changes to the facility or to the radiological analyse The NRC staff has determined that the analysis assumptions in the appendices to this guideprovide an integrated approach to performing the individual analyses and generally expects licensees to address each assumption or to propose acceptable alternative Such alternatives may be justifiable on the basis of plant-specific considerations or updated technical analyse 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 consistency among the assumptions in this guide.The NRC is committed to using probabilistic risk analysis (PRA) insights in its regulatoryactivities and will consider licensee proposals for changes in analysis assumptions that reflect risk 1.195-21insight The staff will not approve proposals that would reduce the defense in depth deemednecessary to provide adequate protection for public health and safet 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 wereapproved by the staff during initial facility licensing or in subsequent licensing proceedings may be used in performing the radiological analyses identified by this guide provided such values remain relevant to the particular accident, its release points, and receptor locatio Methodologies that have been used for determining /Q values are documented in Regulatory Guides 1.3, 1.4, and1.145, "Atmospheric Dispersion Models for Potential Accident Consequence Assessments atNuclear Power Plants" (Refs. 5, 6, and 26), and in the Murphy-Campe paper, "Nuclear PowerPlant Control Room Ventilation System Design for Meeting General Criterion 19" (Ref. 19). Regulatory Guide 1.145 (Ref. 26) and Regulatory Guide 1.XXX, "Atmospheric RelativeConcentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants"(Ref. 27) should be used if the FSAR /Q values are to be revised or if values are to be determinedfor new release points or receptor distance For stack releases, fumigation should be considered where applicable for the EAB and LP For the EAB, the assumed fumigation period should be timed to be included in the 2-hour time perio The NRC computer code PAVAN (Ref. 28)

implements Regulatory Guide 1.145 and its use is acceptable to the NRC staf Regulatory Guide 1.XXX provides guidance on determining control room /Q value The NRC computer codeARCON96 (Ref. 24) may be used in determining control room /Q value Meteorological datacollected in accordance with the site-specific meteorological measurements program described in the facility FSAR should be used in generating accident /Q value Additional guidance isprovided in Regulatory Guide 1.23, "Onsite Meteorological Programs" (Ref. 29). IMPLEMENTATIONThe purpose of this section is to provide information to applicants and licensees regardingthe NRC staff's plans for using this regulatory guid No backfitting is intended or approved in connection with the issuance of this guide.Except when an applicant or licensee proposes an acceptable alternative method forcomplying with the specified portions of the NRC's regulations, the methods described in this guide will be used by the NRC staff in the evaluation of radiological consequences of design basis accidents at light-water nuclear power reactors for which the construction permit or license application is docketed after the issue date of this guide and at plants for which the licensee voluntarily commits to the provisions of this guid .195-22REFERENCES{See the inside front cover of this guide for information on obtaining NRC documents.}1.J.J. DiNunno et al., "Calculation of Distance Factors for Power and Test Reactor Sites,"USAEC TID-14844, U.S. Atomic Energy Commission (now USNRC), 1962.2.USNRC, "Clarification of TMI Action Plan Requirements," NUREG-0737, November1980.3.USNRC, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents atNuclear 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) August 16,1995.5.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequences of aLoss of Coolant Accident for Boiling Water Reactors," Regulatory Guide 1.3, Revision 2,June 1974.6.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequences of aLoss of Coolant Accident for Pressurized Water Reactors," Regulatory Guide 1.4,Revision 2, June 1974.7.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequences of aSteam Line Break Accident for Boiling Water Reactors," Regulatory Guide 1.5,March 1971.8.USNRC, "Assumptions Used for Evaluating the Potential Radiological Consequences of aFuel 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 198 .195-2313.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).15.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 Concentration andDose Conversion Factors for Inhalation, Submersion, and Ingestion," Federal GuidanceReport 11, EPA-520/1-88-020, Environmental Protection Agency, 198 .K.F. Eckerman and J.C. Ryman, "External Exposure to Radionuclides in Air, Water, andSoil," Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency,1993. 18. USNRC, "Steam Generator Tube Integrity," Draft Regulatory Guide DG-1074, December1998.19.K.G. Murphy and K.W. Campe, "Nuclear Power Plant Control Room Ventilation SystemDesign for Meeting General Criterion 19," published in Proceedings of 13th AEC AirCleaning Conference, Atomic Energy Commission (now USNRC), August 1974.20.USNRC, "Computer Codes for Evaluation of Control Room Habitability (HABIT V1.1),"Supplement 1 to NUREG/CR-6210, November 1998.21.S.L. Humphreys et al., "RADTRAD: A Simplified Model for Radionuclide Transport andRemoval and Dose Estimation," NUREG/CR-6604, USNRC, April 1998.22.USNRC, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Unitsof Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water- Cooled Nuclear Power Plants," Regulatory Guide 1.52, Revision 3, June 2001.23.USNRC, "Laboratory Testing of Nuclear-Grade Activated Charcoal," NRC Generic Letter99-02, June 3, 1999.24.J.V. Ramsdell and C.A. Simonen, "Atmospheric Relative Concentrations in BuildingWakes," NUREG-6331, Revision 1, USNRC, May 1997.25.ICRP, "Report of Committee II on Permissible Dose for Internal Radiation," ICRPPublication 2, 1959.26.USNRC, "Atmospheric Dispersion Models for Potential Accident ConsequenceAssessments at Nuclear Power Plants," Regulatory Guide 1.145, Revision 1,November 198 .195-2427.USNRC, "Atmospheric Relative Concentrations for Control Room RadiologicalHabitability Assessments at Nuclear Power Plants," Regulatory Guide 1.XXX, [issue date]. 28.T.J. Bander, "PAVAN: An Atmospheric Dispersion Program for Evaluating Design BasisAccidental Releases of Radioactive Materials from Nuclear Power Stations,"NUREG-2858, USNRC, November 1982.29.USNRC, "Onsite Meteorological Programs," Regulatory Guide 1.23, February 197 Footnote 6 of Regulatory Position 3.2 provides further details concerning assumptions applicable for crediting spray remova A-1Appendix AASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCESOF LWR LOSS-OF-COOLANT ACCIDENTS 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 syste Leaks up to a double-ended rupture of the largest pipe of the reactor coolant system are include The LOCA, as with all design basis accidents (DBAs), is a conservative surrogate accident that is intended to challenge selective aspects of the facility desig Analyses are performed using a spectrum of break sizes to evaluate fuel and emergency core cooling system performanc 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 radionuclides from thefuel 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 assumed tomix instantaneously and homogeneously throughout the free air volume of the primary containment in PWRs or the drywell in BWR This distribution should be adjusted if there are internal compartments that have limited ventilation exchang 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 credite An acceptable model for removal of iodine and particulates 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 systems thathave been designed and are maintained in accordance with Chapter 6.5.2 of the SRP1 A-2(Ref. A-1) may be credite An acceptable model for the removal of iodine andparticulates is described in Chapter 6.5.2 of the SRP. The evaluation of the containment sprays should address areas within the primarycontainment that are not covered by the spray drop 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 justifie 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 the maximumiodine 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 wate This equilibrium is determined by the effective iodine partition coefficien If the methodology in Revision 0 of Chapter 6.5.2 of the SRP (Ref. A-1) is used, the maximum iodine activity in primary containment atmosphere is the iodine activity, described in Regulatory Position 3.2, before the application of the 50%

reduction assumed instantaneously deposited on the walls of containmen .4Reduction in airborne radioactivity in the containment by in-containment recirculation filtersystems 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 suppression poolscrubbing in BWRs is given in Section 6.5.5 of the SRP (Ref. A-1). For suppression 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, orother engineered safety features not addressed above, should be evaluated on an individual case basi 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 containment designs)should be assumed to leak at the peak pressure technical specification leak rate for the first 24 hour2.777778e-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 rat 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 rat Leakage from subatmospheric containments is assumed to terminate when the containment is brought to and maintained at a subatmospheric condition as defined by technical specification A-32.8If the primary containment is purged during power operations, releases via the purgesystem prior to containment isolation should be analyzed and the resulting doses summed with the postulated doses from other release path 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 LOC This inventory should be based on the technical specification reactor coolant system equilibrium activit Iodine spikes need not be considere . ASSUMPTIONS ON DUAL CONTAINMENTSFor facilities with dual containment systems, the acceptable assumptions related to thetransport, 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 byengineered safety feature (ESF) filters, if any, and released to the environment via the secondary containment exhaust system during periods in which the secondary containment has a negative pressure as defined in technical specification 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 secondary containment 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 maintain anegative pressure should be evaluated on a case-by-case basi The wind speed to be assumed is the 1-hour average value that is exceeded only 5% of the total number of hours in the data se 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 adequate means tocause mixing can be demonstrate Otherwise, the leakage from the primary containment should be assumed to be transported directly to exhaust systems without mixin 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 be evaluatedat the bypass leak rate incorporated in the technical specification If the bypass leakage is through water, e.g., via a filled piping run that is maintained full, credit for retention of A-4iodine and particulates may be considered on a case-by-case basi Similarly, deposition ofparticulate 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 secondary containmentbecause of ESF filter systems may be taken into account provided that these 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 are assumed toleak during their intended operatio This release source includes leakage through valve packing glands, pump shaft seals, flanged connections, and other similar component This release source may also include leakage through valves isolating interfacing systems (Ref. A-4). The radiological consequences from the postulated leakage should be analyzed and combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOC 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 reactor powerlevel, is mixed instantaneously and homogeneously in the primary containment sump water (in PWRs) or the suppression pool (in BWRs) at the start of the acciden In lieu of this deterministic approach, suitably conservative mechanistic models for the transport of airborne activity in containment to the sump water may be use 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 from allcomponents in the ESF recirculation systems above which the technical specifications, or licensee commitments to item III.D.1.1 of NUREG-0737 (Ref. A-5), would requiredeclaring such systems out of servic 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 check 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 terminate 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 the liquidthat becomes airborne should be assumed equal to the fraction of the leakage that flashes to vapo 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:

A-5fgffhhhFF21Where: hf1 is the enthalpy of liquid at system design temperature and pressure; hf2 is the enthalpy of liquid at saturation conditions (14.7 psia, 212F); and hfg is the heat ofvaporization at 212F.4.4If the temperature of the leakage is less than 212°F or the calculated FF is less than 10%,the amount of iodine that becomes airborne should be assumed to be 10% of the total iodine activity in the leaked fluid unless a smaller amount can be justified based on the actual sump pH history and area ventilation rates.4.5The radioiodine that is postulated to be available for release to the environment is assumedto be 97% elemental and 3% organi Reduction in release activity by dilution or holdup within buildings, or by ESF ventilation filtration systems, may be credited where applicabl Filter systems used in these applications should be evaluated against the guidance of Regulatory Guide 1.52 (Ref. A-2) and Generic Letter 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 may result ina radioactivity releas 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 LOC 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 leakage shouldbe assumed to be that activity determined to be in the drywell for evaluating containment 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 leakage should be assumed to continue for the duration of the acciden 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 of reduction in concentration allowed will be evaluated on an individual-case basi 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 MSIV leakagecontrol system, or as described in Section 5.5 below, the MSIV leakage should be assumed to be released to the environment as an unprocessed, ground-level releas Holdup and A-6dilution in the turbine building should not be assumed.5.5A reduction in MSIV releases that is due to holdup and deposition in main steam pipingdownstream of the MSIVs and in the main condenser, including the treatment of air ejector effluent by offgas systems, may be credited if the components and piping systems used in the release path are capable of performing their safety function during and following a safe shutdown earthquake (SSE). The amount of reduction allowed will be evaluated on an individual case basi 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 analyze 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 evaluate If the primary containment purging is required within 30 days of the LOCA, the results of this analysis should be combined with consequences postulated for other fission product release paths to determine the total calculated radiological consequences from the LOC 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 Adsorption Unitsof 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 Letter 99-02, 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, 199 (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 the radiologicalconsequences of a fuel handling accident at light-water reactor These assumptions supplement the guidance provided in the main body of this guide.1. SOURCE TERMAcceptable assumptions regarding core inventory and the release of radionuclides from thefuel are provided in Regulatory Position 3 of this guid The following assumptions also apply.1.1The number of fuel rods damaged during the accident should be based on a conservativeanalysis that considers the most limiting cas This analysis should consider parameters such as the weight of the dropped heavy load or the weight of a dropped fuel assembly (plus any attached handling grapples), the height of the drop, and the compression, torsion, and shear stresses on the irradiated fuel rod 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 Position 3.2 ofthis guide and the estimate of the number of fuel rods breache All the gap activity in the damaged rods is assumed to be instantaneously release Radionuclides that should be considered include xenons, kryptons, and iodines.1.3The iodine gap inventory is composed of elemental (99.75%) and organic species (0.25%).1.4The radioactive material available for release is assumed to be from the assemblies with thepeak inventor The fission product inventory for the peak assembly represents an upper limit valu The inventory should be calculated assuming the maximum achievable operational power history at the end of core life immediately preceding shutdow 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, the decontaminationfactor for the elemental and organic species are 400 and 1, respectively, giving an overall effective decontamination factor (DF) of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). If the depth of water is not at least 23 feet, the DF will have to be determined on a case-by-case basis (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 dat These analyses should consider the time for the radioactivity concentration to reach levels corresponding to the monitorsetpoint, instrument line sampling time, detector response time, diversion damper alignment time, and filter system actuation, asapplicabl Containment isolation does not imply containment integrity as defined by technical specifications for non-shutdown modes. The term isolation is used here collectively to encompass both containment integrity and containment closure, typically in placeduring shutdown period 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). 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 is assumed tobe released to the environment over a 2-hour time perio The release rate is generally assumed to be a linear or exponential function over this time period. 4.2A reduction in the amount of radioactive material released from the fuel pool by engineeredsafety feature (ESF) filter systems may be taken into account provided these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2, B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system1 should be determined and accounted for in theradioactivity release analyse .3The radioactivity release from the fuel pool should be assumed to be drawn into the ESFfiltration system without mixing or dilution in the fuel buildin If mixing can be demonstrated, credit for mixing and dilution may be considered on a case-by-case basi 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 isolated2 during fuel handling operations, no radiologicalconsequences need to be analyze Technical specifications that allow such operations usually include administrative controls to close the airlock, hatch, or openpenetrations within 30 minute Such adminstrative controls generally require that a dedicated individual be present, with necessary equipment available, to restore containment closure should a fuel handling accident occu Radiological analysesshould generally not credit this manual isolation.B-35.2If the containment is open during fuel handling operations, but designed to automatically isolate in the event of a fuel handling accident, the release duration should be based on delays in radiation detection and completion of containment isolatio If it can be shown that containment isolation occurs before radioactivity is released to the environment,1 noradiological consequences need to be analyzed for the isolated pathway.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 reactor cavitypool to the containment is released to the environment over a 2-hour time perio The release rate is generally assumed to be a linear or exponential function over this time period. 5.4A reduction in the amount of radioactive material released from the containment by ESFfilter systems may be taken into account provided that these systems meet the guidance of Regulatory Guide 1.52 and Generic Letter 99-02 (Refs. B-2 and B-3). Delays in radiation detection, actuation of the ESF filtration system, or diversion of ventilation flow to the ESF filtration system should be determined and accounted for in the radioactivity release analyse .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 basi Such credit is generally limited to 50% of the containment free volum 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 plenum B-4Appendix B REFERENCESB-1G. Burley, "Evaluation of Fission Product Release and Transport," Staff Technical Paper,197 (NRC Accession number 8402080322 in NUDOCS in NRC's Public DocumentRoom.)B-2USNRC, "Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Unitsof Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water- Cooled Nuclear Power Plants," Regulatory Guide 1.52, Revision 3, June 2001.B-3USNRC, "Laboratory Testing of Nuclear Grade Activated Charcoal," Generic Letter 99-02,June 3, 199 The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximumtechnical specification values, whichever maximizes the radiological consequence In determining the dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be include Activity from projectedfuel damage should not be include If there are forced flow paths from the turbine or condenser, such as unisolated mechanical vacuum pumps or unprocessed airejectors, the leakage rate should be assumed to be the flow rate associated with the most limiting of these path 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 reactor These assumptions supplement the guidance provided in the main body of this guide.1.Assumptions acceptable to the NRC staff regarding core inventory are in RegulatoryPosition 3 of this guid The activity released from the breached, but unmelted, 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 ga The activity release attributed to fuel melting should be based upon the fraction of the fuel material that reaches or exceeds fuel melting temperatur For the secondary system release pathway, 100% of the noble gases and 50% of the iodines contained in that fuel material fraction are released to the reactor coolan .If no or minimal1 fuel damage is postulated for the limiting event, the released activityshould be the maximum coolant activity (typically a pre-accident 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 as follows.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 pressure vessel.3.2Credit should not be assumed for partitioning in the pressure vessel or for removal by thesteam separators.3.3Of the activity released from the reactor coolant within the pressure vessel, 100% of thenoble gases and 10% of the iodine are assumed to reach the turbine and condenser .4Of the activity that reaches the turbine and condensers, 100% of the noble gases and 10%of the iodine are available for release to the environmen The turbine and condensers leak to the environment as a ground-level release at a rate of 1% per day2 for a period of24 hours, at which time the leakage is assumed to terminat No credit should be assumed for dilution or holdup within the turbine buildin Radioactive decay during holdup in the turbine and condenser may be assume C-23.5In lieu of the transport assumptions provided in Sections 3.2 through 3.4 above, a moremechanistic analysis may be used on a case-by-case basi Such analyses account for the quantity of contaminated steam carried from the pressure vessel to the turbine and condensers based on a review of the minimum transport time from the pressure vessel to the first main steam isolation valve (MSIV) and considers MSIV closure tim .6The iodine species released from the reactor coolant within the pressure vessel should beassumed to be 5% particulate, 91% elemental, and 4% organi The release from the turbine and condenser should be assumed to be 97% elemental and 3% organi Minimal fuel damage is defined as an amount of damage that will yield reactor coolant system activity concentration levelsless 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 consequence In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be include Activity from projected fuel damage should not be included.D-1Appendix DASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF ABWR MAIN STEAM 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 reactor 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 guid 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 minimal1 fuel damage is postulated for the limiting event, the released activityshould be the maximum coolant activity allowed by technical specificatio The iodine concentration in the primary coolant is assumed to correspond to the following two cases in the nuclear steam supply system vendor's standard technical specifications.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 andhomogeneously in the reactor coolan Noble gases should be assumed to enter the steam phase instantaneously.2. TRANSPORTAssumptions acceptable to the NRC staff related to the transport, reduction, and release ofradioactive material to the environment are as follows.2.1The main steam line isolation valves (MSIV) should be assumed to close in the maximumtime allowed by technical specifications.2.2The total mass of coolant released should be assumed to be that amount in the steam lineand connecting lines at the time of the break plus the amount that passes through the valves prior to closur D-22.3All the radioactivity in the released coolant should be assumed to be released to theenvironment instantaneously as a ground-level releas No credit should be assumed for plateout, holdup, or dilution within facility building .4The iodine species released from the main steam line should be assumed to be 5%particulate, 91% elemental, and 4% organi Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with theguidance 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 analyse Minimal fuel damage is defined as an amount of damage that will yield reactor coolant system activity concentration levels lessthan the maximum technical specification limits. The activity assumed in the analysis should be based on the activity associatedwith the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequence In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be include Activity from projected fuel damage should not be included.E-1Appendix EASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF APWR STEAM 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 reactor 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 minimal2 fuel damage is postulated for the limiting event, the activity releasedshould be the maximum coolant activity allowed by technical specificatio Two cases of iodine spiking should be assumed.1.1.1A reactor transient has occurred prior to the postulated steam generator tube rupture(SGTR) and has raised the primary coolant iodine concentration to the maximum value (typically 60 µCi/gm DE I-131) permitted by the technical specifications at full power operations (i.e., a pre-accident iodine spike case). 1.1.2The primary system transient associated with the SGTR causes an iodine spike inthe primary syste The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in Curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulate The assumed iodine spike duration should be 8 hour9.259259e-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 released instantaneouslyand homogeneously through the primary coolan 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 breache 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 technical specifications (typically 0.1 µCi/gm DE I-131).1.4Iodine releases from the steam generators to the environment should be assumed to be 97%elemental and 3% organic.2. TRANSPORTAssumptions acceptable to the NRC staff related to the transport, reduction, and release ofradioactive material to the environment are as follows:2.1The primary-to-secondary leak rate in the steam generators should be assumed to be the leakrate limiting condition for operation specified in the technical specification 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 show compliance with leak rate technical specification These tests are typically based on cool liqui Facility instrumentation used to determine leakage is typically located on lines containing room temperature liquid In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).2.3The primary-to-secondary leakage should be assumed to continue until the primary systempressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity from the unaffected steam generatorsshould be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminate 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 with theassumption of a coincident loss of offsite power. 2.5All noble gas radionuclides released from the primary system are assumed to be released tothe environment without reduction or mitigation.2.6The transport model described in this section should be utilized for iodine releases from thesteam generator This model is shown in Figure E-1 and summarized belo Partition Coefficient is defined as:PCmass of I per unit mass of liquidmass of I per unit mass of gas22E-3Steam SpaceBulk WaterPrimaryLeakageScrubbingPartitioningReleaseFigure E-1Transport Model2.6.1A portion of the primary-to-secondary leakage will flash to vapor, based on thethermodynamic conditions in the reactor and secondary coolan During periods of steam generator dryout, all of the primary-to-secondaryleakage is assumed to flash to vapor and be released to the environment with no mitigation.With regard to the unaffected steam generators used for plant cooldown, theprimary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergenc During periods of uncovery, a flash fraction should be determined.2.6.2The leakage that immediately flashes to vapor will rise through the bulk water of thesteam generator and enter the steam spac Credit may be taken for scrubbing in the generator, using the models in NUREG-0409, "Iodine Behavior in a PWR CoolingSystem Following a Postulated Steam Generator Tube Rupture Accident" (Ref. E-1)during periods of total submergence of the tubes.2.6.3The leakage that does not immediately flash is assumed to mix with the bulk water.2.6.4The radioactivity in the bulk water is assumed to become vapor at a rate that is thefunction of the steaming rate and the partition coefficien A partition coefficient foriodine of 100 may be assumed.2.7Operating experience and analyses have shown that for some steam generator designs, tubeuncovery 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 considere The impact of emergency operating procedure restoration strategies on steam generator water levels should be evaluate E-4Appendix 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, 198 Facilities licensed with, or applying for, alternative repair criteria (ARC) should use this section in conjunction with theguidance that is being developed in Draft Regulatory Guide DG-1074, "Steam Generator Tube Integrity" (Ref. F-1), foracceptable assumptions and methodologies for performing radiological analyse Minimal fuel damage is defined as an amount of damage that will yield reactor coolant system activity concentration levelsless 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 consequence In determining dose equivalent I-131 (DE I-131), only the radioiodine associated with normal operations or iodine spikes should be include 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 reactor 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 ofradionuclides from the fuel are provided in Regulatory Position 3 of this regulatory guide. 1.1If no or minimal2 fuel damage is postulated for the limiting event, the activity releasedshould be the maximum coolant activity allowed by the technical specification Two cases 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 the maximum value (typically 60 µCi/gm DE I-131) permitted by the technical specifications at full power operations (i.e., a pre-accident iodine spike case). 1.1.2The primary system transient associated with the MSLB causes an iodine spike in theprimary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in Curies per unit time) increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value (typically 1.0 µCi/gm DE I-131) specified in technical specifications (i.e., concurrent iodine spike case). A concurrent iodine spike need not be considered if fuel damage is postulate The assumed iodine spike duration should be 8 hour9.259259e-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 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> spike exceeds that available for release from the fuel pins assumed to have defect Faulted refers to the state of the steam generator in which the secondary side has been depressurized by a MSLB such thatprotective system response (main steam line isolation, reactor trip, safety injection, etc.) has occurred. F-21.2The activity released from the fuel should be assumed to be released instantaneously andhomogeneously through the primary coolan 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 breache The fuel damage estimate should assume that the highest worth control rod is stuck at its fully withdrawn position.1.3The 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 technical specifications (typically 0.1 µCi/gm DE I-131).1.4Iodine releases from the steam generators to the environment should be assumed to be 97%elemental and 3% organi 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, and release ofradioactive material to the environment are as follows.2.1The bulk water in the faulted3 steam generator is assumed to rapidly blow down to theenvironmen The duration of the blowdown is obtained from thermal-hydraulic analysis code 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) (see Ref. F-1,DG-1074), the primary-to-secondary leak rate in the steam generators should be assumed to be the leak-rate limiting condition for operation specified in the technical specification 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 maximize 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 the faultedsteam generator should be assumed to be the maximum accident-induced leakage derived from the repair criteria and burst correlation For the unaffected steam generators, the leak rate limiting condition for operation specified in the technical specifications is equally apportioned between the unaffected steam generator F-32.3The 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 converte The ARC leak rate correlations are generally based on the collection of cooled liqui Surveillance tests and facility instrumentation used to show compliance with leak rate technical specifications are typically based on cooled liqui In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).2.4The primary-to-secondary leakage should be assumed to continue until the primary systempressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity from unaffected steam generatorsshould be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.2.5All noble gas radionuclides released from the primary system are assumed to be released tothe environment without reduction or mitigation. 2.6The transport model described in this section should be utilized for releases from the steamgenerators. 2.6.1The primary-to-secondary leakage to the faulted steam generator is assumed to flashto vapor and be released to the environment with no mitigation.2.6.2With regard to the unaffected steam generators used for plant cooldown, the primary-to-secondary leakage can be assumed to mix with the secondary water without flashing during periods of total tube submergenc If the tubes are uncovered, a portion of the primary-to-secondary leakage will flash to vapor, based on the thermodynamic conditions in the reactor and secondary coolan The leakage that immediately flashes to vapor will rise through the bulk waterof the steam generator and enter the steam spac Credit may be taken for scrubbing in unaffected generators, using the models in NUREG-0409,

"Iodine Behavior in a PWR Cooling System Following a Postulated SteamGenerator Tube Rupture Accident" (Ref. F-2), during periods of totalsubmergence of the tubes.The leakage to the unaffected generators that does not immediately flash isassumed to mix with the bulk water.The radioactivity in the bulk water of the unaffected generators is assumed tobecome vapor at a rate that is the function of the steaming rate and the 4 Partition Coefficient is defined as:PCmass of I per unit mass of liquidmass of I per unit mass of gas22F-4partition coefficien A partition coefficient4 for iodine of 100 may beassumed. 2.7Operating experience and analyses have shown that for some steam generator designs, tubeuncovery may occur for a short period following any reactor trip (Ref. F-3). The potential impact of tube uncovery on the transport model parameters (e.g., flash fraction, scrubbing credit) needs to be considere The impact of emergency operating procedure restoration strategies on steam generator water levels should be evaluate F-5Appendix F REFERENCESF-1USNRC, "Steam Generator Tube Integrity," Draft Regulatory Guide DG-1074,December 1998.F-2USNRC, "Iodine Behavior in a PWR Cooling System Following a Postulated SteamGenerator Tube Rupture Accident," NUREG-0409, May 1985.F-3USNRC, "Steam Generator Tube Rupture Analysis Deficiency," Information Notice 88-31,May 25, 198 Facilities licensed with, or applying for, alternate repair criteria (ARC) should use this section in conjunction with theguidance 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 reactor 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 guid 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 not requiredas the consequences of this event are bounded by the consequences projected for the main steam line break outside containment.1.2The activity released from the fuel should be assumed to be released instantaneously andhomogeneously through the primary coolan .3Iodine releases from the steam generators to the environment should be assumed to be 97%elemental and 3% organi 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, and release ofradioactive material to the environment are as follows.2.1The primary-to-secondary leak rate in the steam generators should be assumed to be the leak-rate-limiting condition for operation specified in the technical specification 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 show compliance with leak-rate technical specification These tests are typically based on room temperature G-2liqui Facility instrumentation used to determine leakage is typically located on linescontaining cool liquid In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).2.3The primary-to-secondary leakage should be assumed to continue until the primary systempressure is less than the secondary system pressure, or until the temperature of the leakage is less than 100°C (212°F). The release of radioactivity should be assumed to continue untilshutdown cooling is in operation and releases from the steam generators have been terminated.2.4The release of fission products from the secondary system should be evaluated with theassumption of a coincident loss of offsite power.2.5All noble gas radionuclides released from the primary system are assumed to be released tothe environment without reduction or mitigation.2.6The transport model described in assumptions 2.6 and 2.7 of Appendix E should be used foriodin Facilities licensed with, or applying for, alternate repair criteria (ARC) should use this section in conjunction with theguidance that is being developed in Draft Regulatory Guide DG-1074, "Steam Generator Tube Integrity" (USNRC, December1998), 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 reactor Theseassumptions supplement the guidance provided in the main body of this guid Two release paths are considered: (1) release via containment leakage and (2) release via the secondary plan Each release path is evaluated independently as if it were the only pathway availabl The consequences of this event are acceptable if the dose from each path considered separately is less than the acceptance criterion in Table 4 in Regulatory Guide 1.19 . SOURCE TERMAssumptions acceptable to the NRC staff regarding core inventory are in RegulatoryPosition 3 of this guid The activity released from the breached but unmelted 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 ga The activity release attributed to fuel melting should be based on the fraction of the fuel material that reaches or exceeds fuel melting temperatur For this fuel material fraction, the assumption is that 100% of the noble gases and 25% of the iodines contained in that fraction are available for release from containmen For the secondary system release pathway, 100% of the noble gases and 50% of the iodines contained in that fuel material fraction are released to the reactor coolant.1.1If no fuel damage is postulated for the limiting event, a radiological analysis is not requiredas the consequences of this event are bounded by the consequences projected 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 be assumed tobe released instantaneously and homogeneously through the containment atmospher 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 iodid 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 to be 97%elemental and 3% organi H-22. TRANSPORT FROM CONTAINMENTAssumptions acceptable to the NRC staff related to the transport, reduction, and release ofradioactive material in and from the containment are as follows.2.1A reduction in the amount of radioactive material available for leakage from thecontainment that is due to natural deposition, containment sprays, recirculating filter systems, dual containments, or other engineered safety features may be taken into accoun 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 the technicalspecifications at peak accident pressure for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at 50% of this leak rate for the remaining duration of the acciden Peak accident pressure is the maximum pressure defined in the technical specifications for containment leak testin 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, and release 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 for operationspecified in the technical specifications should be assumed to exist until shutdown 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 show compliance with leak rate technical specification These tests typically are based on cooled liqui The facility's instrumentation used to determine leakage typically is located on linescontaining cool liquid In most cases, the density should be assumed to be 1.0 gm/cc (62.4 lbm/ft3).3.3All noble gas radionuclides released to the secondary system are assumed to be released tothe environment without reduction or mitigation.3.4The transport model described in assumptions 2.6 and 2.7 of Appendix E should be usedfor iodin 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 REGULATORY ANALYSISA daft regulatory analysis was published with the draft of this guide when it was issuedfor public comment (DG-1113, January 2002). No changes were necessary, so a separate value/impact statement for this regulatory guide has not been prepare A copy of DG-1113 and the value/impact statement is available for inspection or copying for a fee in the NRC's PublicDocument Room at 11555 Rockville Pike, Rockville, M An electronic version of DG-1113 is available in the NRC's Electronic Reading Room under accession number ML020160023.