Regulatory Guide 1.154

Format and Content of Plant-specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors

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Issue date:	01/31/1987
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U.S. NUCLEAR REGULATORY COMMISSION January 1987REGULATORY GUIDEOFFICE OF NUCLEAR REGULATORY RESEARCHREGULATORY GUIDE 1.154(Task SI 5024)FORMAT AND CONTENT OF PLANT-SPECIFICPRESSURIZED THERMAL SHOCK SAFETY ANALYSISREPORTS FOR PRESSURIZED WATER REACTORSUSNRC REGULATORY GUIDESRegulatory Guides are Issued to describe and make available to thepublic methods acceptable to the NRC staff of Implementingspecific parts of the Commission's regulations, to delineate tech-niques used by the staff In evaluating specific problems or postu-lated accidents. or to provide guidance to applicants. RegulatoryGuides are not substitutes for regulations, and compliance withthem Is not required. Methods and solutions different from those setout In the guides will be acceptable If they provide a basis for thefindings requisite to the Issuance or continuance of a permit orlicense by the CommissIon.This gulde was Issued after consideration of comments received fromthe pubiic. Comments and suggestlons for Improvements In theseguides are encouraged at all times, and guides wiil be revised, asappropriate, to accommodate comments and to reflect new Informa-tion or experience.Written comments may be submitted to the Rules and ProceduresBranch, DRR, ADM, U.S. Nuclear RegulatorY Commission,Washington, DC 20555.The guides are Issued In the following ten broad divisions:1. Power Reactors 6. Products2. Research and Test Reactors 7. Transportation3. Fuels and Materials Facilities 8. Occupational Health4. Environmental and Siting 9. Antitrust and Financial Review5. Materials and Plant Protection 10. GeneralCopies of Issued guides may be purchased from the GovernmentPrinting Office at the current GPO price. Information on currentGPO prices may be obtained by contacting the Superintendent ofDocuments, U.S. Government Printing Office, Post Office Box37082, Washington, DC 20013-7082, telephone 202)275-2060 or(202)275-2171.Issued guides may also be purchased from the National TechnicalInformation Service on a standing order basis. Details on thisservice may be obtalned by writing NTIS, 5285 Port Royal Road,Springfield, VA 22161.-

Table of ContentsPageINTRODUCTION ............................Background and

Purpose

of This Guide. .............Objectives of Plant-Specific PTS Safety Analysis Reports. ...Staff Review of Plant-Specific PTS Safety Analysis Reports andAcceptance Criteria for Continued Operation .........Recommended Format. ......................CHAPTER 1 OVERALL APPROACH, SCOPE OF ANALYSIS, AND REPORTORGANIZATION ..................CHAPTER 2 PLANT DATA. ....................2.12.22.32.42.52.6Systems Pertinent to PTS. .....Reactor Vessel. ..........Fluence ..............Inservice Inspection Results. ...Plant Operating Experience. ....Operating Procedures. .......vvvviviii1.154-11.154-31. 154-31.154-31. 154-41.154-41.154-41.154-5DETERMINATION OF DETAILED PTS SEQUENCES FORANALYSES. ........................1.154-63.1 Approach Used-. ...............3.2 Sequence Delineation. ............1.154-61. 154-63.2.1 Development of Classes of Initiators .3.2.2 Identification of Important InitiatorVariations ..............3.2.3 Definition of Potential TransientsResulting from Each Initiator. ......1.154-6..1.154-7..1.154-73.3 Operator Effects. ..............3.4 Sequence Quantification ...........1.154-81.154-91.154-91.154-91. 154-103.4.1 Initiating Events. .........3.4.2 Equipment Failures .........3.4.3 Operator Actions ..........3.5 Event Tree Collapse .............3.5.1 Specific Sequences ..........3.5.2 Residual Groups. ...........1.154-111.154-111.154-11i i iCHAPTER 3............................................................

Table of Contents (Continued)CHAPTER 4 THERMAL-HYDRAULIC ANALYSIS ...............4.1 Thermal-Hydraulic Analysis Plan ...........4.2 Thermal-Hydraulic Model .............4.3 Simplified Analysis Methods .............4.4 Thermal Stratification Effects. ...........4.5 Thermal-Hydraulic Analysis Results. ........CHAPTER 5 FRACTURE MECHANICS ANALYSIS ...............CHAPTER 6 INTEGRATION OF ANALYSES .................CHAPTER 7SENSITIVITY AND UNCERTAINTY ANALYSES OF THROUGH-WALL CRACKFREQUENCY ........................7.1 Sensitivity Analysis. ............. ...7.2 Uncertainty Analysis. ................Page1.154-121.154-121.154-121.154-131.154-141.154-141.154-161.154-191. 154-201.154-201.154-20CHAPTER7.2.1 Parameter Uncertainties. ...........7.2.2 Modeling Uncertainties (Biases). .......8 EFFECT OF CORRECTIVE ACTIONS ON VESSEL THROUGH-WALLCRACK FREQUENCY .....................8.1 Flux Reduction Program. ..........8.2 Operating Procedures and Training ProgramImprovements. ...............8.3 Inservice Inspection and NondestructiveExamination Program ............8.4 Plant Modifications ............8.5 In Situ Annealing .............CHAPTER 9 FURTHER ANALYSES ....................1.154-201.154-211.154-23 -x...1.154-23 >--)...1. 154-231. 154-241.154-241.154-251.154-26CHAPTER 10 RESULTS AND CONCLUSIONS REGARDING PTS ANALYSES. .10..1 Summary of Analysis. ............10.2 Basis for Continued Operation. .......REFERENCES .........................REGULATORY ANALYSIS. ....................1.154- 71.154-271.154-271.154-281.154-30K)iv............

INTRODUCTIONBackground and

Purpose

of This GuideThe pressurized thermal shock (PTS) rule, § 50.61 of 10 CFR Part 50issued on July 23, 1985 (50 FR 29937), establishes a screening criterionbased on reactor vessel nil-ductility-transition temperature (RTNDT). Thescreening criterion was established after extensive industry and NRC analysesregarding the likelihood of vessel failure due to PTS events in pressurizedwater reactors (PWRs). The analyses were applied generically and containedconservative assumptions to make the results bounding for any PWR. Based onthe results, the NRC concluded that the risk due to PTS events is acceptableat any plant so long as the RTPTS* of the reactor pressure vessel remainsbelow the screening criterion.Extensive safety analyses are required by the rule for any plant thatwishes to operate with RTPTS values above the screening criterion. The recom-mended methods to be used in performing the analyses are outlined in thisguide. The purpose of the analyses is to assess the risk due to PTS eventsfor proposed operation of the plant with reactor vessel RTPTS above the screen-ing criterion. Effective 1 year after the publication of this regulatory guide,Section 50.61 requires that these analyses be completed 3 years before thescreening criterion would be exceeded to allow adequate time for implementationQ on the plant of any corrective actions assumed in the analyses before the plantoperates above the screening criterion.This regulatory guide describes a format and content acceptable to the NRCstaff for these plant-specific PTS safety analyses and describes acceptancecriteria that the NRC staff will use in evaluating licensee analyses and pro-posed corrective measures.The references listed in this guide include a set of analyses sponsored bythe NRC that, taken together, constitute an example of the analyses describedby this guide. The staff recommends that these references be extensively used,along with this guide, by those performing the plant-specific PTS analyses re-quired by the PTS rule, § 50.61. References 1, 2, and 3, for example, eachrepresent an analysis by the Oak Ridge National Laboratory (ORNL) predictingthrough-wall crack frequency for one PWR. These references will provide guid-ance through the analyses. Reference 3 (analysis of H. B. Robinson) should bemost helpful because it was the last one performed and includes the experiencegained in performing the two earlier analyses.Objectives of Plant-Specific PTS Safety Analysis ReportsParagraph 50.61(b)(4) requires that a licensee whose plant will exceed thescreening criterion before expiration of the operating license submit safetyanalyses to determine what, if any, modifications to equipment, systems, and*To avoid confusion among several (preexisting) slightly different definitionsV.- _ °f RTNDT, § 50.61 contains its own definition of an RTNDT (called RTPTS) tobe used when comparing plant-specific vessel material properties with the PTSscreening criterio operation are necessary to prevent potential failure of the reactor vessel as aresult of postulated PTS events if continued operation beyond the screeningcriterion is allowed. These analyses must include the effects of all correctiveactions the licensee believes necessary to achieve an acceptable PTS-relatedrisk for continued operation of the plant. The final objective of the plant-specific PTS study, therefore, is to justify continued operation of the plantby demonstrating that the likelihood of a through-wall crack during continuedoperation is acceptably low. The study must include calculation, for the re-mainder of plant life, of the expected frequency of through-wall cracks due toPTS.In calculating these results, it will be necessary to:o Identify the dominant accident sequences.o Identify operator actions, control actions, and plant features impor-tant to PTS.o Estimate the effectiveness of potential corrective actions in reduc-ing the expected frequency of through-wall cracks.o Identify the sources and approximate magnitude of the majo; uncertain-ties and their effects on the conclusions.o Present and justify the licensee's proposed program for correctivemeasures.o Present and justify the licensee's proposed basis for continued opera-tion at embrittlement levels above the screening criterion. Thismust include comparison with the acceptance criteria described belowof the PTS-related through-wall crack frequency with correctiveactions implemented as necessary.Staff Review of Plant-Specific PTS Safety Analysis Reports and AcceptanceCriteria for Continued OperationThe PTS rule specifies a screening criterion based on RTNDT (called RTPTSfor use as defined within the rule) of 270°F for axial weld and plate materialsand 300°F for circumferential weld materials. As detailed in SECY-82-465(Ref. 4), these values were selected based on generic studies of the expectedfrequency and character of a wide spectrum of transients and accidents thatcould cause pressurized overcooling of the reactor vessel (PTS events) and onoperating experience data. The risk due to PTS events was assessed in termsof probabilistic fracture mechanics calculations of the expected frequency ofthrough-wall crack penetration of the pressure vessel due to the PTS events.In selecting the screening criterion based on those calculations, the conserva-tive assumption was made that any through-wall crack could result in severecore degradation or melt. Core melt itself was viewed as an event to beavoided even though risk to the public due to such an event in terms of person-rems and early and late fatalities was not calculated with any certainty. Theestimated through-wall crack frequency developed as a function of RTNDT foraxial welds (Fig. 8.3 of Ref. 4) is shown in Figure 1. 9vi LONGITUDINAL CRACK EXTENSION NO ARRESTi2SECY-82-465 PRA RESULTSLEGEND:O PRA TOTALO STEAM LINE BREAKS6 S.G. TUBE RUPTURESio-3 V SBLOCA W/WPS-EXTENDED HPI4~ 0 L.cOo-,1~~~~~~~~~110'6175 200 225 250 275 300 325MEAN SURFACE RTNDT(0F)Figure 1vii The RTPTS screening criterion selected by the staff corresponds to a mean(or average) "best estimate" surface RTPTS of 210°F. The staff used a "2-sigma"value (spread between "best estimate" and "upper limit") of 600F;* thus thescreening criterion expressed in terms of RTPTS, which, by definition, is thisupper limit value, was selected at 210 + 60 = 270°F. For axial weld and platematerials, Figure 1 gives a through-wall crack frequency of about 5 x 10-6 perreactor year at 210°F, which corresponds with an RTPTS of 270°F. For circum-ferential welds, the same frequency is believed to be bounded by an RTPTS ofapproximately 300°F (Ref. 4). The Commission concluded that the PTS-relatedrisk at any PWR is acceptable so long as the RTPTS values remain below thespecified screening criterion.It was realized that there are many unknowns and uncertainties inherent inthe probabilistic calculations; thus it was with deliberate intent that conser-vative assumptions such as those stated above were made. The expectation wasthat the true risk at any plant due to PTS events would in all likelihood beconsiderably below that derived from Figure 1 and would therefore be acceptable.Also contributing to the belief that the real PTS risk at any given plant waslower than that resulting from the analysis in Reference 4 was the belief thatmany of the generic plant assumptions made in Reference 4 (e.g., materialproperties, system performance, crack distribution) would prove to be overcon-servative for analysis of a specific plant and that the resulting plant-specificanalysis, when performed, is likely to result in a reduced prediction of PTSrisk.If the plant-specific PTS analyses submitted by licensees in accordancewith § 50.61 using the methodology described in this guide (or acceptable equi-valent methodology) predict that the PTS-related, through-wall crack penetrationmean frequency will remain less than 5 x 10-6 per reactor year for the requestedperiod of continued operation, such operation would be acceptable to the staff.In all the analyses performed, the licensee must justify that the impor-tant input values used are valid for the remaining life of the plant.Recommended FormatThe recommended content of plant-specific PTS safety analyses is presentedin Chapters 1 through 10 of this guide. Use of this format by licensees willhelp ensure the completeness of the information provided, will assist the NRCstaff in locating the information, and will aid in shorteni:q the time neededfor the review process. If the licensee chooses to adopt this format, thenumbering system of this guide should be followed at least down to the sectionlevel. Certain sections may be omitted if they are clearly unnecessary to pro-vide for comprehension of the analysis or if they are repetitive.*Based on preliminary RTNDT data from many plants (see Table P.1 of Enclosure Ato Ref. 4).viii Additional guidance on style, composition, and specifications of safetyanalysis reports is provided in the Introduction of Revision 3 to RegulatoryGuide 1.70, "Standard Format and Content of Safety Analysis Reports for NuclearPower Plants (LWR Edition)."The Advisory Committee on Reactor Safeguards has been consulted concerningthis guide and has concurred in the issuance of this regulatory guide.Any information collection activities mentioned in this regulatory guideare contained as requirements in 10 CFR Part 50, which provides the regulatorybasis for this guide. The information collection requirements in 10 CFR Part 50have been cleared under OMB Clearance No. 3150-0011.ix 1. OVERALL APPROACH, SCOPE OF ANALYSIS, AND REPORT ORGANIZATIONW This chapter is to describe the overall approach to the analysis and out-line the individual tasks in terms of the nature and source of input, the methodsused for analysis, and the nature and subsequent use of the output. The inter-relationship of the tasks should be described and should be illustrated by aflow chart. How the analysis tasks are integrated to achieve the results andconclusions is to be described.Major emphasis should be placed on analyzing event sequences leading tovessel through-wall cracking and corrective actions to prevent this fromoccurring.The report should include both probabilistic and deterministic fracturemechanics analyses. The probabilistic analyses should be used to determinethe statistical likelihood of vessel through-wall crack penetration assuming acrack size distribution appropriately justified for the vessel being analyzedand appropriate uncertainties and distribution of the significant input param-eter such as material properties. The deterministic analyses should be usedto evaluate the critical time interval in the transient during which mitigat-ing action can be effective. The deterministic analyses should be carried outusing the two sigma upper and lower bounding values of the appropriate param-eters such as fluence, copper content, nickel content, fracture initiationtoughness, fracture arrest toughness, and ductile fracture toughness.The input to the probabilistic analysis should be best estimates based on*lli appropriate assumptions. Uncertainties and conservatisms should be explicitlypresented in the decision rationale for the licensee's proposed corrective mea-sures and basis for continued operation.The analysis should include effects of operator actions, control systeminteractions, and support systems such as electric power, instrument air, andservice water cooling.The report should be organized by starting with a description in Chapter 1of how the report chapters and supporting appendices are interrelated and whatmaterial is in the appendices.The main report should describe the objectives and overall approach usedin the study, outline the plant systems analyzed, describe the engineering anal-yses performed, present the results obtained and conclusions drawn, and presentand justify the licensee's proposed program of corrective measures.Appendices should contain data, detailed models, sample calculations, anddetailed results needed to support the various chapters of the report. Appen-dices should contain little supporting text.- Instead, the nature and relevanceoflmaterial in the appendices should be described in the pertinent chapters ofthe main report.Throughout the guide, wherever it is specified or suggested that detailed-descriptive materials should be submitted as part of the licensee's analyses,these detailed materials may be provided by incorporation of reference material..'already submitted to the NRC (for example, in the final safety analysis report).'It remains the responsibility of the licensee to provide a coherent, readable1.154-1 document that does not unduly burden a reviewer with collecting extensivereferences before proceeding with the review. Therefore, care should beexercised in limiting such material provided by reference to the reviewer whois conducting an extensive, detailed evaluation of the submitted work.Certain details (noted in Chapter 1 and in Section 4.3 of this regulatoryguide) that have.not been previously submitted to the NRC may be made availablefor NRC inspection and may also be referenced by the submitted analyses.1. 154-2 2. PLANT-DATAThis chapter is to briefly describe plant systems and operations pertinentto PTS. Chapter 2 of Reference 3 (the H. B. Robinson analysis by ORNL) providesa good example. Supporting appendices or references are to present the designand operating data used in the analysis or needed to understand the analysis.References to other dcuments (e.g., the final safety analysis report (FSAR))should indicate specific sections. (Reliability data, however, are to be inSection 3.4, "Sequence Quantification," or its supporting appendices andreferences.)2.1 Systems Pertinent to PTSSummarize design and operating features of systems pertinent to PTS.Illustrate each system with a simplified process and instrumentation diagramor a single line diagram. Identify on each illustration any interfaces withother systems. For each system, include a table summarizing key design andoperating data. Give the maximum, minimum, and nominal values for those casesin which design data may vary with time (for example, high-pressure injection(HPI) water temperature may vary with season). Such values used in the analysisshould be identified and justified. Refer to appendices or other documents(e.g., specific sections of the FSAR) as necessary for more details. Systemsto be considered should include pertinent portions of:° Reactor cooling system° Condensate and main feedwater systemso Steam systemo Auxiliary feedwater systemo Reactor protection systemo Chemical and volume control systemo Emergency core cooling systemso Instrumentation and-control systemso Support systems-Electric power-Instrument air-Service cooling water2.2 Reactor VesselSummarize the reactor vessel construction and its material properties.Use tables, drawings, or graphs to show:o Vessel design (including weld locations and hot leg and coldleg penetrations).° Vessel materials and chemical composition in the beltline region(including both base and weld material properties).° Vessel fabrication procedures, particularly welding and cladding.1.154-3.:::-a- .,: .:=, ..r.-,:b .. .F i'r .:t O{.-.[Du>: .:,::,, r .-aX $r,:'i!-.,.,,'s; i,'£:':,'t' .:....-0'<,!.;'.Y '.C.W .-; _.,4'n's. :E* r.-.i 'V; 'sS',.: .'M.,b'_'..N',"'._":\ X. ') '..:.:, :>- .i " 'i ,;, 'ts'b.;,L 0 Vessel properties (e.g., RTNDT, initial RTNDT, appropriate fracturetoughness data, including the upper-shelf regime, residual stresses,flaw density distribution, etc.). Describe and justify methods usedto calculate or otherwise determine properties.Available information on the vessel properties should be reexamined indetail to fill any gaps in the supporting data for making an estimate of RTNDTand to support resolution of any disagreements about the validity of valuesused.Few data are currently available and validated to support the selection ofa value for the initial RTNDT. The confidence that can be placed in estimatesof the initial RTNDT depends not only on material tests but also on the accu-rate documentation of welding technique, weld wire used, and weld flux used.The credibility of such estimates could be enhanced by performing more testson archival material, by discovering previously unreported test results onweld specimens from the particular plant, or by evaluating properties of weldsconsidered typical of the plant-specific weld.2.3 FluencePresent (or incorporate by reference to a submitted report) the currentand projected fluence on the vessel using benchmarked computer programs and.methodology and information from neutron flux surveillance dosimetry. Use theweld locations and fluence values to identify the critical welds. Show how thefluence varies along the length and depth of the critical welds. Describe thebasis for these estimates and their uncertainty. These fluence values shouldbe benchmarked, for example, through use of ENDF/B-IV or V cross sections, toquantify the error.2.4 Inservice Inspection ResultsTo the extent pertinent to the probabilistic analysis nd proposed correc-tive actions, summarize:o Results -The number, size, depth, and location of any flaws foundshould be well defined and described.o Methods used -The method used to perform the inspection should bewell described with documentation of any validation information.Note: Only those inservice inspections (ISIs) that have actually been per-formed should be discussed in this section. Improved ISI programs as proposedby the licensee should be described under corrective measures in Chapter 8,"Effect of Corrective Actions on Vessel Through-Wall Crack Frequency."2.5 Plant Operating ExperienceSummarize overcooling transients that have occurred at this station andsimilar stations. Also, summarize lessons learned from these and other tran-sients, and indicate actions taken to prevent recurrence or minimize severityof overcooling transients.1.154-46.-.l 2.6 Operating ProceduresThis section provides procedural data, e.g., what the operator is supposedto do and when. This section, for example, should present and describe theimportant operator actions as defined by existing procedures associated withpotential overcooling transients. Also emphasize how the procedures wereevaluated and optimized in light of any competing risks that might arise fromevents other than PTS events to ensure that overall plant safety is appropriatelybalanced. The conditions under which the operator takes each action, the expectedtime for performing the action, and how the time was derived should be identified.Some examples of these operator actions are:o Trip reactor coolant pumps.o Throttle/terminate emergency core coolant.o Throttle/terminate main and emergency feedwater.o Restore main and emergency feedwater.o Isolate break (primary or secondary).Supply a summary of training materials associated with overcooling eventsin general and with respect to principal initiators. In addition, a summary ofsimulator exercises associated with potential overcooling events should beprovided.Note: Proposed improvements in procedures, diagnostic instrumentation, displaysystems, and operator training should be presented in Section 8.2 under theO licensee's program of corrective measures.-1.154-5 3. DETERMINATION OF DETAILED PTS SEQUENCES FOR ANALYSESThis chapter is to present the methods and analyses used to identify thosetransient sequences that could contribute significantly to the PTS risk. Agood example is presented in Chapter 3 of Reference 3. The scope includes iden-tifying initiating events, developing event trees, modeling and quantifying thereliability of relevant systems and operator actions, and collapsing the eventtrees to identify specific relevant sequences. Detailed models, data, and samplecalculations should be included in appendices or referenced. However, the logicof the analysis, criteria used, results, and insights gained are to be describedin the main report.3.1 Approach UsedDescribe how the material presented in this chapter fits into the overallPTS study. Provide a general description of the process used to identify PTSsequences. It should be made clear how the approach used will result in com-pleteness of identification of all classes of events that could contribute sig-nificantly to PTS risk, how specific events are selected for more detailed anal-ysis to represent each class, and finally how the events so analysed are usedto determine total PTS risk at the plant.3.2 Sequence DelineationIdentify potential overcooling transients in a well-defined manner, anddocument them in such a way that it is clear to a reviewer that all importantpotential overcooling conditions have been considered. Classes of initiators Ishould be developed, important variations of initiators within each class shouldbe identified, and potential transients resulting from these initiators shouldbe defined.Operating experience at the specific plant and at similar plants shouldbe carefully examined to aid in the identification of potentially significantPTS initiators, contributing failures, and potential corrective actions. TheORNL contribution to Systematic Evaluation Program reviews (Ref. 5, for example)is a technique that can be used for this purpose.3.2.1 Development of Classes of InitiatorsAny class of transients that could lead to overcooling of the reactor ves-sel should be considered in the analysis. It should, however, be appropriateto use logical arguments to eliminate classes of transients as actual PTSinitiators whenever justifiable. Examples of initiators that should be includedare:o Loss-of-coolant accidents (LOCAs), including steam generator tuberupture accidents.o Steam line breaks.o Overfeeds.o Combinations of these, including possible return to criticality.1.154-6 3.2.2 Identification of Important Initiator VariationsAfter the classes of potential initiators have been identified, it is im-portant to consider variations within any individual class. These variationsshould include:1. Decay heat level -The decay heat level, determined by recent operat-ing history of the plant, can have a major impact on the potential consequencesof a given event. Thus, various decay heat conditions should be considered.Clearly, decay heat associated with a reactor trip from full power (assumingoperation at full power for some considerable time) should be examined. Zerodecay heat represents the opposite extreme but for all practical purposes occursonly once at the beginning of life for the plant when PTS is not important.Therefore, the analyst may choose to use some other level of decay heat thatwould cover potential decay heat conditions after the initial startup of theplant. The reasons for choosing particular decay heat levels for analysisshould be documented. Each identified initiator should be examined at all decayheat levels defined whenever appropriate.2. Power level -Power level may be important since certain equipmentconditions or configurations may only exist at certain power levels, e.g., hotstandby. As in the case of decay heat level identification, the reasons forthe selection of specific power levels for analysis purposes should be stated.It should be noted that under certain conditions a reactor system may be at ahigh power level with a low decay heat condition.3. Location of event -In many instances the location of the event isdefined. For example, an event consisting of a failed open turbine bypass valvehas the location defined since it is a specific valve failure. However, forsome events such as pipe breaks, the location is not defined and could have animpact on the progression of the event. In the case in which location is notdefined, all locations that could be significant should be considered. Eachlocation should then be elimihated by logical argument, bounded by consequencesassociated with another location, or treated as a separate event.4. Magnitude of event -Many of the initiators can occur to variousdegrees. For example, a LOCA can range from a very small break to a full guil-lotine pipe break. Break sizes should be examined to identify categories ofsizes that lead to similar system conditions. In the.case of the LOCA event,special consideration should be given to the identification of break sizes thatcould lead to loop flow stagnation. The larger-sized LOCAs typically do not'contribute to PTS risk since the pressure cannot be maintained because of thelarge flow out of the break.3.2.3 Definition of Potential Transients Resulting from Each InitiatorAfter the complete set of significant initiators has been defined, eventtrees are required to identify potential sequences resulting from each initia-tor. The development of the event tree headings and branches should be donein a consistent and logical manner. This was done in the ORNL studies (Refs. 1,2, and 3) by using what have been called system state trees. These trees definethe potential states of each plant system of interest conditional on specific-':thermal-hydraulic conditions. Initiator-specific event trees can then bedeveloped by examining the system state trees with respect to each initiating1. 154-7 event. A similar or equivalent approach should be used to ensure traceabilityof the event trees and to ensure that important sequences are not inadvertentlyeliminated.Support system failures should also be presented within some type of eventtree structure. If the event trees are developed as previously described, anysupport system failure would most likely lead to a sequence of events that isalready mapped out on the event trees, but in many instances with a higher pro-bability of occurrence. In other cases, it may be necessary to define eventtrees resulting from a support system failure. In either case, it is importantthat the support systems be examined to identify their potential impact on over-cooling conditions. The results of this examination should be presented as aseparate section with the identification of specific support system failuresequences of interest. The support system review should at least include:o The electrical supply system.o The compressed air instrument system.o The component and service water systems.3.3 Operator EffectsThe operator effects are analyzed in two separate sections. In this sec-tion the potential operator actions are identified. These actions are furtheranalyzed in Section 3.4 in which the probabilities associated with the perfor-mance of an operator action are developed.The operator can improve, aggravate, or initiate an overcooling transient.All three of these categories should be discussed in this section.1. Procedures and/or the operators' general knowledge can lead to actionsthat improve the conditions associated with an overcooling event. Explanationshould be included as to why it is perceived that this action would be taken.Where appropriate, these operator actions should be either included directlyon the event trees or presented as separate operator action trees that can laterbe coupled with the principal event trees.2. Although the ORNL studies (Refs. 1, 2, and 3) did not include operator-initiated events or events aggravated by operator actions contrary to procedures,this category of events should also be examined as part of a plant-specificanalysis.3. The analyses should include a quantitative approximation of the PTSrisk resulting from operator acts of commission. Also included should be thepossibility that an operator could initiate or exacerbate some milder eventinto a more severe PTS-type event. Since there is no generally accepted wayto perform such analyses, the approximation used by the licensee for thispurpose should be discussed and justified for applicability to this particularplant. The "confusion matrix" approach (Ref. 6) used in human reliabilityanalysis could provide an acceptable structure for identifying and analyzingthese potential operator actions.1.154-8 w 3.4 Sequence QuantificationQuantify the event trees by using identified initiating event frequencies,appropriate conditional probabilities associated with the success or failureof specific equipment operations, and success and failure probabilities asso-ciated with operator actions. Plant-specific data should be used wheneverappropriate to define these probabilities, including appropriately adjustedsimulator studies. This should be supplemented by vendor-specific or PWR-generic data bases when plant-specific data do not appear to provide an adequatedata base. Reference 7 includes guidance about treatment of generic and plant-specific data. Its appendices include an updated generic data base that shouldbe used.Identify by specific reference or provide in appendices all the reliabilitydata used as input to quantify the event sequences. An explanation should besupplied as to how the data were derived for each data point.3.4.1 Initiating EventsInitiating event frequencies should be developed based on the number ofobserved events within selected periods of operation for similar plants underconsideration. If no failures have been observed and no other information isavailable with which to estimate a probability, a standard statistical methodsuch as the Poisson distribution can be used to determine a probability, or thetechnique described in Appendix B to Reference 3 for estimating plant-specificinitiating event frequencies can be used. For some initiators, it may be neces-sary to estimate the frequency of events in a particular operating mode, e.g.,hot zero power. The data should be researched to identify trends associatedwith the occurrence of the event and the operating mode. In addition, theinitiator itself should be examined to identify physical conditions that mightfavor failure in one mode rather than another. If this examination reveals noevidence of correlation between frequency and operating mode, the fraction oftime spent in each operating mode can be used as a weighting factor.3.4.2 Equipment FailuresFollowing each initiating event, certain components are designed to performin a defined manner. Failure of a component to perform its required functioncould lead-to PTS considerations. Thus, it is necessary to assign a failure andsuccessful operation probability for each component on a per-demand basis. Theseprobabilities can be obtained by estimating the number of failures observed withina period of time, combined with an estimate of the number of demands expectedwithin that same period, or by developing fault trees. If no failures havebeen observed and no other information is available with which to estimate afailure-on-demand probability, a standard statistical method can be used todevelop a probability.is, As with all event trees, the probability associated with a particular branchis conditional on the prior branches in the sequence. Questions of conditionalprobability should be carefully considered before a failure probability isOt>a' assigned.The potential for coupled or common cause failures within a system orbetween systems should be examined in the analysis. Careful consideration1.154-9 should be given to increasing the failure potential of a component, given thefailure of one or more components of the same type in the same system or inother systems being subjected to the same environment or fault causes. Asadditional components of a particular type are postulated to fail, the proba-bility for the next component of the same type to fail should increase. Basedon the ORNL analysis, a simplified approach would be to assume that the failureprobability of the second component, given that the first component has failed,might be as high as 0.1. The third component might be assumed to fail with a0.3 probability, given the failure of two identical components. One could thenassume that, after the failure of three components of the same type, all remainingcomponents of that type in the same or in other systems being subjected to thesame environment or fault causes would fail with a probability of 1.0. Thelicensee should discuss how these types of coupled failures are handled in theanalysis.Common cause failures of a different type may occur, as previously dis-cussed, through the failure of a support system or a control signal. An anal-ysis of these potential failures should be made and the branch probabilitiesshould be adjusted whenever appropriate.3.4.3 Operator ActionsOperator action probabilities are particularly difficult to determinebecause of the lack of a data base. The problem is further complicated whentime becomes an important variable. The procedure outlined below representsone approach to quantifying operator actions. This procedure should be conser-vative for any operator action performed as required by procedures assumingthat the equipment required is operational. For operator actions that mightnot be associated with procedural steps, it is not clear that this simplifiedapproach would produce conservative frequencies. Therefore, the approachdescribed would only be recommended for operator actions associated with proce-dural steps. Regardless of the method used, the human error probabilities usedin these analyses should be supported by data validated for the plant beinganalyzed.1. Identify operator actions -In this step the procedures associatedwith each initiator would be reviewed to identify those operator actions thatwould have an impact on downcomer temperature.2. Identify time constraint -In the case of each operator action, thetransient would be reviewed assuming no operator action to identify the time-frame available for successful completion of the operator action.3. Assign screening failure probabilities -In this step a conservativevalue for the failure of the operator action would be identified. For operatoractions required by procedures to be performed within the first 5 minutes ofthe transient, the time-reliability curve as presented in NUREG/CR-2815 (Ref. 7)could be used to identify a screening value. After 5 minutes, a value of 0.9for success and 0.1 for failure would be assumed for all operator actions. Theentire PTS analysis would then be completed using these screening values.4. Identify dependency factors -In some instances, there may be coupled (failures associated with operator actions just as there were coupled failures1.154-10 of an operator action may be linked, to various degrees, to the success or fail-ure of a previous operator action. Thus, it is recommended that each operatoraction be reviewed with respect to dependency. This can be accomplished usingthe dependency tables as presented in the human reliability handbook (Ref. 8).5. If any of the dominant sequences involve the failure of an operatoraction, a more comprehensive evaluation of the failure would be performed forthat operator action. When necessary, the comprehensive evaluation should beperformed using a human reliability methodology. The acceptability of thismethodology for the purpose should be justified by the licensee (Refs. 9through 13).3.5 Event Tree CollapseCollapse the event trees using a frequency screening criterion to form alist of specific sequences and a set of residual groups to be analyzed. Thisis important since the event trees may generate thousands of end states thatcannot be individually analyzed. A screening value of 1.OE-7/reactor year isrecommended. This value should ensure that important sequences are treatedindividually, and it should also help to keep the size of the residual small.This is particularly important since it may be necessary to treat the residualusing a bounding consequence condition.3.5.1 Specific SequencesThose sequences that survive the frequency screening should be defined andtheir frequency noted. It is recommended that some identification be assignedto each sequence to enhance its traceability through the remainder of the anal-ysis. Grouping and identifying each sequence with respect to initiator typemay also prove helpful.3.5.2 Residual GroupsThose sequences that do not survive the frequency screening must also beconsidered. They should be grouped together based on transient characteristicsto form a set of residual groups. The residual groups should be reviewed toidentify sequences that should be grouped with previously defined sequencesi--because of transient similarity or should be specifically evaluated because oftheir severe consequence. It is important to attempt to reduce the size of eachresidual group since it will be necessary to assign a bounding consequence thatwould apply within each group. Each residual group should be defined and its-f.requency noted.1.154-11associated wit-h ninment fi1uresTn mn.v in.stan.rcs th.e nntential, filure 4. THERMAL-HYDRAULIC ANALYSISThis chapter is to present the reactor coolant pressures, temperatures, (and heat transfer coefficients at the vessel's interior surface in the beltlineregion for the set of overcooling sequences that envelops the plant's potentialfor experiencing a PTS event. A good example is presented in Chapter 4 ofReference 3. Also the chapter is to present the details of the analysismethods used to obtain these fluid conditions and is to include the followingsections:1. The thermal-hydraulic analysis plan and logic.2. A description and evaluation of the thermal-hydraulic models.3. A description of any simplified analysis methods used in the study.4. A description of the methods used to evaluate the effects of thermalstratification and mixing.5. Graphs of all the best-estimate thermal-hydraulic results with theirassociated uncertainties and a detailed explanation of the transient behaviorobserved.4.1 Thermal-Hydraulic Analysis PlanThis section should outline the logic and identify the subtasks in thethermal-hydraulic analysis. Subtasks include detailed thermal-hydraulic systemsanalysis, simplified thermal-hydraulic systems analysis, and thermal stratifica- (tion analysis. The logic should describe the sampling plan used to selectsequences for detailed or simplified analysis. ORNL experience favors selectingdetailed thermal-hydraulic analysis sequences, including at least a few severeexamples of each type of postulated overcooling transient in order to understandand benchmark the plant behavior for subsequent simplified calculations. Theorder in which the scenarios are evaluated can result in a considerable reduc-tion in expenditures. By first analyzing the scenarios that are expected to bethe bounding cases (i.e., the most severe), calculations for an entire class ofovercooling scenarios may be deemed unnecessary if the bounding case is not ofPTS concern. Similarly, careful selection of the first set of scenarios to beevaluated can permit simple extrapolation or interpolation of the results toother scenarios that share common controlling thermal-hydraulic phenomena.During the analysis, the sequence identification analyst and the thermal-hydraulic analyst should coordinate activities to ensure that pertinent detailsof the delineated sequences are thoroughly understood. Similarly, close coor-dination must be maintained between the thermal-hydraulic analyst and the frac-ture mechanics analyst so that the transient fluid conditions are calculatedat the appropriate vessel locations.4.2 Thermal-Hydraulic ModelsThis section and supporting appendices should present a detailed descrip-tion of the thermal-hydraulic computer models used in this analysis. The models1.154-12 should include an accurate representation of the pertinent parts of the primaryand secondary systems. This includes the condensate system, the main and auxil-iary feedwater systems, and parts of the steam system. The model should includeappropriate secondary-side metal heat capacity. Particular attention should begiven to the modeling of control system logic and characteristics such as valveclosure times and liquid level measurements. References 14 through 17 illustratesome of the modeling details included in such a study. The thermal-hydraulicmodels should be capable of predicting single and two-phase flow behavior andcritical flow as required. The models should be capable of predicting plantbehavior for LOCAs, steamline breaks, and steam generator tube ruptures. Ingeneral, a one-dimensional code is suitable for most overcooling transientcalculations. However, if any of the control systems are dependent solely onthe fluid conditions in a single loop (e.g., reactor coolant pump restart crite-ria), a method of estimating the three-dimensional effects in the downcomermay be necessary for some of the asymmetric cooldown scenarios encountered inthe PTS study. Sensitivity of calculated results to the nodalization schemesused should be discussed. The thermal-hydraulic models should be coupled, whereappropriate, with neutronic models that have the capability to analyze pressuresurges resulting from any relevant sequences involving recriticality.This section of the report must also present the results of benchmarkingthe computer models against suitable plant data or data from experimentalfacilities or incorporate this information by reference to an NRC-approvedtopical report. As a minimum, the plant data comparison should fully exercisethe modeling features that are employed in the thermal-hydraulic computer pro-grams such as the pressurizer (including heaters and sprays), feedwater heatersand liquid level controls, the steam generator liquid level controls, and theturbine bypass (i.e., steam dump) controls under steady-state and transient con-ditions. If overcooling transients have occurred at the plant or at a similarplant, they should be benchmarked against the computer models. The licensee isencouraged to use codes and methods accepted by the NRC at the time the calcula-tion is performed.The models should be capable of accurately predicting condensation at allsteam-water interfaces in the primary system, especially in the pressurizerduring the repressurization phase of an overcooling event or during refillingof the primary system with cold safety-injection water. The effects of noncon-densible gases, if present, on system pressure and temperature calculationsshould be addressed.All code input and modeling assumptions should be documented and availablefor NRC review during the analysis review period (normally starting 3 yearsbefore the plant exceeds the screening limit and continuing until the evaluationresults and any requisite actions are approved by the Commission).4.3 Simplified Analysis MethodsI; This section should present the technical bases for any simplified analysismethods that are applied in the study. This includes the grouping of similarsequences by controlling phenomena and any extrapolations used to modify exist-ing calculations. If a simplified thermal-hydraulic plant model is used to pre-dict portions of the plant transients, all the simplifying assumptions inherent1.154-13m to this model should be stated and justified. Reference 18 provides examples of-how to group sequences and develop a simplified thermal-hydraulic model suitablefor portions of the analysis.4.4 Thermal Stratification EffectsTransient thermal-hydraulic computer programs available to analyze LWRresponse to overcooling scenarios do not model fluid behavior with sufficientdetail to predict the onset of HPI thermal fluid stratification ithe cold legand the subsequent cold leg and downcomer behavior. As a result, additionalanalysis methods may be needed to determine which transients are affected bythermal stratification and the extent of such effects.This section should describe and justify the thermal fluid mixing analysisX 4 methods that have been applied in the study. References 19 through 24 describethe results of recent mixing analyses and experiments. Reference 19 identifiesa useful stratification criterion to determine which overcooling transients willrequire the additional mixing analysis. Particular attention should be givento scenarios that involve HPI under very low flow or stagnant loop conditions.When stagnation is partial (i.e., not all loops stagnate), stratification isexpected only within the cold legs of the stagnant loops. However, scenariosinvolving complete loop stagnation will require the evaluation of a transientcooldown in the presence of stratified layers both in the cold legs and in aportion of the downcomer. The mixing model should include the effect of metalheating on the mixing behavior, particularly in a stagnant flow situation. Also,the effect of noncondensible gases, if present, should be included. References 1through 23 describe tools that have been used for such an analysis.This section should also document the heat transfer correlations appliedin the mixing analysis. The research efforts described in References 18 through23 indicated that the downcomer heat transfer coefficients generally exceeded300 Btu/hr-ft2-0F. These values of heat transfer coefficient were generallyhigh enough to keep the vessel wall surface temperatures within a few degreesof the downcomer fluid temperature. Furthermore, because the vessel wall cool-down was controlled by conduction processes rather than convection processes,the vessel wall surface temperatures were insensitive to heat transfer coef-ficient variations due to changes in flow and heat transfer regimes.4.5 Thermal-Hydraulic Analysis ResultsThis section should present graphs of the best-estimate downcomer pressures,fluid temperatures, and heat transfer coefficients and their associated uncer-tainty ranges as a function of time at the critical weld areas. This includesthe results of the detailed thermal-hydraulic model, the simplified model, andmixing analysis calculations.The duration assumed for each overcooling scenario should be justified. Itis assumed that a scenario duration of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> may be reasonable for many casessince the overcooling transient would probably be identified and mitigated priorto that time. However, there may be scenarios requiring lengthier evaluationperiods because the controlling phenomena delay the scenario's evolution.1.154-14 Also provide a discussion of the accuracy of the results, including ademonstration that nodalization and error estimation methods chosen are appro-priate, and how the predicted plant behavior compared to plant history and oper-ating experience. Time-dependent uncertainty estimates for the downcomer pres-sure, fluid temperature, and heat transfer coefficients at the critical weldsshould be provided for each scenario. These uncertainties are often limited byphysical phenomena. For example, the pressurizer power-operated relief valveI (PORV) setpoints will limit the system pressure for certain high-pressure sce-narios. Therefore, the uncertainty is limited by PORV operating character-istics. References 16 and 18 describe some uncertainty analysis techniques.FE;.';, .541 5. FRACTURE MECHANICS ANALYSISFor each sequence identified in Chapter 3, "Determination of Detailed PTSSequences for Analyses," calculate (or for unimportant sequences, estimate usingbounding conditions) the conditional probability of through-wall crack penetra-tion given the occurrence of the event versus fluence or RTNDT. (Althoughlicensees were required to use the method of determining RTNDT (RTp15) specifiedin paragraph 50.61(b)(2) when evaluating their vessel properties with respectto the screening limits, in performing these plant-specific calculations, theyare encouraged to use any alternative methods/data/correlations for which theyprovide justification of applicability to their specific plant.) Specificsequences identified in Section 3.5.1 should be calculated individually indetail. Less important events such as the residual groups identified in Sec-tion 3.5.2 may be conservatively bounded without a calculation for each sequencein the group. A good example is provided in Chapter 5 of Reference 3. Inputfor these calculations includes the primary system pressure, the temperatureof the coolant in the reactor vessel downcomer, the fluid-film heat transfercoefficient adjacent to the vessel wall, all as a function of time, and thevessel properties. The calculations should be performed with a probabilisticfracture mechanics code such as OCA-P or VISA-II (Refs. 25 ad 26).An acceptable procedure to be followed in the fracture mechanics analysisis as follows: A one-dimensional thermal and stress analysis for the vesselwall should be performed. The effect of cladding should be accounted for inboth the thermal and stress analyses. The fracture mechanics model can be basedon linear elastic fracture mechanics with a specified maximum value of K andKIa to account for upper-shelf behavior. Plastic instability should be consid-ered in the determination of failure. Warm prestress should not be assumed inevaluations of the postulated transients. Acceptable types of material pro-perties are given in the study of the H. B. Robinson reactor (Ref. 3).In the Monte Carlo portion of the analysis, as a minimum, each of thefollowing should be assigned distribution functions:KIC = Static crack initiation fracture toughnessKIa = Crack arrest fracture toughnessRTNDT = Nil-ductility reference temperatureCu.= Concentration of copper, wt-%Ni = Concentration of nickel, wt-%F = Fast neutron fluenceThe functions used should be justified. Examples of these distributions arefound in Reference 3.1.154-16 The following additional information should be supplied:1. Flaw density -The number of cracks per unit surface area should beestablished for use in the calculations and should be justified. A value of0.2 flaw per square meter of 8-inch-thick material (one flaw/cubic meter) wasselected in References 1, 2, and 3.2. Flaw depth density function -The flaw depth density distributionshould be established. The function to be used can be that specified inReferences 1, 2, and 3.3. Flaw size, shape, and location -Axial flaws with depths less than20 percent of the wall thickness and all circumferential flaws should be modeledin infinite length. Axial flaws with depths greater than 20 percent of the wallthickness may be modeled in infinite or finite length depending on the relativetoughness of the weld regions and plate material. For instance, the length ofan axial flaw in an axial weld that suffers severe radiation damage relative tothe plate can be limited to the length of the weld. The flaws should be assumedto be located at the inner surface of the vessel and should extend through thecladding to the inner surface of the vessel.Reference 20 provides a comprehensive discussion of recommendations forinput distributions to be used in probabilistic fracture mechanics calculations.4. All regions of the beltline should be considered.and circumferential welds as well as the base material.This includes axialThe following relationships are required:KIc f(T,RTNDTO ARTNDT), andKIa = f(T, RTNDTO,A RTNDT)whereT = Wall temperatureRTNDTo = Initial nil-ductility reference temperatureARTNDT = Increase in nil-ductility reference temperature due to radiationdamage, f(Cu,Ni,fluence). If plant surveillance data meet thecriteria for credibility given in Reference 27, they may be usedas described therein.Examples of these functions are described in References 3 and 27.In reporting the results, the methods used for the probabilistic vessel-integrity analysis should be described, their limitations for this analysisidentified, and the impact of uncertainties in the resulting vessel failureprobabilities estimated. Discussion of the analysis should include a listingof the assumptions used, their bases, and a discussion of the sensitivity ofthe results to variations in the assumptions. Vessel dimensions and materialproperties used should be given.1.154-17I;UI;pIt .i,:_

For each transient of interest, a deterministic analysis that includes aset of critical crack-depth curves as functions of time (see Refs. 1, 2, and 3),i.e., a plot of crack depths corresponding to initiation and arrest events versustime, should be carried out. This plot should also have curves indicating thedepth of crack at which upper-shelf toughness is effective. These resultsshould correspond to minus two sigma values for KIC and Ka, plus two sigmavalues for RTNDI, and plus two sigma values for the copper and nickel contentsas well as plus two sigma for the fluence value.These curves, which graphically represent the worst-case condition foreach transient of interest, will be used in the evaluation of the critical timeinterval from the initiation of the transient during which mitigating actioncan occur.IiII .iII1.154-18

• 6. INTEGRATION OF ANALYSESIn this chapter, the event frequencies are coupled with the results of thefracture mechanics analysis to obtain an integrated frequency of vessel through-wall cracking due to PTS. An example of one acceptable method is presented inChapter 6 of Reference 3. A table that supplies the following information foreach specific sequence and residual group identified in Section 3.5 should beprovided. These results are to be provided for the operating time at which thereactor will reach the PTS screening criterion and for any additional operationlife being requested:o-- °Sequence identification.ott ° Type of initiator (small-break LOCA with low decay heat,large steamline break at full power, etc.).o; ° Estimated sequence frequency.o Method used to determine conditional through-wall crack penetrationprobability.o ° Sequence conditional through-wall crack penetration probability.*° Frequency of through-wall cracking due to sequence obtained by theproduct of sequence frequency and sequence conditional through-wallcrack penetration probability.For each dominant sequence, a section or table should be provided that sup-plies (1) specific reference to the graph of temperature, pressure, and flow asprovided in Chapter 4, "Thermal-Hydraulic Analysis"; (2) a time-line descriptionof the accident sequence noting important operator actions, control actions,protection system actions, equipment faults, and vessel failure; and (3) fre-quency of through-wall crack penetration as a function of fluence or RTNDT.Results should then be summed within each initiator type to provide a fre-quency of through-wall crack penetration as a function of initiator type.The discussion should explain why each initiator type is or is not impor-tant to PTS.Finally, the results should be summed over all initiator types to providean integrated frequency of through-wall cracking for the vessel. This inte-grated value should be reported as a function of fluence, or RINOT, and plottedwith uncertainty values as determined in Chapter.7, "Sensitivity and UncertaintyAnalyses of Through-Wall Crack Frequency," and included on the plot. The dis-cussion should identify important operator actions, control actions, and plantfeatures that can cause or prevent vessel failure.'WThe conditional through-wall crack penetration probability is the probabilityof a through-wall crack as determined by the fracture mechanics analysis,given that the event occurs.1.154-19 7. SENSITIVITY AND UNCERTAINTY ANALYSES OF THROUGH-WALL CRACK FREQUENCY 6In order for the results of the probabilistic analysis to be useful forregulatory decisionmaking, the sensitivity of the results to input parametersand assumptions should be determined, the major sources of uncertainty should beidentified, and the magnitude of the uncertainty should be estimated. In thischapter, the results and the procedures used to perform each of these processesare to be documented. A good example is given in Chapter 7 of Reference 3.Portions of that analysis, or other analyses, may be referenced in lieu of por-tions of the analysis described in this chapter, provided the licensee demon-strates the applicability of the referenced analyses to the specific plant.7.1 Sensitivity AnalysisPerform a sensitivity analysis to estimate the change in the through-wallcrack frequency for a known change of a single parameter. Parameters examinedin the sensitivity analysis should include (1) the initiating event and eventtree branch frequencies, (2) the thermal-hydraulic variables (temperature, pres-sure, etc.), and (3) the fracture mechanics variables (fluence, flaw density,etc.). Where appropriate, 68th percentile (1-sigma) values should be used torepresent the change in the parameter. This should provide a sufficient changeto illustrate the effects of the change, and the use of the 68th percentilevalue whenever possible will help to define the important variabilities. Inthe case of temperature and pressure, however, the 68th percentile values mayvary from one sequence to another. In this case, it may be easier to identifya representative change in the parameter that could then be used for allsequences rather than to try to use the 68th percentile values.Each variable examined in the sensitivity analysis should be listed alongwith the change in the variable. In the cases in which changes are representedby using 68th percentile values, some explanation should be provided to documentthe reasons the value is considered a 68th percentile value. In those cases inwhich something other than a 68th percentile value is chosen, discussion shouldcenter around the reasons for choosing the value used.Sensitivity factors should be obtained by dividing the through-wall crackfrequency obtained with the changed variable by the through-wall crack frequencyobtained with each variable at its mean value. Supply the sensitivity factorsobtained for both positive and negative changes in each of the variables. Thesensitivity factors obtained for changes made in the PTS-adverse directionshould be. ranked according to magnitude and provided in table form.7.2 Uncertainty Analysis7.2.1 Parameter UncertaintiesEach step in the probabilistic analysis should include an uncertainty anal-ysis. This should include uncertainty in frequency of occurrence of a sequence,uncertainty in temperatures and pressures reached during the sequence, includingthat resulting from the nodalization scheme chosen as discussed in Section 4.5,and uncertainty in the fracture mechanics model for vessel failure given thetransients. d1.154-20 X w For the following reasons, a Monte Carlo simulation is appropriate forportions of the PTS uncertainty analysis.o The temperature and pressure error distributions are not symmetric.o The fracture mechanics results are nonlinear with respect tovariations in input parameters, particularly the temperature andpressure time histories.o The results of the Monte Carlo analysis can indicate the shape ofthe output distribution.The Monte Carlo approach would involve four steps as described below:1. Develop a statistical distribution for each variable used in thecalculation -This step will involve the representation of each variable as adistribution with 5th and 95th percentiles as previously identified. The shapesof the distributions selected should be discussed.2. Select a random value from each distribution -A random sampling codeshould be used to sample from each of the distributions.3. Calculate a through-wall crack frequency estimate based on valuesobtained in the previous step -In this step, the through-wall crack frequencyis obtained based on'the randomly selected variables. This requires under-standing the form of the relationship between each input variable and through-wall crack frequencies. For some variables such as initiating event and branchfrequencies and flaw density, this is-simple since the through-wall crackfrequency is directly proportional to the value of these parameters over therange of variable values considered. Other variables such as temperature andpressure may require the development of an appropriate relationship. In suchcases in which the effect of a variable change may be dependent on the value ofanother variable, response-surface techniques may be used to estimate importantinteraction effects.4. Summarize the resulting estimates and approximate frequency distribu-tion -Steps 2 and 3 are repeated until a statistically valid number of trialshave been performed. A distribution of through-wall crack frequencies is thenproduced from the results of the trials. The 95th and 5th percentiles and themean (expected value) of this distribution should be identified and discussed.7.2.2 Modeling Uncertainties (Biases)During the process of performing the PTS analysis, the analyst will makesimplifying assumptions in order to make the analysis tractable. Such assump-tions include decisions on thermal-hydraulic models, fracture mechanics models,grouping of sequences both for thermal-hydraulic analysis and fracture mechanicsanalysis, nodalization in the thermal-hydraulic models, etc. These assumptionscan introduce conservative or nonconservative biases into the analysis. Thesebiases should be identified and their potential impact on the results discussed.In this section, important assumptions made as part of the analysis should belisted. Each assumption should be identified as being either conservative ornonconservative. A discussion should be supplied for each assumption withrespect to its impact on the overall value of through-wall crack frequency.1.154-21 Whenever excess conservatism or nonconservatism is suspected to be present inan assumption, an alternative assumption should also be used in the full calcu-lation procedure and the impacts on the overall result compared.1.154-22 8. EFFECT OF CORRECTIVE ACTIONS ON VESSEL THROUGH-WALL CRACK FREQUENCYThis chapter is to summarize the licensee's program of corrective measures.Each corrective measure considered by the licensee should be presented and ex-plained. In each case, the reasons for considering the action as a correctivemeasure are to be documented, and the estimated impact of the action with respectto through-wall crack .frequency provided. Corrective actions that are to beconsidered include, but are not limited to, those discussed in the remainingsections of the chapter. An example can be found in Chapter 8 of Reference 3.8.1 Flux Reduction ProgramEarly analysis and implementation of such flux reductions as are reasonablypracticable to avoid reaching the screening criterion are already being requiredand accomplished in accordance with the PTS rule, § 50.61. Further flux reduc-tions to critical areas of the vessel wall that would reduce the risk of con-tinued operation beyond the screening criterion should be considered. If suchadditional flux reductions are needed, in view of the irreversibility ofembrittlement, the licensee should consider early implementation before reachingthe screening criterion. For licensees who are considering applications toextend the operating license beyond its present expiration date, it may be pru-dent to implement the reduction as early as possible to avoid the necessity ofvessel annealing or replacement.8.2 Operating Procedures and Training Program ImprovementsOperator actions and associated plant response play a key role in theinitiation and mitigation of PTS events. Therefore, ensure that the actionsare based on approved technical guidelines that include an integrated;evaluationof relevant technical considerations, including, but not limited to, PTS, corecooling, environmental releases, and containment integrity. The evaluationshould address the following types of concerns:° Frequent realistic "team" training should be conducted, exposing theoperators to potential PTS transients and their precursor events.The training should give the operators actual practice in controllingreactor system pressure and cooldown rates during PTS situations.Specific training should include, but not be limited to, reactor cool-ant pump trip criteria, the HPI throttling criterion, control ofnatural circulation, recovery from inadequate core cooling, recoveryfrom solid plant operations, and the use of PORVs to control primaryoverpressure.o Instructions should be based on analyses that include considerationof system response delay times (e.g., loop transport time, thermaltransport time).o Whether or not there is a need for cooldown rate limits for periodsshorter than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> should be evaluated.o Methods for controlling cooldown rates should be provided. Referenceshould be made to these methods with respect to the dominant PTSrisk sequences whenever possible.1.154-23 o Guidance should be provided for the operator if cooldown rates orpressure-temperature limits are exceeded. These guidelines shouldtake into account potential core cooling, environmental release, orcontainment integrity problems that could exist as a result of respond-ing to the abnormal cooldown rate. These guidelines should leavelittle doubt as to when PTS concerns are more important than othersafety issues and when other safety issues assume primary importanceover PTS concerns. It should be emphasized how the guidelines wereevaluated and optimized in light of any competing risks that mightarise from events other than PTS events to ensure that plant safetyis appropriately balanced.o The desired region of operation between the pressure-temperature limitand the limit determined by avoidance of saturation conditions shouldbe evaluated to determine if it can-be revised to minimize total riskdue to plant operation from PTS plus non-PTS events.o Instructions for controlling pressure following depressurizationtransients should be provided.o Instructions should be available for the condition where naturalcirculation is lost and the primary system main circulation pumpsare not available.Portions of the above may be provided by incorporation by reference, forexample, to the plant-specific Emergency Response Guidelines. However, a sum-mary discussion relating the referenced material to the overall subject shouldbe provided.8.3 Inservice Inspection and Nondestructive Examination ProgramThe use of state-of-the-art nondestructive examination (NDE) techniquescould provide an opportunity to decrease any conservatism that might exist inthe flaw density value used in the analysis. This decrease in conservatism,however, may be less important than the decrease in uncertainty in the actualflaw density that may result from an examination of this type.Existing inservice inspection programs should be reevaluated to considerincorporation of state-of-the-art examination techniques for inspecting theclad-base metal interface and the near-surface area. This includes plant-uniqueconsideration of the clad surface conditions. Consideration should be given toincreased frequency of inspections.The reliability of the NDE method selected to detect small flaws should bedocumented.8.4 Plant ModificationsAll plant modifications should be evaluated and optimized in light of any com-peting risks that might arise from events other than PTS events to ensure thatoverall plant safety is appropriately balanced. Plant modifications that maybe considered include the following:1. 154-24 1. Instrumentation, Controls, and Operationa. Reactor vessel downcomer water temperature monitor.b. Instantaneous and integrated reactor coolant system cooldownrate monitors.c. Steam dump interlock.d. Feedwater isolation/flow control logic.e. Reactor coolant system pressure and temperature monitors.f. Control system to prevent repressurization of the reactorprimary coolant system during overcooling events.g. Monitor to measure margin between vessel inner-surfacetemperature and current RTNDT at that location.h. Diagnostic instrumentation and displays.i. Primary coolant system pump trip logic.j. Automatic isolation of auxiliary feedwater to broken steamlines/generators.2. Increased Temperature of Emergency Core Cooling Water and EmergencyFeedwaterIf plant modifications are proposed to prevent overcooling, the reportshould include an evaluation of undesirable side effects (i.e., undercooling)and a discussion of steps planned to ensure that the modifications represent anet improvement in safety when PTS and non-PTS related events are considered.8.5 In Situ AnnealingIf in situ annealing is part of the licensee's program of corrective mea-sures, the licensee should describe the program to ensure that annealing willachieve the planned increase in vessel toughness, the surveillance program tomonitor vessel toughness after annealing, the program directed toward coderequalification after annealing, and the program to ensure that annealing doesnot introduce other safety problems.1. 154-25 9. FURTHER ANALYSESThe PTS rule (§ 50.61 of 10 CFR Part 50) requires Commission approval forplant operation with RTp15values above 270°F. This regulatdry guide outlinesthe analyses that should be performed in support of any request to operate atRTprs values in excess of 270°F, as required in paragraphs 50.61(b)(4) and50.61(b)(5), and states that the staff's primary acceptance criterion will belicensee demonstration that expected through-wall crack frequency will be below5 x 10-6 per reactor year for such operation.In the event that a licensee is unable to meet this primary acceptancecriterion, he may request Commission approval for continued operation under theprovisions of paragraph 50.61(b)(6), which allows the submittal of further anal-yses. The content of these further analyses'-would be determined by the licenseeand might include topics such as overall plant risk analyses that are beyondthe scope of the vessel failure analyses covered by this regulatory guide.,, .~~~~~~~~~~~~~~~~~~~~~~~~,'~~~~~~~~~~~~~~~~~~~~1. 154-26 10. RESULTS AND CONCLUSIONS REGARDING PTS ANALYSES.This chapter is to summarize the models used and the results obtained andprovide the conclusions reached with respect to continued operation of the plant.10.1 Summary of AnalysisIn this section the major findings of each aspect of the PTS analysis, asdescribed in the previous chapters, should be presented. These should include:o Expected (mean) value of frequency of reactor vessel through-wallcrack penetration versus time, with uncertainty bound (95thpercentile).o Identification of dominant accident sequences.o If sensitivity/uncertainty analysis shows that slightly differentassumptions could lead to different dominant sequences, identificationof these assumptions and discussion of the impact on results given thedifferent assumptions.o Identification of important operator actions, control actions, andplant features that can increase or decrease the frequency or severityof overcooling transients, and whether these have been appropriatelybalanced to ensure optimum overall plant safety.o Major sources and magnitudes of uncertainty in the analysis.o The relative effectiveness of potential alternative corrective measuresin reducing the expected (mean) value of through-wall crack penetration.° The program of planned corrective measures.10.2 Basis for Continued OperationFinally, as part of the plant-specific analysis package, the licenseeshould provide a basis for concluding whether or not continued plant operationis justified. The basis for continued operation should include comparison withNRC's PTS acceptance criteria given in the Introduction-to this guide.1.154-27 F~~~~~ -ilX ~2.14I 3.4.'l ~~7.3 ~~8.ii 9.S ~~10., '~~11.12.tt ~13.REFERENCEST. J. Burns et al., "Preliminary Development of an Integrated Approach tothe Evaluation of Pressurized Thermal Shock Risk As Applied to the OconeeUnit 1 Nuclear Power Plant," Oak Ridge National Laboratory, U.S. NuclearRegulatory Commission (USNRC) Report NUREG/CR-3770 (ORNL/TM-9176),May 1986.D. L. Selby et al., "Pressurized Thermal Shock Evaluation of the CalvertCliffs Unit Nuclear Power Plant," Oak Ridge National Laboratory, USNRCReport NUREG/CR-4022 (ORNL/TM-9408), November 1985.D. L. Selby et al., "Pressurized Thermal Shock Evaluation of the H. B.Robinson Unit 2 Nuclear Power Plant," Oak Ridge National Laboratory,USNRC Report NUREG/CR-4183 (ORNL/TM-9567), November 1985.USNRC, "Pressurized Thermal Shock (PTS)," SECY-82-465, November 23, 1982.Appendix F to-"Integrated Plant Safety Assessment Report, SystematicEvaluation Program, San Onofre Nuclear Generating Station Unit 1," USNRCReport NUREG-0829, April 1985.L. Potash, "Confusion Matrix," Section C1.2 of Appendix C in "Oconee PRA,"Electric Power Research Institute, Palo Alto, CA, and Duke Power Co.,Charlotte, NC, NSAC/60, Vol. 4, 1984.R. A. Bari et al., "Probability Safety Analysis ProceduresBrookhaven National Laboratory, Revision 1 to USNRC ReportAugust 1985.Guide,"NUREG/CR-2815,A. D. Swain and H. E. Guttmann, "Handbook of Human Reliability Analysiswith Emphasis on Nuclear Power Plant Applications," Sandia NationalLaboratories, USNRC Report NUREG/CR-1278 (SAND80-0200), October 1983.M. K. Comer et al., "Generating Human Reliability Estimates Using ExpertJudgment," General Physics Corporation, USNRC Report NUREG/CR-3688, Vols.and 2, January 1985.1D. E. Embrey, "The Use of Performance Shaping Factors and QuantifiedExpert Judgment in the Evaluation of Human Reliability: An InitialAppraisal," Brookhaven National Laboratory, USNRC Report NUREG/CR-2986(BNL-NUREG-51591), October 1983.D. A. Seaver and W. G. Stillwell, "Procedures for Using Expert JudgmentTo Estimate Human Error Probabilities in Nuclear Power Plant Operations,"Sandia National Laboratories, USNRC Report NUREG/CR-2743 (SAND82-7054),April 1983.Organisation for Economic Co-operation and Development, Nuclear EnergyAgency, Committee on the Safety of Nuclear Installations, "Assessing HumanReliability in Nuclear Power Plants," May 1983.Organisation for Economic Co-operation and Development, Nuclear EnergyAgency, Committee on the Safety of Nuclear Installations, "ExpertJudgment of Human Reliability," CSNI Report No. 88, January 1985.1.154-28E,(/

14. B. Bassett et al., "TRAC Analyses of Severe Overcooling Transients forthe Oconee 1 PWR," Los Alamos Scientific Laboratory (LASL), USNRC ReportNUREG/CR-3706, August 1985.15. C. D. Fletcher et al., "RELAP 5 Thermal-Hydraulic Analysis of PTS Sequencesfor the Oconee 1 PWR," EG&G, USNRC Report NUREG/CR-3761, July 1984.16. J. Koenig, G. Spriggs, and R. Smith, "TRAC-PFI Analyses of Potential PTSTransients at a Combustion Engine6ring PWR," LASL, USNRC ReportNUREG/CR-4109, April 1985.17. C. D. Fletcher et al., "RELAP 5 Thermal-Hydraulic Analyses of PTS Sequencesfor H. B. Robinson Unit 2 PWR," EG&G, USNRC Report NUREG/CR-3977, April 1985.18. C. D. Fletcher, C. B. Davis, and D. M. Ogden, "Thermal-Hydraulic Analysesof Overcooling Sequences for the H. B. Robinson Unit 2 PTS Study," EG&G,USNRC Report NUREG/CR-3935, July 1985.19. T. G. Theofanous et al., "Decay of Buoyancy Driven Stratified Layers withApplication to PTS," Purdue University, USNRC Report NUREG/CR-3700, May 1984.20. T. G. Theofanous et al., "REMIX: Computer Program for TemperatureTransients Due to High Pressure Injection in a Stagnant Loop," PurdueUniversity, USNRC Report NUREG/CR-3701, May 1986.21. T. G. Theofanous et al., "Buoyancy Effects on Overcooling TransientsCalculated for the USNRC Pressurized Thermal Shock Study," PurdueUniversity, USNRC Report NUREG/CR-3702, May 1986.22. Bart Daly, "Three-Dimensional Calculations of Transient Fluid-ThermalMixing in the Downcomer of the Calvert Cliffs-1 Plant Using SOLA-PTS,"LASL, USNRC Report NUREG/CR-3704, April 1984.23. Martin Torrey and Bart Daly, "SOLA-PTS: A Transient 3-D Algorithm forFluid-Thermal Mixing and Wall Heat Transfer in Complex Geometries," LASL,USNRC Report NUREG/CR-3822, July 1984.24. F. X. Dolan et a., "Facility and Test Design Report: 1/2-Scale ThermalMixing Project," USNRC Report NUREG/CR-3426, Vols. 1 and 2, September 1985.-25. R. D. Cheverton and D. G. Ball, "OCA-P, A Deterministic and ProbabilisticFracture-Mechanics Code for Application to Pressure Vessels," Oak RidgeNational Laboratory, USNRC Report NUREG/CR-3618 (ORNL-5991), July 1984.26. F. A. Simonen et al., "VISA-II -A Computer Code for Predicting the Proba-bility of Reactor Vessel Failure," Battelle Pacific Northwest Laboratories,USNRC Report NUREG/CR-4486, April 1986.27. USNRC Regulatory Guide 1.99, "Effects of Residual Elements on PredictedRadiation Damage to Reactor Vessel Materials."1.154-29 REGULATORY ANALYSIS 4The pressurized thermal shock (PTS) rule, § 50.61 of 10 CFR Part 50 qg(July 23, 1985--50 FR 29937), requires collection and reporting of materialproperties data, analyses of flux reduction options, and detailed plant-specificPTS risk analyses for those plants that reach the screening criterion based onRTNDT,* as specified in the rule, during the term of the operating license. Theregulatory guide addresses the detailed plant-specific risk analysis requirement,providing recommendations regarding how licensees should perform and how the NRCstaff should review those analyses.Neither the PTS rule nor the regulatory guide requires specific correctiveactions. The guide merely provides guidance for the performance of the analysesrequired by the rule to identify and select necessary corrective actions. There-fore, in accordance with the Commission's Regulatory Analysis Guidelines (NUREG/BR-0058, Revision 1), this regulatory analysis does not provide extensive anddetailed assessment of required, specific corrective actions.The background material, nature of the problem, objectives, and costs, etc.,of the PTS rule's requirements are covered in the regulatory analysis preparedas part of the rulemaking proceeding (Enclosure B to SECY-83-288, ProposedPressurized Thermal Shock (PTS) Rule, July 15, 1983, and Enclosure D toSECY-85-60, Final Pressurized Thermal Shock (PTS) Rule, February 20, 1985).This regulatory analysis therefore addresses only (1) the need for publishingguidance regarding how licensees should perform the required plant-specificanalyses, (2) the appropriateness of this particular guidance, and (3) the basisfor the NRC staff acceptance criteria provided in the subject guide.1. Need for GuidanceThe NRC staff has gained considerable experience concerning PTS riskanalyses. This experience has come from performance of analyses by the staff,from prototype plant-specific analyses performed by national laboratories andsponsored by NRC, and from review of industry-sponsored analyses. The regula-tory guide reflects the lessons learned from this experience and will aidlicensees in performing analyses that will efficiently derive risk estimatesin the form the NRC needs for use in evaluating their cor,formance with theregulations.This need for guidance is particularly acute since the plant-specific PTSanalyses should use a probabilistic risk analysis (PRA) approach, as opposedto the more traditional design basis accident (BA) approach, as explainedbelow.The PTS risk is developed as the sum of the small risks resulting fromeach of a large number of possible (but unlikely) PTS events. The regulatoryguide accordingly describes acceptable methods to identify as many as possibleof the potential PTS events, group them, calculate the frequencies and conse-quences of each group, determine the risk due to each group by multiplying thepredicted frequency by the calculated consequences, and then sum the results*Reference Temperature for the Nil Ductility Transition, a measure of the 4temperature range in which the materials' ductility changes most rapidly withchanges in temperature.1.154-30I!'I from 'llIgrouosito-zobtain total PTS risk estimates that can be compared withthe acceptance criteria given in the regulatory guide.-TheDBA approach, on the other hand, would attempt to define a worst cred-ibleevent (the "design basis accident") and then show that (1) consequencesfrom:that event are acceptable and (2) all other credible events are less severeand therefore acceptable. The staff has determined that this DBA approach isnot appropriate for plant-specific PTS analyses because the total risk from allcredible PTS events can be significant even though each event individually isless severe than the DBA. The NRC staff therefore believes that this guide willencourage licensees to use the acceptable PRA approach and not waste time andresources on the more traditional DBA approach.2. Justification of This Particular GuidanceV. The NRC staff has performed prototype plant-specific analyses for threeplants. They constitute the most detailed, thorough analyses performed to date,and the lessons learned in their performance are reflected in the guide. TheNRC staff has incorporated into the guide descriptions of the best methods foundregarding how to assemble details of a plant's design (and to what level thosedetails should be included), how to use event tree methodologies to identifyand group potential PTS events, how to calculate severity of the events, how tointegrate the resulting risk, and many other subjects. The staff believes thatt- the benefit of this experience is presented in this guide, and its use by i-censees will enable them to avoid many of the false starts and errors made bythe staff and their contractors in performing the prototype analyses, therebysaving time and resources.3. Justification of Acceptance CriteriaThe guide states that, in judging the acceptability of continued operationbeyond the PTS screening criterion, the staff will accept any analyses performedwith acceptable methods such as those described in the subject regulatory guidethat predict a through-wall crack penetration frequency less than 5 x 10-6 perreactor year.The mean frequency of reactor vessel through-wall crack penetration isused as the principal acceptance criterion because the staff's analyses predictthat there is a high likelihood of core damage in the event of such cracks.Core damage events have potential public health and safety consequences thatt. are difficult to analyze with certainty. They would also have severe economicimpacts upon the licensee and the public who will pay for cleanup and replace-ment power. For all these reasons, reactor vessel through-wall crack penetra-tion frequency is used as the principal acceptance criterion. The particularvalue of 5 x 10-6 mean frequency per reactor year was selected as an achievable,realistic goal that will result in an acceptable level of risk. It is believedthat this value is acceptably low considering that pressure vessel failure isnot part of the design basis of the plant and therefore must have a frequencylow enough to be considered incredible. When the various (unquantifiable)biases that are inherent in the analyses are taken into account at leastqualitatively, such as the implicit assumption that "core damage" is equivalent1.154-31 to "core melt," this value probably results in a core melt mean frequency closeto one per million reactor years.In the opinion of the NRC staff, there are no practical quantities on whichto base the acceptance criteria other than reactor vessel through-wall cracks(i.e., vessel failure).41. 154-32