NRC Generic Letter 1985-12: Difference between revisions
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IIIW~ &1985TO ALL APPLICANTS AND LICENSEES WITH WESTINGHOUSE (W) DESIGNED NUCLEAR STEAMSUPPLY SYSTEMS (NSSSs)
SUBJECT: IMPLEMENTATION OF TMI ACTION ITEM II.K.3.5, "AUTOMATIC TRIP OFREACTOR COOLANT PUMPS" (GENERIC LETTER NO. 85-12 )Gentlemen:The purpose of this letter is to inform you of (1) the staff's conclusionsregarding the Westinghouse Owners Group (WOG) submittals on reactor coolantpump trip in response to Generic Letters 83-lOc and d, and (2) provideguidance concerning implementation of the reactor coolant pump trip criteria.Our Safety Evaluation (SE) on this subject is enclosed for your use.With regard to the WOG submittals referenced in Section V of the enclosed SE,we conclude that the methods employed by the WOG to Justify manual reactorcoolant pump (RCP) trip are consistent with the guidelines and criteriaprovided in Generic Letters 83-lOc and d. The approved Westinghouse SmallBreak LOCA Evaluation Model was used to demonstrate compliance with10 CFR 50.46 and Appendix K to 10 CFR Part 50.We have determined that the information provided by the WOG in support of thealternative RCP trip criteria is acceptable on a generic basis. A suitablereactor coolant pump trip criterion can be selected by each licensee tominimize reactor coolant pump trip during steam generator tube ruptures andnon-LOCA events, while still providing for RCP trip for small break LOCAs.With regard to implementation, we note that the WOG RCP trip methodologyallows applicants/licensees to select among three alternate RCP trip criteria.The selection is based upon obtaining maximum discrimination between a smallbreak LOCA (which requires RCP trip) and a steam generator tube rupture(which does not require RCP trip). In reviewing the WOG RCP trip criteria,we note that the process of criterion selection involves a number ofconsiderations which were assigned plant-specific status by the WOG duringthe process of the trip criteria review.Accordingly, we request that operating reactor licensees select and implementan appropriate RCP trip criterion based upon the WOG methodology. Schedulesfor submittal of information requested in Section IV of the SE (refer toAppendix A for considerations associated with Generic Letters83-10c and d)should be developed with your individual project managers within 45 days from areceipt of this letter. The requested information does not constitute a n
CONTACT
- D. JaffeX28140 1-.VII I--2 ---UR 2 8 1985requirement but only identifies information specified in Generic Letters 83-10cand d which has not been provided under the WOG generic program. In the eventthat licensees decide not to trip the RCP (an option provided for in GenericLetters83-10c and d), they should respond to the questions in Section IV ofthe SE and refer to Appendix B of the SE. Applicants should provide theappropriate response to the extent that this information is known at this time.Those applicants and licensees who chooseshould submit a schedule for submittal ofJustification for non-trip of RCPs withinnot to endorse theplant specific RCP45 days of receiptWOG methodologytrip criteria orof this letter.This request for information was approved by the Office of Management andBudget under clearance number 3150-0065 which expires September 30, 1985.Comments on burden and duplication may be directed to the Office ofManagement and Budget, Reports Management, Room 3208, New Executive OfficeBuilding, Washington, D.C. 20503.If you believe further clarification regarding this issue is necessary ordesirable, please contact Mr. D. Jaffe (301 492-8140).
Sincerely,Original Signed byHugh L. Thompson, Jr.Hugh L. Thompson, Jr., DirectorDivision of Licensing
Enclosure:
SafetX Evaluationcc w/enclosure:Service ListsDistribution:DJaffeMemo FIlePKreutzerGLainasORPMsBSheronRetyped by CSchum 06/27*SEE PREVIOUS PAGE FOR CONCURRENCEDL: AD/SA*DCrutchfield06/27/85D:D oHThompsonl/4/ 185ORB#3:DL*PKreutzer06/26/85ORB#3:DL*DJaffe06/26/85C-RSB*BSheron06/26/85ORB#3:DL*EButcher06/26/85AD:OR:DL*GCLainas06/26/85
-2 -information specified in Generic Letters 83-lOc and d which has not beenprovided under the WOG generic program. In the event that licensees decidenot to trip the RCP (an option provided for in Generic Letters83-10c andd), they should respond to the questions in Section IV of the SE and refer toAppendix B of the SE. Applicants should provide the appropriate response tothe extent that this information is known at this time.Those applicants and licensees who chooseshould submit a schedule for submittal ofjustification for non-trip of RCPs withinnot to endorse the WOG methodologyplant specific RCP trip criteria or30 days.This request for information was approved by the Office of Management andBudget under clearance number 3150-0065 which expires September 30, 1985.Comments on burden and duplication may be directed to the Office ofManagement and Budget, Reports Management, Room 3208, New Executive OfficeBuilding, Washington, D.C. 20503.If you believe further clarification regarding this issue is necessary ordesirable, please contact Mr. D. Jaffe (301 492-8140).
Sincerely,Hugh L. Thompson, Jr., DirectorDivision of Licensing
Enclosure:
Safety Evaluationcc w/enclosure:Service ListsDistribution:DJaffeMemo FIlePKreutzerGLainasORPMsBSheronRetyped by CSchum 06/27*SEE PREVIOUS PAGE FOR CONCURRENCEDL i(p DC field& /11/85ORB#3:DL*PKreutzer06/26/85ORB#3:DL*DJaffe06/26/85C-RSB*BSheron06/26/85ORB#3:DL*EButcher06/26/85AD:OR:DL*GCLainas06/26/85D: DLHThompson/ /85
-2 -decide not to trip the RCP (an option provided for in Generic Letters83-10c andd), they should respond to the questions in Section IV of the SE and refer toAppendix B of the SE. Applicants should provide the appropriate response tothe extent that this information is known at this time.Those applicants and licensees who choose not to endorse the WOG methodologyshould submit a schedule for submittal of plant specific RCP trip criteria orjustification for non-trip of RCPs within 30 days.This request for information was approved by the Office of Management andBudget under clearance number 3150-0065 which expires September 30, 1985.Comments on burden and duplication may be directed to the Office ofManagement and Budget, Reports Management, Room 3208, New Executive OfficeBuilding, Washington, D.C. 20503.If you believe further clarification regarding this issue is necessary ordesirable, please contact Dr. B. Sheron (301 492-7460).
Sincerely,Hugh L. Thompson, Jr., DirectorDivision of Licensing
Enclosure:
Safety Evaluationcc w/enclosure:Service ListsDistribution:DJaffeMemo FIlePKreutzerGLainasORPMsBSheronORB#3:DL C-RSB A DL D:DLPK~rezer eJaffe BSheron E t9her nas HThompson54 /t / 85 6 /z1/85 I2X/85 / /85 UNITED STATESNUCLEAR REGULATORY COMMISSION/ WASHINGTON, D. C. 20555SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONFOR THEWESTINGHOUSE WNERS GROUPREACTOR COOLANT PUMP TRIPI. INTRODUCTIONTMI Action Plan Item II.K.3.5 of NUREG-0737 required all licensees toconsider other solutions to the small-break loss-of-coolant-accident(LOCA) problems because tripping the reactor coolant pumps (RCPs) was notconsidered the ideal solution. Automatic trip of the RCPs in the case ofa small-break LOCA was recommended until a better solution was found. Asummary of both the industry programs and the NRC programs concerning RCPtrip is provided in Generic Letters 83-lOa, b, c, d, and e, which areincluded in the NRC report, SECY-82-475, from W. J. Dircks to the NRCCommissioners, "Staff Resolution of the Reactor Coolant Trip Issue"(November 30, 1982). SECY-82-475 also provided the NRC guidelines andcriteria for the resolution of TMI Action Item II.K.3.5, "AutomaticTrip of Reactor Coolant Pumps."In SECY-82-475 the NRC concluded: "...that appropriate pump tripsetpoints can be developed by the industry that would not require RCPtrip for those transients and accidents where forced convection circulationand pressurizer pressure control is a major aid to the operators, yetwould alert the operators to trip the RCPs for those small LOCAs wherecontinued operation or delayed trip might result in core damage."SECY-82-475 also stated: "The resolution provided in the enclosures(Generic Letters 83-10 is intended to ensure that for whatever mode ofpump operation a licensee elects, a) a sound technical basis for thatdecision exists, b) the plant continues to meet the Commission's rules andregulations, and c) as a minimum, the pumps will remain running for thosenon-LOCA transients and accidents where forced convection cooling andpressurizer pressure control would enhance plant control. This wouldinclude steam generator tube ruptures (SGTR) up to approximately thedesign basis event (one tube)."The Westinghouse Owners Group (WOG) submitted two reports to the NRC inresponse to the Westinghouse specific Generic Letters, 83-lOc and d. Thefirst report provided an "Evaluation of Alternate RCP Trip Criteria"(Reference 1). The second report provided the "Justification of ManualRCP Trip for Small Break LOCA Events" (Reference 2). The WOG also pro-vided additional information (Reference 3) in response to our request forthis information, based on the review of the WOG submittals. We have alsoperformed analyses of selected events to support our review (Reference 4).
-2 -Appendix A to this report summarizesSection I of the enclosure to GenericLetter 83-10 for "Pump-Operation Criteria that Can Result in RCP TripDuring Transients and Accident," and Appendix B summarizesSection II,"Pump-Operation Criteria That Will Not Result in RCP Trip DuringTransients and Accident."II. SUMMARYThe WOG has developed a set of three alternative reactor coolant pump(RCP) trip criteria -each one being reported to be equally suitable inmeeting the intent of Generic Letter 83-10. The revised criteria replacethe current RCP trip criterion of low reactor coolant system (RCS)pressure, which could result in RCP trip for SGTR and non-LOCA events.The objective of the WOG study was to evaluate alternative RCP tripcriteria to determine if a criterion could be established to reduce theprobability of RCP trip for SGTRs and non-LOCA events, while stillproviding for RCP trip for small break LOCAs.The parameters which were considered for the evaluation of alternative RCPtrip criteria included RCS pressure, RCS subcooling, and secondarypressure dependent RCS pressure (RCS/secondary pressure differential).Because SGTRs and most non-LOCA events will not result in adversecontainment conditions, the normal instrument uncertainties associatedwith the measurement of these parameters can be utilized in evaluating theeffectiveness of these alternative criteria in preventing pump trip formost SGTRs and non-LOCAs.The alternative RCP trip criteria which were evaluated are:1. RCS pressure with normal instrument uncertainties.This criterion would be established in the same manner as the currentRCS pressure criterion, with the exception that the normal instrumentuncertainties would be utilized in determining the RCP trip setpointfor normal containment conditions. The instrument uncertaintiesassociated with post-accident containment conditions would continueto be used to determine the RCP trip setpoint for adverse containmentconditions.2. Reactor coolant subcooling.This method would provide a direct indication of the need for pumptrip, since pump trip is not required as long as the reactor coolantremains subcooled. The RCP trip setpoint would be established aszero degrees subcooling in the RCS hot legs, plus the uncertainty inthe subcooling monitor to assure that the pumps are tripped beforesubcooling is actually lost. The normal instrument uncertainties-would be used for normal containment conditions, whereas the instru-ment uncertainties associated with post-accident containmentconditions would be used for adverse containment condition . Secondary pressure dependent RCS pressure.With the current method of using RCS pressure, the trip criteria isconservatively derived assuming that the secondary pressure is at thelowest secondary safety valve set pressure. However, the secondarypressure may actually be significantly less than this value, particu-larly if the condenser steam dump system is in operation. With thismethod, the RCS pressure setpoint for pump trip would be continuouslyevaluated based on the actual secondary pressure. The alternate RCPtrip criterion can also be expressed as the RCS/secondary pressuredifferential. The combined instrument uncertainties for the RCS andsecondary pressure measurements would be included in determining theRCP trip setpoint. The normal instrument uncertainties would be usedfor normal containment conditions, whereas the instrumentuncertainties associated with post-accident containment conditionswould be used for adverse containment conditions.The results of the small break LOCA analysis demonstrate that each of thealternative criteria is adequate in providing an indication for theoperator to trip the RCPs for small break LOCAs. Thus, the selection ofthe RCP trip criterion can be based on the capability to preclude a pumptrip for SGTRs and non-LOCAs. The minimum values of each of the parametersused to evaluate the alternative criteria were also determined for SGTRsand non-LOCA events for each category of plants in the study. A methodologywas provided to determine the RCP trip setpoints for each of the threealternative criteria. Using the results of the evaluation and the setpointscalculated for each of the alternative criteria, each utility can determinewhich of the criteria will prevent pump trip for SGTRs and non-LOCA eventsfor their respective plants. iThe criterion that is considered mostappropriate in providing pump trip discrimination between LOCAs and SGTRor non-LOCA events can then be selected by the utility for each plant.Based on these studies, the WOG concluded that the RCP trip criterion cangenerally be implemented using existing qualified instrumentation alreadyavailable in the plants, and additional instrumentation is not required.The WOG followed the guidelines provided in Generic Letter 83-lOc and dto Justify manual RCP trip for small-break LOCAs. (See Appendix A,Section D.) The WOG studies concluded that:1. Every Westinghouse plant's FSAR ECCS analysis demonstrates compliancewith 10 CFR 50.46 if operator action to trip the RCPs is taken withintwo minutes after the RCP trip criterion is reached.2. Most probable best estimate analyses indicate that in allWestinghouse plants the RCPs may be tripped at any time during asmall break LOCA event without reaching clad temperatures of 2200'F.The highest PCT calculated with most probable best estimateassumptions was 12550 The WOG concluded that automatic reactor coolant pump trip is not requiredsince adequate time for manually tripping the RCPs is demonstrated using10 CFR Part 50, Appendix K assumptions as well as most probable bestestimate analysis results. It was also concluded that the most probablebest estimate analysis results demonstrate that the RCPs can be tripped atany time during the LOCA (if the operator should fail to trip the pumpswhen the trip criterion is reached) without incurring unacceptable cladtemperature results. Therefore, the WOG concludes that the existingguidelines in Revision 1 of the Emergency Response Guidelines (ERGs) aresufficient and complete with respect to RCP status for all accidentsituations, and no additional "missed RCP trip setpoint" steps arerequired.The methods (References 5 and 6) employed by the WOG to Justify manual RCPtrip are consistent with the guidelines and criteria provided in GenericLetters 83-1Oc and d.We have reviewed thk assumptions and models employed by the WOG to studysteam generator tube rupture (SGTR) and non-LOCA events. The LOFTRANcomputer program (Reference 7) was used to study these events. Bestestimate assumptions and models were used. However, the SGTR break flowmodel incorporated into LOFTRAN does result in higher than expected massflow rates for a given break size. (This model was approved for SAR SGTRanalyses, where the high flow rates result in a conservative evaluationof offsite dose.) The WOG position with respect to the use of thisconservative model is that the analysis results are bounding for thedesign basis SGTR event of a single tube.The WOG considered all other FSAR Chapter 15 non-LOCA events for evaluationagainst the alternate RCP trip criteria. It was concluded that thefeedline and steamline breaks needed to be considered because theirtransient characteristics would be the most limiting with respect to thethree criteria. For the steamline break accident, a "credible" (4.5 inchdiameter) break size was considered -essentially equivalent in size to asingle steam generator PORV failing open. For the feedline break, a fulldouble-ended rupture of the main feedwater line was considered.We have reviewed the assumptions, models and plant groupings used toperform the SGTR and non-LOCA studies and have determined that theinformation provided is acceptable. Table 1 provides a summary of the WOGstudies. We believe that the three RCP trip criteria may be marginal forsome plants for the SGTR event. We base our conclusion on the following:1. The SGTR event gives the minimal values for all three alternative RCPtrip criteria for all but a few plants, and2. The uncertainty analysis of instrument error provided by the WOG foruse to evaluate the trip set points for each alternative criteria(for both normal and adverse containment conditions) may not bebounding for all plant In particular, the reactor coolant system pressure set point RCP tripcriterion appears to offer the least in reducing unnecessary RCP trip.This confirms our position as discussed in SECY-82-475 and Generic Letter83-10.The WOG objective for the SGTR and non-LOCA analyses was to consider designbasis accidents with more realistic assumptions, to enable the developmentof a RCP trip criterion which would provide reasonable assurance ofcontinued pump operation for these accidents. While it is possible thatother accident conditions could result in more limiting parameter values,the design basis accidents which were defined for the analyses combinedwith the conservatisms which are incorporated in the analytical modelprovide assurance that the analysis results will be bounding for mostSGTR and non-LOCA events. The WOG does not consider it to be practicalor necessary to develop a RCP trip criterion which will provide forcontinued pump operation for all possible SGTR and non-LOCA events. Itwould not be a safety problem if RCP trip should occur for a SGTR ornon-LOCA event, since the plant safety systems are designed to handlethose accidents with a loss of offsite power and, therefore, with RCPtrip. The objective was to demonstrate that the RCPs will remain on formost of the expected cases of these accidents, so that the operator canretain normal pressurizer pressure control and will not be required toopen the pressurizer PORVs. In addition, maintaining forced reactorcoolant system flow will reduce the likelihood of generating voids in thereactor vessel upper head region.The WOG response to our concern that none of the three alternativecriteria would prevent RCP trip for a SGTR or a non-LOCA event (on a plantspecific bases) is the recommendation to use the criterion whichdemonstrates the greatest discrimination capability.In doing so, the WOG expects that a large range of SGTRs and non-LOCA eventsstill would not require RCP trip. In the event of RCP trip occurring forSGTRs and non-LOCAs, the WOG position is that the Emergency ResponseGuidelines (ERGs) provide specific contingency actions to recover theplant even though RCP operation is not available. Also, specific RCPrestart steps are built into the ERGs where deemed beneficial althoughthey are not required for safe plant shutdown. The WOG expects, however,that at least one of the alternative criteria will be successful inpreventing pump trip for SGTRs and non-LOCA events for each of the plants.The studies performed by the WOG to determine the transient characteristicsfor the SGTR and non-LOCA events were based on best estimate inputassumptions and models (to the extent practical with the computer programsused). Based on our experiences, with other thermal-hydraulic programsused to perform similar types of analyses, we believe there areuncertainties associated with the numerical results of any calculatedsystem transient. Each licensee must consider these uncertainties whenselecting the criterion which demonstrates the greatest discriminationcapability, and be prepared to explain how they were considered duringfuture inspection The generic nature of the WOG submittals concerning RCP trip by nature donot include any plant specific information, other than that needed todetermine plant groupings for analysis. We have therefore included asection (Implementation) in this report which describes those plantspecific items we require each licensee to address when Incorporating theRCP trip criteria into the plant procedures.III CONCLUSIONSWe have determined that the information provided by the WOG for theJustification of manual reactor coolant pump trip is acceptable. Themethods employed by the WOG to justify manual reactor coolant pump tripare consistent with the guidelines and criteria provided in Generic Letters83-10c and d. The approved Westinghouse Small Break LOCA EvaluationModel was used to demonstrate compliance with 10 CFR 50.46 and Appendix K to10 CFR Part 50.We have determined that the information provided by the WOG in support ofthe alternative reactor coolant pump trip criteria is acceptable. Asuitable reactor coolant pump trip criterion can be selected by eachlicensee to minimize reactor coolant pump trip during steam generator tuberuptures and non-LOCA events, while still providing for RCP trip for smallbreak LOCAs.The results presented by the WOG, for the plant groups studied, imply thatone of the alternative RCP criteria would prevent RCP trip for the designbasis SGTR and for design basis non-LOCA events. This would be a truestatement if the numerical results from the calculation performed wereerror free and if each plant responds exactly as the simulation modelpredicts. Also, the uncertainty analysis for instrument error would haveto be bounding for each plant, with normal containment conditions.Adverse containment conditions are not expected for design basis SGTRsor non-LOCA events.We believe the analysis tools employed by the WOG are capable ofqualitatively providing the appropriate information to evaluate thealternate RCP criteria. It should be obvious however that thequantitative values provided cannot be considered absolute. In ourjudgement, the alternate RCP trip criteria, as defined, may provideonly marginal assurance of preventing RCP trip for the design baseSGTR event for some Westinghouse plants.We have concluded that the WOG has developed acceptable criteria fortripping the reactor coolant pumps during small-break LOCAs and tominimize reactor coolant pump trip for SGTR and non-LOCA events.IV IMPLEMENTATIONThe generic information presented by the WOG does not address plantspecific concerns about instrumentation uncertainties, potential reactorcoolant pump problems and operator training and procedures as requestedin Generic Letter 83-10. Appendix A contains a summary related to theseissues and may be used as a guideline to assure that these issues areadequately addresse In order to complete the response to Generic Letters 83-lOc and d, eachW licensee is required to submit the following information to the NRCTor plant specific reviews:A. Determination of RCP Trip Criteria1. Identify the instrumentation to be used to determine the RCPtrip set point, including the degree of redundancy of eachparameter signal needed for the criterion chosen.2. Identify the instrumentation uncertainties for both normal andadverse containment conditions. Describe the basis for theselection of the adverse containment parameters. Address, asappropriate, local conditions such as fluid jets or pipe whipwhich might influence the instrumentation reliability.3. In addressing the selection of the criterion, consideration touncertainties associated with the WOG supplied analyses valuesmust be provided. These uncertainties include bothuncertainties in the computer program results and uncertaintiesresulting from plant specific features not representative of thegeneric data group.If a licensee determines that the WOG alternative criteria aremarginal for preventing unneeded RCP trip, it is recommendedthat a more discriminating plant-specific procedure bedeveloped. For example, use of the NRC-required inadequate-core-cooling instrumentation may be useful to indicate the need forRCP trip. Licensees should take credit for all equipment(instrumentation) available to the operators for which thelicensee has sufficient confidence that it will be operableduring the expected conditions.B. Potential Reactor Coolant Pump Problems1. Assure that containment isolation, including inadvertentisolation, will not cause problems if it occurs for non-LOCAtransients and accidents.a. Demonstrate that, if water services needed for RCP operationsare terminated, they can be restored fast enough once anon-LOCA situation is confirmed to prevent seal damage orfailure.b. Confirm that containment isolation with continued pumpoperation will not lead to seal or pump damage or failure.2. Identify the components required to trip the RCPs, includingrelays, power supplies and breakers. Assure that RCP trip, whendetermined to be necessary, will occur. If necessary, as aresult of the location of any critical component, include theeffects of adverse containment conditions on RCP trip reliability.Describe the basis for the adverse containment parameters selecte C. Operator Training and Procedures (RCP Trip)1. Describe the operator training program for RCP trip. Includethe general philosophy regarding the need to trip pumps versusthe desire to keep pumps running.2. Identify those procedures which include RCP trip related operations:(a) RCP trip using WOG alternate criteriaMb) RCP restartc Decay heat removal by natural circulation(d) Primary system void removal(e) Use of steam generators with and without RCPs operating(f) RCP trip for other reasons V REFERENCES1. Westinghouse Owners Group, Letter OG-110, "Evaluation of Alternate RCPTrip Criteria," December 1, 1983.2. Westinghouse Owners Group, Letter OG-117, "Justification of Manual RCPTrip for SBLOCA Events," March 9, 1984.3. Westinghouse Owners Group, Letter OG-137, "Response to NRC Question on RCPTrip," October 25, 1984.4. Lime, J. F., "TRAC Analysis of Small-Break and Tube-Rupture Accidents forthe Evaluation of Westinghouse Alternate Reactor-Coolant-Pump TripCriteria," Los Alamos National Laboratory, February 1985 (DRAFT).5. Esposito, V., et al., "WFLASH -A FORTRAN IV Computer Program forSimulation of Transients in a Multi-Loop PWR," WCAP-8200 Rev. 2, June1984.6. Bordelon, F. M., et al., "LOCTA -IV Program: Loss-of-Coolant TransientAnaylsis," WCAP-8301, June 1984.7. Burnett, T. W. T., et al., "LOFTRAN Code Description," WCAP-7907-P-A,April 1984.
Attachments:
Appendix AAppendix B TABLE ISl'flHART TABLE or voc NEACTQR COOLANT PUMP TRIP SM 9ISNO. DOCKET OR TACELANT LOOr NO. NO.POWERLEVEL(IWT)51CUTorrHEADPUMPWEIRtI/ACC.C1A"I pIESTiN (PSIUCUPHEADTEOPRICS.C. S.C. NO MINIMUM ICS MINIMUM RtS MIN RCSISECONDARYVALVE TUBE LOAD PRESSURE SUSCOOLINE Dair. PRESSURESET PT I.D. TATE ---- (PSIA) ---- ----- (F) ---- ---- (PSI)--(PSIA) (IN) (F) SCTR SLI ILU lSUR SLR rLe SCm SLI fLBVOGTLE IVOGTLE 2SEABROOK ISEABROOK IMILLSTONE 3WOLF CREEKCALLAVlABYRON IBYRON IBRAIDVOOD IIRAIDWOOD IMc GUIRE InC tUIRE 2CATAVIA ICATAV0A 2MARBLE HILL IMARBLE RILL 2WATTS BAR IWATTS BAR 2COMANCHE PEAK 1COMANCHE PEAK I4 50-4244 50-4254 50-4434 50-4444 50-4234 50-4824 50-4830 34110 34110 34110 341t0 34110 34110 3411Hi YHI iRl iH iTHliTN 400N I40N 405N t00N t00N 40lN to04 50-4544 56-4554 50-4564 56-4574 50-3414 50-3704 50-4134 55-4144 50-5464 50-5474 .50-3904 51-3914 50-4454 50-444000049 is300006000034113411341134113411341134113411341134113411341134113411HI IHI YNl TR iTH iTHI THI 7HR rH iTHi THi 7HI THi TNI rIIIIIIIN 400N 400N 400N t00400401400405N tooN t00460405400N 4oC 12O0C 120C 1206C 1200C 1200C 1200C 1200C 1200C 1200C 1205C 1205C 10OEC 1200C 1200C 1200C 1200C 1200C 1200C 1200C 1206C 1200.406 557 1753 1774 1164.444 557 146 1774 141856 71 4t 415 811 81757 71 4? 44 Sit 817(TROJAN4 50-344 47477 3301 Hi V t N 400 H 1100.775 557 1524 1774 1t4e52 71 4? 535 111 817(ZION IZION 24 50-213 441113 3250 NI b N N 400 N 1100 .775 547 1407 1774 1141 ll 71 49 404 ill I174 50-304 47614 3250 HI N N t00 H 1101DIABLO CANTON 2SALEM ISALEM 2SEOUOTAH ISEGUOYAN 24 50-3234 50-2724 50-3114 50-3274 50-3210 3411 NI4I174 3411 HI40470 3411 Ni47944 3411 HI51321 3411 IllIIIN 400 R 1100 .775 547 1473 1774 13IJ 57 71 0 5614 111 117N d00 H 1106N 466 R 1100N T 406 c 119lN T 400 C 1100 TABLE I (CON'T)ASUMART TABLE Of VOD REACTOR COOLANT PUMP TRIP STUDIES51ACC. UP S.C. S.C. NOMINIMHU RCSPOVER CUT FUM? CAS HEAD VALV TUBE LOAD FRESSURENO. DOCKET OR TAC LEVEL OFF VEIR URI PRES TEt? SET PT I.D. TATE ---- (PSIA) ----MINIMNU RCS 111t RCSISECONDARYSUICOOLIN; DIET. PRESSURE.. ()- iS I-- ---- 1P5 --SGTR SLI fIl SGTR SLI FlUPLANTLOOP No. NO. (IYT) READtiN TIN (FSI)CHIC (PSIA) (IN) (fl SCUT SLI 1l3D C. COOK IB.C. COOK IDIABLO CANYON ISOUTH TE2AS ISOUTH TEIAS 2INOIAN POINT 2INDIAN POINT 3SUMMERSHERON HARRIS ISHERON HARRIS 2FARLEY IFARLEY 2NORTH ANNA INORTH ANNA 2SURR7 ISURRY 2DEAVER VALLEY IBEAVER VALLEY 2H.R. ROIINSONTURKEY POINT 3TURKEY POINT 44 50-315 49701 3250 Hi4 50-311 49702 3391 Hi4 50-275 41454 3335 HiN N I00 H 1100 .775 547 1443 1774 1011N N I05 H 1100Y N 400 R 11004 5.-4954 50-4990 3500 IN0 3800 IN4 50-247 41472 2758 LO4 50-2ll 41457 3025 LO3 50-375 41115 2775 Hi3 50-400 0 2775 Hi3 50-401 0 2775 Hi3 50-348 41145 2152 NI3 50-314 4141V 2452 HI3 50-331 41445 2775 Hi3 50-337 40144 2775 Hi3 50-250 41481 2441 HI3 50.281 41412 2441 Hi3 50-334 49415 2452 HIN 400 H 1300 .444 517 1407 1530 1551N 400 H 1300N 45t H 1100 .775 547 1110 1155 1157N t00 H 1100 .775 547 1211 1155 1557N 400 C 1200 .444 557 1543 1431 1123N 100 C 1200N 605 C 1206N 501 R 1100 .775 547 1234 1436 1723N 400 H 1100N 400 R 1100N 4l6 H 1100N N 400 R 1100N N t0 N 1100N 400 U floeN 101 R 1100 .775 547 1147 1434 1123N !1P a 1100 .775 547 1147 1741 I'IMN 40a R 1101N 410 N 110055 71 41 542 111 11740 44 I2 453 414 55731 10 1 2193 1050 15432 90 10 315 1050 154SI 70 71 541 711 52237 70 71 350 710 12231 70 71 271 710 52231 52 10 301 1150 102(3 50-4120 2l5t Hi(3 50-241 404l0 2321 LO ' T3 30-250 49479 200 10 L3 50-251 41460 2200 LO TPRAIRIE ISLAND IPRAIRIE ISLAND I2 50-i22 41443 1450 IN2 50-304 41444 1450 INN N 700 aN N 711 R1100 .775 547 1341 1117 1751110031 85 It 351 1100 1047I 50-30S 40411 14S0 INN N 700 R 1100 .775 547 1253 1777 175131 15 14 341 110t 1147R.E. CINMAPOINT BEACH IPOINT BEACH 22 50-144 4101 1310 LO T2 S0-244 40454 1520 O II 50-301 49457 1520 O IN 700 R 1100 .775 II? 11t7 1111 I123N 100 H 1100N 700 R 110021 74 1 1305 440 t00 APPENDIX APUMP-OPERATION CRITERIA THAT CAN RESULT IN RCP TRIPDURING TRANSIENTS AND ACCIDENTSA. The NRC staff has concluded that if sufficient time exists, manualaction is acceptable for tripping the RCPs following a LOCA providedcertain conditions are satisfied.B. Potential problem areas should be considered in developing RCP-tripsetpoints and methods.1. Tripping RCPs causes loss of pressurizer sprays.a. This produces a need to use PORVs in some plants to controlprimary pressure.b. PORVs have frequently failed to close.c. Despite testing, PORV operational reliability has not improvedsignificantly.2. Tripping RCPs tends to produce a stagnant region of hot coolant inthe reactor-vessel upper elevations.a. Hot stagnant coolant has flashed and partially voided the uppervessel region during depressurization or cooldown operationalevents.b. Operators are not completely familiar with the significance ofan upper-head steam bubble.c. Operators have difficulty controlling coolant conditions toavoid or control flashing.d. Operators may take precipitous actions when a steam bubbleexists.3. After tripping the RCPs, decay-heat removal by natural circulation isrequired. This procedure is used less frequently than controllingwith the RCPs and it places more demand on the operators to controlthe primary-system conditions.C. Consider the following guidelines in developing RCP-trip setpoints.1. Demonstrate and justify that proposed RCP-trip setpoints are adequatefor small-break LOCAs but will not cause RCP trip for other non-LOCAtransients and accidents such as SGTRs.-
Ia. Assure that RCP trip will occur for all primary-coolant lossesin which RCP trip is considered necessary.b. Assure that RCP trip will not occur for SGTRs up to and includ-ing the design-basis SGTR.c. Assure that RCP trip will not occur for other non-LOCA tran-sients where it is not considered necessary.d. Perform safety analyses to prove that a, b, and c above areachieved.e. Consider using partial or staggered RCP-trip schemes.f. Assure that training and procedures provide direction for use ofindividual steam generators with and without operating RCPs.g. Assure that symptoms and signals differentiate between LOCAs andother transients.h. (Westinghouse plant specific) RCP trip is expected to occur forthe design-basis SGTR for some Westinghouse plants that have SIpumps with lower shutoff heads. The exact rupture size abovewhich RCP trip would be required has not been determined.(1) NRC informed Westinghouse that RCPs should not be trippedfor SGTR events such as that which occurred at Ginna,which was essentially equivalent to a design-basis SGTR.(2) NRC informed Westinghouse that methods should be examinedfor either improving the RCP-trip setpoints or modifyingthe plants so that RCPs need not be tripped fordesign-basis SGTRs.(3) Restart permission was granted for Ginna with therequirement that supplementary guidelines be developed forRCP trip to assure RCPs would not be tripped for thedesign-basis SGTR.(4) NRC agrees with Westinghouse that in the long term usingthe reactor-vessel-liquid-inventory system to helpdetermine when to do a RCP trip will increase theprobability of maintaining RCP operation during non-LOCAs.i. (Westinghouse plant specific) NRC has concluded that recentinformation by Westinghouse about wide-range pressureuncertainty indicates that analyses confirming Westinghouse'sconclusions about RCP trip setpoints for high-head-SI plants areprobably necessar . Exclude extended RCP operation in a voided system where pump head ismore than 10% degraded unless analyses or tests can justify pump andpump-seal integrity when operating in voided systems.3. Avoid challenges to the PORVs where possible.a. If setpoints lead to RCP trip even though it is neither requirednor desirable for transients or accidents with offsite poweravailable, assure that challenges to the PORVs are avoided thatwould normally be handled by using pressurizer sprays.b. Challenges to PORVs could be eliminated by using heatedauxiliary pressurizer sprays from a source other than the RCPdischarge.c. If submittal recommends use of PORVs to depressurize, thenlicensees need to develop a program for upgrading the PORVs'operational reliability.4. Establish guidelines and procedures for cases where RCP trip can leadto hot, stagnant fluid regions at primary-system high points.a. Describe symptoms of primary-system voiding caused by flashingof hot, stagnant fluid regions including effects on thepressurizer.b. Specify guidance for detecting, managing and removing the voids.c. Train operators concerning the significance of primary-systemvoids for both non-LOCA and LOCA conditions.5. Assure that containment isolation will not cause problems if itoccurs for non-LOCA transients and accidents.a. Demonstrate that, if water services needed for RCP operation areterminated, they can be restored fast enough once a non-LOCAsituation is confirmed to prevent seal damage or failure.b. Confirm that containment isolation with continued pump operationwill not lead to seal or pump damage or failure.6. RCP-trip decision parameters should provide unambiguous indicatorsthat a LOCA has occurred and the NRC-required inadequate-core-coolinginstrumentation should be used where useful in indicating the needfor a RCP trip.7. NRC recommends that the licensee use event trees to systematicallyevaluate their set points to minimize the potential for undesirableconsequences because of a misdiagnosed even a. Evaluate set points for events with RCP trip when it ispreferable the RCPs remain operational.b. Evaluate set points for events where early RCP trip does notoccur and a delayed trip may lead to undesirable consequences.D. NRC's guidance for justification of manual RCP trip in the-licenseesubmittals is summarized in this section. This guidance had two purposes.It was intended to assist plants that can and should rely on manual tripto justify it, and it was also intended to help identify those few plantsthat may not be able to rely on manual trip.1. Analyses should demonstrate that the limits set forth in 10 CFR 50.46are not exceeded for the limiting small-break size and location usingthe RCP-trip set points developed with the guidance of part C above.a. Assume manual RCP trip does not occur earlier than 2 minutesafter the RCP-trip set point is reached.b. Include allowances for instrument error.c. Generic analyses are acceptable if they are shown to bound theplant-specific evaluations.2. Determine the time available to the operator to trip the RCPs for thelimiting cases if manual RCP trip is proposed.a. Perform the analysis for the limiting small-break size andlocation identified in D.1 above.b. Use the most probable best-estimate analysis to determine thetime available to trip the RCPs following the time when theRCP-trip signal occurs.c. Most probable plant conditions should be identified and justi-fied by each licensee.d. NRC will accept conservative estimates in the absence of justi-fiable most probable plant conditions.e. Justify that the time available to trip the RCPs is acceptableif it is less than the Draft ANSI Standard N660.(1) Include an evaluation of operating experience data.(2) Address the consequences if RCP trip is delayed beyond thistime.(3) Develop contingency procedures and make them available forthe operator to use in case the RCPs are not tripped in thepreferred time fram (4) No justification is required if the time available to tripthe RCPs exceeds the Draft ANSI Standard N660.E. Assure that good engineering practice has been used for the followingareas.1. Establish the quality level for the instrumentation that will signalthe need for RCP trip.a. Identify the basis for the sensing-instruments' design featureschosen.b. Identify the basis for the sensing-instruments' degree ofredundancy.c. Licensees can take credit for all equipment available to theoperators that they have sufficient confidence in its operabilityduring the expected conditions.2. Ensure that emergency operating procedures exist for the timelyrestart of the RCPs when conditions warrant.3. Instruct operators in their responsibility for tripping RCPs forsmall-break LOCAs including priorities for actions after theengineered safety features actuation occur APPENDIX BPUMP-OPERATION CRITERIA THAT WILL NOT RESULT IN RCP TRIPDURING TRANSIENTS AND ACCIDENTSConsider the following guidelines if the submittal concludes that keeping theRCPs running is both the preferred and safest method of pump operation forsmall-break LOCAs and other transients and accidents.A. Evaluate inventory loss.1. Complete evaluation of LOFT Test L3-6 through the ECCS recoveryphase.2. Evaluate all modeling differences expected between LOFT and aPWR analysis.B. Evaluate pump integrity.1. Justify how the pump-seal and pump structural integrity will beassured during extended two-phase flow performance.2. Include the consequences of pump and/or pump-seal failure in theanalyses if their integrity cannot be assured.3. Include one of the following if continuous RCP operation isexpected even with a containment isolation signal.a. Evaluate the capability to continue RCP operation withoutessential water services.b. Evaluate the capability to rapidly restore essentialwater services.4. Evaluate the RCP's capability to operate in the accidentenvironment.5. Evaluate the consequences of RCP failure at any time during theaccident if continuous operation in the accident environmentcannot be assured.C. Ensure acceptability of results.1. Analyses should demonstrate that the 10 CFR 50.46 ECCS acceptancecriteria are met with a model in compliance with Appendix K to10 CFR Part 50.2. Assume continuous pump operation and also RCP trip at varioustimes if continuous pump operation cannot be assured.3. NRC will consider a request for an exemption to 10 CFR 50.46requirements if analyses indicate compliance cannot be achieved.a. Submittal concludes that compliance with 10 CFR 50.46 wouldrequire operating the plant in a less safe condition. This
-2-needs to be supported with a risk/benefit analysis that cantake credit for all equipment expected to remainoperational during the accident.b. Submittal concludes that design modifications would not becost-effective to implement from a safety standpoint.