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| issue date = 07/23/1990
| issue date = 07/23/1990
| title = Provides Results of Offsite Dose Calculation for Reactor Coolant Pump Locked Rotor Event for Facility Cycle 8.Util Identified Previously Issued SERs Addressing Short Term Containment Analysis & LOCA Containment Integrity
| title = Provides Results of Offsite Dose Calculation for Reactor Coolant Pump Locked Rotor Event for Facility Cycle 8.Util Identified Previously Issued SERs Addressing Short Term Containment Analysis & LOCA Containment Integrity
| author name = ALEXICH M P
| author name = Alexich M
| author affiliation = INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
| author affiliation = INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
| addressee name = MURLEY T E
| addressee name = Murley T
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000316
| docket = 05000316
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=Text=
=Text=
{{#Wiki_filter:ACCELERATEDDIRIBUTIONDEMONSTTIONSYSTEMREGULATORYINFORMATIONDISTRIBUTIONSYSTEM(RXDS)ACCESSIONNBR:9007260232DOC.DATE:90/07/23NOTARIZED:NODOCKETFACIL:50-316DonaldC.CookNuclearPowerPlant,Unit2,Indiana605000316AUTH.NAMEAUTHORAFFILIATIONALEXXCH,M.P.XndianaMichiganPowerCo.(formerlyIndiana6MichiganEleRECIP.NAMERECIPXENTAFFILIATXONMURLEY,T.E.DocumentControlBranch(DocumentControlDesk)
{{#Wiki_filter:ACCELERATED DI                  RIBUTION DEMONST                TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RXDS)
ACCESSION NBR:9007260232              DOC.DATE: 90/07/23    NOTARIZED: NO          DOCKET FACIL:50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana 6                05000316 AUTH. NAME          AUTHOR AFFILIATION ALEXXCH,M.P.         Xndiana Michigan Power Co. (formerly Indiana            6 Michigan Ele RECIP.NAME            RECIPXENT AFFILIATXON MURLEY,T.E.             Document      Control Branch (Document Control Desk)


==SUBJECT:==
==SUBJECT:==
ProvidesresultsofoffsitedosecalculationforreactorcoolantpumplockedrotoreventforfacilityCycle8.DISTRIBUTIONCODE:A009DCOPIESRECEIVED:LTRENCLSIZE:TITLE:OR/LicensingSubmittal:AppendixINOTESRECIPIENTIDCODE/NAMEPD3-1LAGIITTER,J.COPIESLTTRENCL1011RECIPIENTIDCODE/NAMEPD3-1PDCOPIESLTTRENCL55INTERNALNRR/BRADFUTEiJ11NRR/DREP/PRPBll11OC/LFM10G11NRR/DREPDIR10ENUDOCS-ABSTRACTOGC/HDS1RGN3DRSS/RPB10111011EXTERNALEGGGAKERSiDNRCPDR1111LPDR11NOTETOALL"RIDS"RECIPIENTS:PLEASEHELPUSTOREDUCEWASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATEYOURNAMEFROMDISTRIBUTIONLISTSFORDOCUMENTSYOUDON'TNEED!TOTALNUMBEROFCOPIESREQUIRED:LTTR18ENCL14 indianaMichiganPowerCompanyP.O.Box16631Columbus,OH43216AEP:NRC:1071KDonaldC.CookNuclearPlantUnit2docketNo.50-316License.No.DPR-74OFFSITEDOSECALCULATIONFORTHEREACTORCOOLANTPUMPLOCKEDROTOREVENTFORUNIT2CYCLE8U.S.NuclearRegulatoryCommissionAttn:DocumentControlDeskWashington,D.C.20555Attn:T.E.Murley.July23,l990
Provides results of          offsite dose calculation for reactor coolant pump locked          rotor event for facility Cycle 8.
DISTRIBUTION CODE: A009D          COPIES RECEIVED:LTR        ENCL      SIZE:
TITLE: OR/Licensing Submittal: Appendix            I NOTES RECIPIENT              COPIES          RECIPIENT          COPIES ID  CODE/NAME            LTTR ENCL      ID  CODE/NAME      LTTR ENCL PD3-1 LA                    1    0    PD3-1 PD                5    5 GIITTER,J.                   1    1 INTERNAL  NRR/BRADFUTE i J            1    1    NRR/DREP DIR10E        1    0 NRR/DREP/PRPBll              1    1    NUDOCS-ABSTRACT        1    1 OC/LFM                      1    0    OGC/HDS1                1    0 G                          1    1    RGN3    DRSS/RPB        1    1 EXTERNAL  EGGG AKERS i D              1    1    LPDR                    1    1 NRC PDR                      1    1 NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED: LTTR              18  ENCL    14


==DearDr.Murley:==
indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 AEP:NRC:1071K Donald C. Cook Nuclear Plant Unit        2 docket No. 50-316 License. No. DPR-74 OFFSITE DOSE CALCULATION FOR THE REACTOR COOLANT PUMP LOCKED ROTOR EVENT FOR UNIT 2 CYCLE 8 U. S. Nuclear Regulatory Commission Attn:      Document Control Desk Washington, D. C. 20555 Attn: T.      E. Murley .
ThepurposeofthisletteristotransmittheresultsoftheoffsitedosecalculationforthereactorcoolantpumplockedrotoreventforUnit2Cycle8.ThelicensebasisanalysisoflockedrotoreventincludedinAppendixB,SectionB.3.5.2ofAttachment4ofoursubmittalAEP:NRC:1071EdoesnotassumefuelfailurebasedonDNB.However,Condition9oftheSERforWCAP10444-P-A,"ReferenceCoreReport,Vantage5FuelAssembly,"states"WithregardtotheRCSpumpshaftseizureaccident,thefuelfailurecriterionshouldbethe95/95DNBRlimit.ThemechanisticmethodmentionedinWCAP-20444isnotacceptable."Suchananalysisshowing11$rodsinfailurewassubmittedasAttachment3tooursupplementarysubmittalAEP:NRC:1071I.Ourfuelcontractor,WestinghouseElectricCorporationhasalsocalculatedtheexpectedoffsitedoseresultingfromthefailureof11$ofthefuelrods.Attherequestofthestaff,wearesubmittingadiscussionofthiscalculationasanattachmenttothisletter.ApreliminarycopyofthisdiscussionwastelecopiedtoTimothyColburn,CookNuclearPlantProjectManager,onJuly18,1990.InadiscussionwiththestaffonJuly16,1990,concernwasexpressedregardingourreferencestothe"Shore-termContainmentAnalysis"andthe"LOCAContainmentIntegrity"onpage34ofAttachment4tooursubmittalAEP:NRC:1071E.Fortheconvenienceofthestaff,weidentifiedpreviouslyissuedSERsaddressingtheseanalyses.Thefirstofthesewasattachedtoaletterdated"9007260232900723PDRADOCK050003it~P'DCgdofI/I Dr.T.E.Murley-2-AEP:NRC:1071KJanuary30,1989fromJohnF.StangoftheNRCstafftoMr.MiltonP.Alexich,VicePresidentofIndianaMichiganPowerCompany,regardinga"ChangetoLicensingBasisforSIandRHRCrossties."ThisSERapprovestheLOCAcontainmentintegrityanalysisforbothunits.ThesecondSERwasattachedtoAmendmentNo.126toFacilityOperatingLicenseNo.DPR-58,CookNuclearPlantUnitl.Thestaffconcluded"thattheD.C.CookNuclearPlant'sdesignbasispertainingtocontainmentshorttermresponse...isadequateforRTPoperationandthereforeisacceptable."TheSERspecifiesUnit1initsdiscussion.However,asindicatedinWCAP11902,(Attachment2tooursubmittalAEP:NRC:1067),onpg3.4-2,"TheanalysisaddressedtheCookNuclearPlants[sic]Units1and2,atreratedconditionsassuminganNSSSpowerlevelof3600MWt,forarangeofconditionswhichbound[thereducedtemperatureandpressureconditions)...."TheanalysisalsoboundstheoperatingconditionsproposedforUnit2Cycle8.TheassumptionsalsoboundtheconditionsdocumentedinAttachment4toAEP:NRC:1071EforUnit2.Wealsonotethatasfarastheyimpactthisanalysis,theUnit1andUnit2containmentsareidentical.OneoftheseSERsandpertinentportionsoftheotherweretelecopiedtothestaffonJuly18,1990fortheconvenienceofthereviewer.ThisdocumenthasbeenpreparedfollowingCorporateproceduresthatincorporateareasonablesetofcontrolstoensureitsaccuracyandcompletenesspriortosignaturebytheundersigned.Sincerely,M.P.AlexichVicePresidentMPA/ehcc:D.H.Williams,Jr.A.A.Blind-BridgmanJ.R.PadgettG.CharnoffNFEMSectionChiefA.B.Davis-RegionIIINRCResidentInspector-Bridgman ATTACHMENTTOAEP:NRC:1071KLOCKEDROTORDOSEANALYSISFORDONALDC.COOKNUCLEARPLANTUNIT2CYCLES8AND9
July 23, l990
~'eactorCoolantPumockedRotorOffsiteRadiationOoseAnalsisAconservativeanalysisofthepotentialradiologicalconsequencesofareactorcoolantpumplockedrotoreventhasbeenperformed.Thesalientassumptionsusedtocalculatetheactivityreleasesandoffsitedosesfollow.1.Priortotheaccident,theprimarycoolantiodineconcentrationisassumedtoequaltheTechnicalSpecificationtransientlimitforfullpoweroperation-60uCi/gramofdoseequivalentI-131(TechnicalSpecificationFigure3.4-1)2.Priortotheaccident,thesecondarycoolantiodineconcentrationisassumedtoequaltheTechnicalSpecificationlimit(LCO3.7.1.4)forfullpoweroperation-0.1uCi/gramofdoseequivalentI-1313.ElevenpercentofthecoreispredictedtofailasaresultofDNB.Thisresultsinthereleaseof11%ofthecoregapactivitytotheprimarycoolant.Thefractionofcoreactivitycontainedinthegap(gapfraction)isassumedtobe10%forallnuclides.Thus,atotalof1.1%'fthecoreactivityisreleased.4.Thetotalprimary-to-secondaryleakrateisassumedtobeattheTechnicalSpecificationlimit(LCO3.4.6.2itemc)of1gpmforthedurationoftheevent.5.Aniodinepartitioncoefficientof100betweenthesteamgeneratorliquidandsteamphasesisassumed.ThisvalueissuggestedinStandardReviewPlan(NUREG-0800)Section15.6.3,Revision2.Itisaconservativeestimateofthepartitioncoefficientduetosecondarycoolantboil-off.6.Offsitepowerislost.7.Steamreleasetotheenvironment:0to2hr-600,000ibm2to8hr-1,200,000ibm8.EighthoursaftertheaccidenttheResidualHeatRemovalSystemisassumedtostart.Noadditionalsteamorradioactivityisreleasedtotheenvironmentafter8hrs.Reactorcoreandcoolantfissionproductinventories,doseconversionfactors,gammaenergies,atmosphericdispersionfactors,andbreathingratesareconsistentwiththevaluesusedinWCAP-12135.ResultsDoseinremSiteboundaryThyroid3.0Whole-bodygamma0.3Lowpopulationzone6.00.2 ConclusionThedoseacceptancecriteriaisbasedontherecommendationsofStandardReviewPlanSection15.5.3,i.e.,30remthyroidand2.5remwhole-body.Thecalculateddosesforthelockedrotoreventarewithintheacceptancecriteria.}}
 
==Dear Dr. Murley:==
 
The purpose      of this letter is to transmit the results of the offsite dose      calculation for the reactor coolant pump locked rotor event for Unit 2 Cycle 8.
The    license basis analysis of locked rotor event included in Appendix B, Section B.3.5.2 of Attachment 4 of our submittal AEP:NRC:1071E does not assume fuel failure based on DNB. However, Condition 9 of the SER for WCAP 10444-P-A, "Reference Core Report, Vantage 5 Fuel Assembly," states "With regard to the RCS pump shaft seizure accident, the fuel failure criterion should be the 95/95    DNBR limit. The mechanistic method mentioned in WCAP-20444 is  not    acceptable." Such an analysis showing 11$ rods in failure was submitted as Attachment 3 to our supplementary submittal AEP:NRC:1071I. Our fuel contractor, Westinghouse Electric Corporation has also calculated the expected offsite dose resulting from the failure of 11$ of the fuel rods. At the request of the staff, we are submitting a discussion of this calculation as an attachment to this letter. A preliminary copy of this discussion was telecopied to Timothy Colburn, Cook Nuclear Plant Project Manager, on July 18, 1990.
In  a  discussion with the staff on July 16, 1990, concern was expressed      regarding our references to the "Shore-term Containment Analysis" and the "LOCA Containment Integrity" on page 34 of Attachment 4 to our submittal AEP:NRC:1071E. For the convenience of the staff, we identified previously issued SERs addressing these analyses.       The first of these was attached to a letter dated "9007260232 900723 PDR P         'DC ADOCK 050003it~                                                                gdof I/I
 
Dr. T. E. Murley                                   AEP:NRC:1071K January 30, 1989 from John F. Stang of the NRC staff to Mr. Milton P. Alexich, Vice President of Indiana Michigan Power Company, regarding a "Change to Licensing Basis for SI and RHR Crossties."
This SER approves the LOCA containment integrity analysis for both units. The second SER was attached to Amendment No. 126 to Facility Operating License No. DPR-58, Cook Nuclear Plant Unit l.
The staff concluded "that the D. C. Cook Nuclear Plant's design basis pertaining to containment short term response... is adequate for  RTP  operation and therefore is acceptable." The SER specifies Unit  1  in its discussion. However, as indicated in WCAP 11902, (Attachment 2 to our submittal AEP:NRC:1067), on pg 3.4-2, "The analysis addressed the Cook Nuclear Plants [sic] Units 1 and 2, at rerated conditions assuming an NSSS power level of 3600 MWt, for a range of conditions which bound [the reduced temperature and pressure conditions) . . . ." The analysis also bounds the operating conditions proposed for Unit 2 Cycle 8. The assumptions also bound the conditions documented in Attachment 4 to AEP:NRC:1071E for Unit 2. We also note that as far as they impact this analysis, the Unit 1 and Unit 2 containments are identical.
One of these SERs and pertinent portions of the other were telecopied to the staff on July 18, 1990 for the convenience of the reviewer.
This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, M. P. Alexich Vice President MPA/eh cc:   D. H. Williams, Jr.
A. A. Blind - Bridgman J. R. Padgett G. Charnoff NFEM Section Chief A. B. Davis - Region  III NRC  Resident Inspector - Bridgman
 
ATTACHMENT TO AEP:NRC:1071K LOCKED ROTOR DOSE ANALYSIS FOR DONALD C. COOK NUCLEAR PLANT UNIT  2 CYCLES 8 AND 9
 
~
  'eactor Coolant    Pum    ocked Rotor  Offsite Radiation  Oose Anal    sis A    conservative analysis of the potential radiological consequences            of a reactor coolant pump locked rotor event has been performed.
The    salient assumptions      used  to calculate the  activity releases    and  offsite doses follow.
: 1. Prior to the accident, the primary coolant iodine concentration is assumed    to equal the Technical Specification transient limit for full power operation - 60 uCi/gram of dose equivalent I-131 (Technical Specification Figure 3.4-1)
: 2. Prior to the accident, the secondary coolant iodine concentration is assumed to equal the Technical Specification limit (LCO 3.7. 1.4) for            full power operation - 0.1 uCi/gram of dose equivalent I-131
: 3. Eleven percent of the core is predicted to fail as a result of DNB. This results in the release of 11% of the core gap activity to the primary
    'fcoolant. The fraction of core activity contained in the gap (gap fraction) is assumed to be 10% for all nuclides. Thus, a total of 1. 1%
the core activity is released.
: 4. The  total primary-to-secondary leak rate is      assumed  to  be at the Technical Specification      limit (LCO 3.4.6.2 item c) of    1  gpm  for the duration of the event.
: 5. An  iodine partition coefficient of 100 between the steam generator liquid and steam phases      is assumed. This value is suggested in Standard Review Plan (NUREG-0800)     Section 15.6.3, Revision 2. It is a conservative estimate of the partition coefficient due to secondary coolant boil-off.
: 6. Offsite    power  is lost  .
: 7. Steam  release to the environment:     0  to 2 hr - 600,000 ibm 2  to 8 hr - 1,200,000 ibm
: 8. Eight hours after the accident the Residual Heat Removal System is assumed  to start. No additional steam or radioactivity is released to the environment after 8 hrs.
Reactor core and coolant        fission product inventories, dose conversion factors, gamma energies, atmospheric dispersion factors, and breathing rates are consistent with the values used in WCAP-12135.
Results Dose  in rem Site boundary        Low population zone Thyroid                  3.0                  6.0 Whole-body gamma        0.3                  0.2
 
Conclusion The dose acceptance criteria is based on the recommendations of Standard Review Plan Section 15.5.3, i.e., 30 rem thyroid and 2.5 rem whole-body. The calculated doses for the locked rotor event are within the acceptance criteria.}}

Latest revision as of 05:08, 29 October 2019

Provides Results of Offsite Dose Calculation for Reactor Coolant Pump Locked Rotor Event for Facility Cycle 8.Util Identified Previously Issued SERs Addressing Short Term Containment Analysis & LOCA Containment Integrity
ML17328A342
Person / Time
Site: Cook American Electric Power icon.png
Issue date: 07/23/1990
From: Alexich M
INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To: Murley T
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AEP:NRC:1071K, NUDOCS 9007260232
Download: ML17328A342 (6)
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Text

ACCELERATED DI RIBUTION DEMONST TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RXDS)

ACCESSION NBR:9007260232 DOC.DATE: 90/07/23 NOTARIZED: NO DOCKET FACIL:50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana 6 05000316 AUTH. NAME AUTHOR AFFILIATION ALEXXCH,M.P. Xndiana Michigan Power Co. (formerly Indiana 6 Michigan Ele RECIP.NAME RECIPXENT AFFILIATXON MURLEY,T.E. Document Control Branch (Document Control Desk)

SUBJECT:

Provides results of offsite dose calculation for reactor coolant pump locked rotor event for facility Cycle 8.

DISTRIBUTION CODE: A009D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR/Licensing Submittal: Appendix I NOTES RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 0 PD3-1 PD 5 5 GIITTER,J. 1 1 INTERNAL NRR/BRADFUTE i J 1 1 NRR/DREP DIR10E 1 0 NRR/DREP/PRPBll 1 1 NUDOCS-ABSTRACT 1 1 OC/LFM 1 0 OGC/HDS1 1 0 G 1 1 RGN3 DRSS/RPB 1 1 EXTERNAL EGGG AKERS i D 1 1 LPDR 1 1 NRC PDR 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 18 ENCL 14

indiana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 AEP:NRC:1071K Donald C. Cook Nuclear Plant Unit 2 docket No. 50-316 License. No. DPR-74 OFFSITE DOSE CALCULATION FOR THE REACTOR COOLANT PUMP LOCKED ROTOR EVENT FOR UNIT 2 CYCLE 8 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555 Attn: T. E. Murley .

July 23, l990

Dear Dr. Murley:

The purpose of this letter is to transmit the results of the offsite dose calculation for the reactor coolant pump locked rotor event for Unit 2 Cycle 8.

The license basis analysis of locked rotor event included in Appendix B, Section B.3.5.2 of Attachment 4 of our submittal AEP:NRC:1071E does not assume fuel failure based on DNB. However, Condition 9 of the SER for WCAP 10444-P-A, "Reference Core Report, Vantage 5 Fuel Assembly," states "With regard to the RCS pump shaft seizure accident, the fuel failure criterion should be the 95/95 DNBR limit. The mechanistic method mentioned in WCAP-20444 is not acceptable." Such an analysis showing 11$ rods in failure was submitted as Attachment 3 to our supplementary submittal AEP:NRC:1071I. Our fuel contractor, Westinghouse Electric Corporation has also calculated the expected offsite dose resulting from the failure of 11$ of the fuel rods. At the request of the staff, we are submitting a discussion of this calculation as an attachment to this letter. A preliminary copy of this discussion was telecopied to Timothy Colburn, Cook Nuclear Plant Project Manager, on July 18, 1990.

In a discussion with the staff on July 16, 1990, concern was expressed regarding our references to the "Shore-term Containment Analysis" and the "LOCA Containment Integrity" on page 34 of Attachment 4 to our submittal AEP:NRC:1071E. For the convenience of the staff, we identified previously issued SERs addressing these analyses. The first of these was attached to a letter dated "9007260232 900723 PDR P 'DC ADOCK 050003it~ gdof I/I

Dr. T. E. Murley AEP:NRC:1071K January 30, 1989 from John F. Stang of the NRC staff to Mr. Milton P. Alexich, Vice President of Indiana Michigan Power Company, regarding a "Change to Licensing Basis for SI and RHR Crossties."

This SER approves the LOCA containment integrity analysis for both units. The second SER was attached to Amendment No. 126 to Facility Operating License No. DPR-58, Cook Nuclear Plant Unit l.

The staff concluded "that the D. C. Cook Nuclear Plant's design basis pertaining to containment short term response... is adequate for RTP operation and therefore is acceptable." The SER specifies Unit 1 in its discussion. However, as indicated in WCAP 11902, (Attachment 2 to our submittal AEP:NRC:1067), on pg 3.4-2, "The analysis addressed the Cook Nuclear Plants [sic] Units 1 and 2, at rerated conditions assuming an NSSS power level of 3600 MWt, for a range of conditions which bound [the reduced temperature and pressure conditions) . . . ." The analysis also bounds the operating conditions proposed for Unit 2 Cycle 8. The assumptions also bound the conditions documented in Attachment 4 to AEP:NRC:1071E for Unit 2. We also note that as far as they impact this analysis, the Unit 1 and Unit 2 containments are identical.

One of these SERs and pertinent portions of the other were telecopied to the staff on July 18, 1990 for the convenience of the reviewer.

This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

Sincerely, M. P. Alexich Vice President MPA/eh cc: D. H. Williams, Jr.

A. A. Blind - Bridgman J. R. Padgett G. Charnoff NFEM Section Chief A. B. Davis - Region III NRC Resident Inspector - Bridgman

ATTACHMENT TO AEP:NRC:1071K LOCKED ROTOR DOSE ANALYSIS FOR DONALD C. COOK NUCLEAR PLANT UNIT 2 CYCLES 8 AND 9

~

'eactor Coolant Pum ocked Rotor Offsite Radiation Oose Anal sis A conservative analysis of the potential radiological consequences of a reactor coolant pump locked rotor event has been performed.

The salient assumptions used to calculate the activity releases and offsite doses follow.

1. Prior to the accident, the primary coolant iodine concentration is assumed to equal the Technical Specification transient limit for full power operation - 60 uCi/gram of dose equivalent I-131 (Technical Specification Figure 3.4-1)
2. Prior to the accident, the secondary coolant iodine concentration is assumed to equal the Technical Specification limit (LCO 3.7. 1.4) for full power operation - 0.1 uCi/gram of dose equivalent I-131
3. Eleven percent of the core is predicted to fail as a result of DNB. This results in the release of 11% of the core gap activity to the primary

'fcoolant. The fraction of core activity contained in the gap (gap fraction) is assumed to be 10% for all nuclides. Thus, a total of 1. 1%

the core activity is released.

4. The total primary-to-secondary leak rate is assumed to be at the Technical Specification limit (LCO 3.4.6.2 item c) of 1 gpm for the duration of the event.
5. An iodine partition coefficient of 100 between the steam generator liquid and steam phases is assumed. This value is suggested in Standard Review Plan (NUREG-0800) Section 15.6.3, Revision 2. It is a conservative estimate of the partition coefficient due to secondary coolant boil-off.
6. Offsite power is lost .
7. Steam release to the environment: 0 to 2 hr - 600,000 ibm 2 to 8 hr - 1,200,000 ibm
8. Eight hours after the accident the Residual Heat Removal System is assumed to start. No additional steam or radioactivity is released to the environment after 8 hrs.

Reactor core and coolant fission product inventories, dose conversion factors, gamma energies, atmospheric dispersion factors, and breathing rates are consistent with the values used in WCAP-12135.

Results Dose in rem Site boundary Low population zone Thyroid 3.0 6.0 Whole-body gamma 0.3 0.2

Conclusion The dose acceptance criteria is based on the recommendations of Standard Review Plan Section 15.5.3, i.e., 30 rem thyroid and 2.5 rem whole-body. The calculated doses for the locked rotor event are within the acceptance criteria.

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