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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:ACCELERATED DIRIBUTIONDEMONSTTIONSYSTEMREGULATORY INFORMATION DISTRIBUTION SYSTEM(RXDS)ACCESSION NBR:9007260232 DOC.DATE: | | {{#Wiki_filter:ACCELERATED DI RIBUTION DEMONST TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RXDS) |
| 90/07/23NOTARIZED: | | 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) |
| NODOCKETFACIL:50-316 DonaldC.CookNuclearPowerPlant,Unit2,Indiana605000316AUTH.NAMEAUTHORAFFILIATION ALEXXCH,M.P.
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| XndianaMichiganPowerCo.(formerly Indiana6MichiganEleRECIP.NAME RECIPXENT AFFILIATXON MURLEY,T.E.
| |
| DocumentControlBranch(Document ControlDesk)
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| ==SUBJECT:== | | ==SUBJECT:== |
| Providesresultsofoffsitedosecalculation forreactorcoolantpumplockedrotoreventforfacilityCycle8.DISTRIBUTION CODE:A009DCOPIESRECEIVED:LTR ENCLSIZE:TITLE:OR/Licensing Submittal:
| | Provides results of offsite dose calculation for reactor coolant pump locked rotor event for facility Cycle 8. |
| AppendixINOTESRECIPIENT IDCODE/NAME PD3-1LAGIITTER,J.
| | DISTRIBUTION CODE: A009D COPIES RECEIVED:LTR ENCL SIZE: |
| COPIESLTTRENCL1011RECIPIENT IDCODE/NAME PD3-1PDCOPIESLTTRENCL55INTERNALNRR/BRADFUTE iJ11NRR/DREP/PRPBll 11OC/LFM10G11NRR/DREPDIR10ENUDOCS-ABSTRACT OGC/HDS1RGN3DRSS/RPB10111011EXTERNALEGGGAKERSiDNRCPDR1111LPDR11NOTETOALL"RIDS"RECIPIENTS:
| | 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: |
| PLEASEHELPUSTOREDUCEWASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATE YOURNAMEFROMDISTRIBUTION LISTSFORDOCUMENTS YOUDON'TNEED!TOTALNUMBEROFCOPIESREQUIRED:
| | 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! |
| LTTR18ENCL14 indianaMichiganPowerCompanyP.O.Box16631Columbus, OH43216AEP:NRC:1071K DonaldC.CookNuclearPlantUnit2docketNo.50-316License.No.DPR-74OFFSITEDOSECALCULATION FORTHEREACTORCOOLANTPUMPLOCKEDROTOREVENTFORUNIT2CYCLE8U.S.NuclearRegulatory Commission Attn:DocumentControlDeskWashington, D.C.20555Attn:T.E.Murley.July23,l990
| | TOTAL NUMBER OF COPIES REQUIRED: LTTR 18 ENCL 14 |
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| ==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 . |
| Thepurposeofthisletteristotransmittheresultsoftheoffsitedosecalculation forthereactorcoolantpumplockedrotoreventforUnit2Cycle8.ThelicensebasisanalysisoflockedrotoreventincludedinAppendixB,SectionB.3.5.2ofAttachment 4ofoursubmittal AEP:NRC:1071E doesnotassumefuelfailurebasedonDNB.However,Condition 9oftheSERforWCAP10444-P-A, "Reference CoreReport,Vantage5FuelAssembly,"
| | July 23, l990 |
| states"WithregardtotheRCSpumpshaftseizureaccident, thefuelfailurecriterion shouldbethe95/95DNBRlimit.Themechanistic methodmentioned inWCAP-20444 isnotacceptable." | | |
| Suchananalysisshowing11$rodsinfailurewassubmitted asAttachment 3tooursupplementary submittal AEP:NRC:1071I.
| | ==Dear Dr. Murley:== |
| Ourfuelcontractor, Westinghouse ElectricCorporation hasalsocalculated theexpectedoffsitedoseresulting fromthefailureof11$ofthefuelrods.Attherequestofthestaff,wearesubmitting adiscussion ofthiscalculation asanattachment tothisletter.Apreliminary copyofthisdiscussion wastelecopied toTimothyColburn,CookNuclearPlantProjectManager,onJuly18,1990.Inadiscussion withthestaffonJuly16,1990,concernwasexpressed regarding ourreferences tothe"Shore-term Containment Analysis" andthe"LOCAContainment Integrity" onpage34ofAttachment 4tooursubmittal AEP:NRC:1071E.
| | |
| Fortheconvenience ofthestaff,weidentified previously issuedSERsaddressing theseanalyses.
| | 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. |
| Thefirstofthesewasattachedtoaletterdated"9007260232 900723PDRADOCK050003it~
| | 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. |
| P'DCgdofI/I Dr.T.E.Murley-2-AEP:NRC:1071K January30,1989fromJohnF.StangoftheNRCstafftoMr.MiltonP.Alexich,VicePresident ofIndianaMichiganPowerCompany,regarding a"ChangetoLicensing BasisforSIandRHRCrossties." | | 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 |
| ThisSERapprovestheLOCAcontainment integrity analysisforbothunits.ThesecondSERwasattachedtoAmendment No.126toFacilityOperating LicenseNo.DPR-58,CookNuclearPlantUnitl.Thestaffconcluded "thattheD.C.CookNuclearPlant'sdesignbasispertaining tocontainment shorttermresponse...
| | |
| isadequateforRTPoperation andtherefore isacceptable."
| | 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." |
| TheSERspecifies Unit1initsdiscussion.
| | 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. |
| However,asindicated inWCAP11902,(Attachment 2tooursubmittal AEP:NRC:1067), | | 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. |
| onpg3.4-2,"Theanalysisaddressed theCookNuclearPlants[sic]Units1and2,atreratedconditions assuminganNSSSpowerlevelof3600MWt,forarangeofconditions whichbound[thereducedtemperature andpressureconditions)
| | 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. |
| ...."Theanalysisalsoboundstheoperating conditions proposedforUnit2Cycle8.Theassumptions alsoboundtheconditions documented inAttachment 4toAEP:NRC:1071E forUnit2.Wealsonotethatasfarastheyimpactthisanalysis, theUnit1andUnit2containments areidentical. | | 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. |
| OneoftheseSERsandpertinent portionsoftheotherweretelecopied tothestaffonJuly18,1990fortheconvenience ofthereviewer.
| | Sincerely, M. P. Alexich Vice President MPA/eh cc: D. H. Williams, Jr. |
| Thisdocumenthasbeenpreparedfollowing Corporate procedures thatincorporate areasonable setofcontrolstoensureitsaccuracyandcompleteness priortosignature bytheundersigned.
| | A. A. Blind - Bridgman J. R. Padgett G. Charnoff NFEM Section Chief A. B. Davis - Region III NRC Resident Inspector - Bridgman |
| Sincerely, M.P.AlexichVicePresident MPA/ehcc:D.H.Williams, Jr.A.A.Blind-BridgmanJ.R.PadgettG.CharnoffNFEMSectionChiefA.B.Davis-RegionIIINRCResidentInspector | | |
| -Bridgman ATTACHMENT TOAEP:NRC:1071K LOCKEDROTORDOSEANALYSISFORDONALDC.COOKNUCLEARPLANTUNIT2CYCLES8AND9 | | ATTACHMENT TO AEP:NRC:1071K LOCKED ROTOR DOSE ANALYSIS FOR DONALD C. COOK NUCLEAR PLANT UNIT 2 CYCLES 8 AND 9 |
| ~'eactorCoolantPumockedRotorOffsiteRadiation OoseAnalsisAconservative analysisofthepotential radiological consequences ofareactorcoolantpumplockedrotoreventhasbeenperformed. | | |
| Thesalientassumptions usedtocalculate theactivityreleasesandoffsitedosesfollow.1.Priortotheaccident, theprimarycoolantiodineconcentration isassumedtoequaltheTechnical Specification transient limitforfullpoweroperation
| | ~ |
| -60uCi/gramofdoseequivalent I-131(Technical Specification Figure3.4-1)2.Priortotheaccident, thesecondary coolantiodineconcentration isassumedtoequaltheTechnical Specification limit(LCO3.7.1.4)forfullpoweroperation | | '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. |
| -0.1uCi/gramofdoseequivalent I-1313.Elevenpercentofthecoreispredicted tofailasaresultofDNB.Thisresultsinthereleaseof11%ofthecoregapactivitytotheprimarycoolant.Thefractionofcoreactivitycontained inthegap(gapfraction) isassumedtobe10%forallnuclides. | | The salient assumptions used to calculate the activity releases and offsite doses follow. |
| Thus,atotalof1.1%'fthecoreactivityisreleased. | | : 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) |
| 4.Thetotalprimary-to-secondary leakrateisassumedtobeattheTechnical Specification limit(LCO3.4.6.2itemc)of1gpmforthedurationoftheevent.5.Aniodinepartition coefficient of100betweenthesteamgenerator liquidandsteamphasesisassumed.Thisvalueissuggested inStandardReviewPlan(NUREG-0800) | | : 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 |
| Section15.6.3,Revision2.Itisaconservative estimateofthepartition coefficient duetosecondary coolantboil-off.
| | : 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 |
| 6.Offsitepowerislost.7.Steamreleasetotheenvironment: | | '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% |
| 0to2hr-600,000ibm2to8hr-1,200,000 ibm8.EighthoursaftertheaccidenttheResidualHeatRemovalSystemisassumedtostart.Noadditional steamorradioactivity isreleasedtotheenvironment after8hrs.Reactorcoreandcoolantfissionproductinventories, doseconversion factors,gammaenergies, atmospheric dispersion factors,andbreathing ratesareconsistent withthevaluesusedinWCAP-12135.
| | the core activity is released. |
| ResultsDoseinremSiteboundaryThyroid3.0Whole-body gamma0.3Lowpopulation zone6.00.2 Conclusion Thedoseacceptance criteriaisbasedontherecommendations ofStandardReviewPlanSection15.5.3,i.e.,30remthyroidand2.5remwhole-body.
| | : 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. |
| Thecalculated dosesforthelockedrotoreventarewithintheacceptance criteria.}}
| | : 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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ML17326A1531999-09-16016 September 1999
Submits Info Pertaining to Plant Proposed Operator Licensing Exam Requirements Through Yr 2003.NRC Form 536, Operator Licensing Exam Data, Which Provides Required Info Encl
ML17326A1101999-08-27027 August 1999
Forwards LER 99-021-00, GL 96-01 Test Requirements Not Met in Surveillance Tests. List of Commitments Identified in LER Provided
ML17326A0991999-08-26026 August 1999
Forwards LER 99-020-00,re EDGs Being Declared Inoperable. Commitments Made by Util Are Listed
ML17326A1221999-08-23023 August 1999
Forwards Revised Page 2 to 1998 Annual Environ Operating Rept, for DC Cook Nuclear Plant,Correcting Omission to App I
ML17326A0981999-08-23023 August 1999
Forwards fitness-for-duty Program Performance Data for Period of 990101-0630 for DC Cook Nuclear Plants,Units 1 & 2,per 10CFR26.71(d)
ML17326A0891999-08-16016 August 1999
Forwards LER 99-019-00,re Victoreen Containment High Range Monitors Not Beign Environmentally Qualified to Withstand post-LOCA Conditions.Commitments Made by Util Are Listed
ML17326A0811999-08-10010 August 1999
Notifies NRC of Changes in Commitments Made in Response to GL 98-01,supplement 1, Yr 2000 Readiness of Computer Sys Ar Npps, Dtd 990623
ML17326A0821999-08-0606 August 1999
Informs That Util Is Submitting Encl Scope & Objectives for 991026 DC Cook Nuclear Plant Emergency Plan Exercise to G Shear of NRC Plant Support Branch.Exercise Will Include Full State & County Participation
ML17326A1451999-08-0404 August 1999
Requests Withholding of WCAP-15246, Control Rod Insertion Following Cold Leg Lbloca.
ML17326A0751999-08-0404 August 1999
Forwards LER 98-029-01, Fuel Handling Area Ventilation Sys Inoperable Due to Original Design Deficiency. Supplemental Rept Represents Extensive Rev to Original LER & Replaces Rept in Entirely.Commitment Listed
ML17326A0721999-07-29029 July 1999
Forwards LER 99-018-00 Re Refueling Water Storage Tank Suction Motor Operated Valves Inoperable,Due to Inadequate Design.Listed Commitments Were Identified in LER
ML17326A0711999-07-27027 July 1999
Responds to 980123 RAI Re NRC GL 87-02, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Unresolved Safety Issue (USI) A-46.
ML17326A0601999-07-22022 July 1999
Forwards UFSAR, IAW 10CFR50.71(e) & Rept of Changes,Tests & Experiments as Required by 10CFR50.59(b)(2) for DC Cook Nuclear Plant,Units 1 & 2.Without UFSAR
ML17326A0631999-07-22022 July 1999
Forwards LER 98-014-03, Response to High-High Containment Pressure Procedure Not Consistent with Analysis of Record. Revised Info Marked by Sidebars in Right Hand Margin. Commitments Made by Util,Listed
ML17326A0311999-07-0101 July 1999
Forwards LER 99-004-01 Re Failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed. Commitments Made by Util Are Listed
ML17326A0281999-06-28028 June 1999
Provides Response to 981116 & 960228 RAIs Re GL 92-01. Revised Pressurized Thermal Shock Evaluation Based on New Weld Chemistry Info & Copy of W Rept WCAP-15074, Evaluation of 1P3571 Weld Metal from Surveillance Programs... Encl
ML17326A0241999-06-23023 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants. Y2K Readiness Disclosure for Plant & List of Commitments Encl
ML17326A0121999-06-18018 June 1999
Forwards LER 99-014-00 Re Requirement of TS 4.0.5 Not Met for Boron Injection Tank Bolting.Commitments Identified in Submittal Listed
ML17326A0111999-06-11011 June 1999
Provides Response to NRC RAI Re GL 97-01, Degradation of Crdm/Cedm Nozzle & Other Vessel Closure Head Penetrations.
ML17325B6401999-06-0101 June 1999
Forwards LER 99-013-00 Re Safety Injection & Centrifugal Charging Throttle Valve Cavitation During LOCA Could Lead to ECCS Pump Failure.Listed Commitments Identified in Submittal
ML17325B6281999-06-0101 June 1999
Forwards LER 99-S03-00,re Nonconforming Vital Area Barriers.Commitments Made by Util Are Listed
ML17325B6331999-05-28028 May 1999
Forwards LER 99-S02-00,re Vulnerability in Safeguard Sys That Could Allow Unauthorized or Undetected Access to Protected Area.Commitments Made by Util Are Listed
ML17265A8201999-05-24024 May 1999
Forwards LER 98-037-01,representing Extensive Rev to Original LER & Replacing Rept in Entirety.Listed Commitments Identified in Submittal
ML17325B6111999-05-21021 May 1999
Forwards Annual Radioactive Effluent Release Rept for 980101-1231 for DC Cook Nuclear Plant,Units 1 & 2. Transmittal of Submittal Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17325B6031999-05-21021 May 1999
Provides Response to NRC GL 98-04, Potential for Degradation of ECCS & Containment Spray Sys After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment.
ML17325B5971999-05-20020 May 1999
Forwards LER 99-012-00,re Auxiliary Building ESF Ventilation Sys Not Being Capable of Maintaining ESF Room Temps post-accident.Commitment,listed
ML17335A5281999-05-12012 May 1999
Forwards DC Cook Nuclear Plant Fitness for Duty Program Performance Dtd for six-month Period of 980701-1231,IAW 10CFR26.71(d).Info Was Delayed Due to Administrative Error in Regulatory Affairs Dept
ML17335A5271999-05-11011 May 1999
Forwards Details Re Sources & Levels of Insurance Maintained for DC Cook,Units 1 & 2,as of 990401,per 10CFR50.54(w)(3). Info Was Delayed Beyond Required Date Due to Internal Oversight
ML17325B5841999-05-10010 May 1999
Forwards LER 99-002-00 Re TS 4.0.5 Requirements Not Being Met Due to Improperly Performed Test.Commitments Identified in Ler,Listed
ML17325B5871999-05-0707 May 1999
Forwards Current Revs of Expanded Sys Readiness Review (Essr) Implementing Procedures,For Info Purposes to Support Current NRC Insps.Current Esrr Schedule Provided for Info Purposes,Reflecting Revised Target Dates
ML17325B5821999-05-0404 May 1999
Provides Addl Background,Description & Clarification of Previous & Revised Commitments Re UFSAR Revalidation Effort. Commitment Change Involved Alignment of UFSAR Revalidation Program Methodology to Strategy Contained in Current Plan
ML17325B5791999-05-0404 May 1999
Forwards LER 99-011-00,concerning Air Sys for EDG Not Supporting Long Term Operability.Commitments Made by Util Listed
ML17325B5741999-05-0303 May 1999
Forwards LER 99-010-00 Re RCS Leak Detection Sys Sensitivity Not in Accoradnce with Design Requirements.Listed Commitments Identified in Submittal
ML17325B5631999-04-22022 April 1999
Forwards Results of Independent Chemical Evaluations Performed from Sept 1997 Through Feb 1999,re Resolution of Issues Related to License Amend 227
ML17325B5561999-04-16016 April 1999
Forwards LER 99-006-00, Fuel Crane Loads Lifted Over SFP Could Impact Energies Greater than TS Limits, IAW 10CFR50.73.Submittal Was Delayed to Allow for Resolution of Questions.Commitment Made by Licensee,Listed
ML17325B5451999-04-12012 April 1999
Forwards LER 99-009-00 Re as-found RHR Safety Relief Valve Lift Setpoint Greater than TS Limit.Commitments Identified in Submittal Listed
ML17325B5301999-04-0707 April 1999
Forwards LER 99-S01-01, Vulnerability in Locking Mechanism of Four Vital Area Gates, Per 10CFR50.73.Commitments Made by Util,Listed
ML17325B5241999-04-0505 April 1999
Forwards Revs 0 & 1 to Cook Nuclear Plant Restart Plan, Dtd 980307 & 0407.Rev 5 Is Current Cook Nuclear Plant Restart & Supercedes Previous Revs in All Respects
ML17325B5121999-04-0101 April 1999
Forwards LER 99-007-00, Calculations Show That Divider Barrier Between Upper & Lower Containment Vols May Be Overstressed. Commitments Made by Util Are Listed
ML17325B5141999-03-30030 March 1999
Forwards Rept on Status of Decommissioning Funding.Attached Rept Includes Amount of Decommissioning Funds Estimated to Be Required Pursuant to 10CFR50.75(b) & (C)
ML17325B5191999-03-29029 March 1999
Forwards LER 99-001-00,re Degraded Component Cw Flow to Containment Main Steam Line Penetrations.Commitment, Listed
ML20204F6401999-03-19019 March 1999
Responds to NRC 981013 NOV & Proposed Imposition of Civil Penalty.Violations Cited in Subject NOV Were Initially Identified in Referenced Five Insp Repts.Corrective Actions: Ice Condensers Have Been Completely Thawed of Any Blockage
ML17325B4751999-03-18018 March 1999
Forwards LER 99-004-00,re Failure to Perform TS Surveillance Analyses of Reactor Coolant Chemistry with Fuel Removed. Commitment Made by Util,Listed
ML17325B4721999-03-18018 March 1999
Forwards LER 99-005-00,re Reactor Trip Breaker Manual Actuations During Rod Drop Testing Not Previously Reported. Listed Commitments Identified in Submittal
ML17325B4641999-03-17017 March 1999
Withdraws Response to Issue 1 of NRC Cal,Dtd 970919. Comprehensive Design Review Effort in Progress to Validate Resolution of Issue for Future Operation
1999-09-30
[Table view]
Category:UTILITY TO NRC
MONTHYEARML17328A4481990-09-21021 September 1990
Requests Withdrawal of Mode 6 Proposed Tech Spec Change Since Issue Currently Being Addressed by New STSs
ML17328A4331990-08-27027 August 1990
Informs of Preliminary Assessment of 900713 Electrical Contact Accident at Facility.Investigation Concludes That No Safety Rules Violated
ML17328A4001990-08-24024 August 1990
Responds to 900725 Ltr Re Commitments Made in Response to Generic Ltr 88-14 on Instrument Air Supply Problems Affecting safety-related Equipment.Simulator Training to Be Provided as Stated
ML17334B3781990-08-24024 August 1990
Forwards Info to Certify Plant Simulator Facility.Results of Evaluation of Dual Unit Simulation Facilities Demonstrate That Plant Simulator Performance Compares Favorably W/Units
ML17328A4231990-08-21021 August 1990
Forwards Under Separate Cover,Semiannual Radioactive Effluent Release Rept for Jan-June 1990
ML17328A3901990-08-17017 August 1990
Responds to NRC 900720 Ltr Re Violations Noted in Onsite Audit of Spds.Corrective Actions:Apparent Disparity in Selection of Reactor Trip & Sys Capacity or Anticipatory Values Rectified
ML17328A3891990-08-15015 August 1990
Forwards Performance Data Sheets for Plant fitness-for-duty Program for Period Jan-June 1990,per 10CFR26.Encl Includes Statistics on Various Categories of Testing,Substances Tested for & cut-off Levels Used
ML17328A3571990-08-0202 August 1990
Responds to Open Items in Safety Evaluation of Util Response to Unresolved Issues on post-fire Safe Shutdown Methodology. Open Items 1 & 21 Will Be Closed by Implementing Plant Procedures Providing Equivalent Degree of Protection
ML17328A3431990-07-23023 July 1990
Requests Withdrawal of 890830 Proposed Tech Spec Changes Re Sections 3.0 & 4.0.Util Will Resubmit Generic Ltr 87-09 Recommended Improvement Items on Case by Case Basis
ML17328A3421990-07-23023 July 1990
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
ML20055G7551990-07-18018 July 1990
Responds to NRC Re Violations Noted in Insp Repts 50-315/90-10 & 50-316/90-10.Corrective Actions:Required Review Performed & Updated Procedure 12 Mhp 5021.019.001 Revised & Issued
ML17328A3191990-07-12012 July 1990
Responds to NRC 900601 Ltr Re Violations Noted in Insp Repts 50-315/89-31 & 50-316/89-31.Corrective Actions:Electrical Testing Techniques Improved & Surveillance Procedures for Feedwater Pumps Will Be Rewritten
ML17328A3181990-07-0909 July 1990
Responds to Generic Ltr 90-03,Suppl 1, Relaxation of Staff Position in Generic Ltr 83-28,Item 2.2,Part 2, 'Vendor Interface for Safety-Related Components.' Licensee Will Contact safety-related Vendors on Annual Basis
ML17328A3111990-07-0303 July 1990
Provides Certification of Funding Plan for Decommissioning of Plant,Per 10CFR50.33 & 50.75
ML17328A2961990-06-25025 June 1990
Forwards Response to Generic Ltr 90-04 Re Closeout of Generic Safety Issues.Util Supports Concern & Desire to Close long-standing Generic Safety Issues
ML17328A2921990-06-22022 June 1990
Forwards WCAP-12483, Analysis of Capsule U from American Electric Power Co DC Cook Unit 1 Reactor Vessel Radiation Surveillance Program.
ML20044A3521990-06-22022 June 1990
Submits Info Re Sensitivity Study Performed on Number of Fuel Axial Intervals,Per Topical Rept, American Electric Power Reactor Core Thermal-Hydraulic Analysis Using Cobra III-C/MIT-2 Computer Code.
ML17328A2821990-06-15015 June 1990
Submits Ltr Re Proposed Control Room Habitability Tech Spec Changes & Supporting Analyses,Per 900521 Discussion.Revised Calculations of Control Room Thyroid Doses Will Be Submitted within 60 Days of Receipt of Proposed Generic Ltr
ML17328A2741990-06-12012 June 1990
Submits Followup,Per 900205 & 0308 Ltrs & Provides Update Re Inoperable Fire Barrier.Fire Seal Repaired & Restored to Operability on 900419
ML17328A2551990-06-0505 June 1990
Forwards Addl Info Re Util 900126 Revised Response to NRC Bulletin 88-002,per NRC 900509 Request
ML17328A7361990-06-0101 June 1990
Forwards Addl Info Re 890825 & 1212 Applications for Amends to Licenses DPR-58 & DPR-74,per Request.Amends Make Changes to Administrative Controls
ML17328A7331990-05-29029 May 1990
Forwards Nonproprietary WCAP-12577 & Proprietary WCAP-12576, Westinghouse Revised Thermal Design Procedure Instrument... Methodology for American Electric Power DC Cook Unit 2 Nuclear Power Station, Per 900419 Commitment
ML17328A7321990-05-24024 May 1990
Responds to NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. No Rosemount Model 1153 Series B or D or Model 1154 Transmitters Installed at Facility.Transmitters Purchased as commercial-grade Units
ML17328A6991990-05-0909 May 1990
Provides Suppl to 900103 Response,Certifying That Fitness for Duty Program Implemented at Plant.Change Also Clarifies When More Stringent cut-off Levels Adopted at Plant Program Apply
ML17328A7001990-05-0909 May 1990
Responds to NRC 900406 Ltr Re Inadequacies of Spds,Per Audit on 900221-22.Corrective Actions:Software Mod Will Prevent User from Accessing Displays Other than Iconic Displays from SPDS Dedicated Terminal
ML17328A7071990-05-0707 May 1990
Forwards Monthly Operating Rept for Apr 1990 & Corrected Rept Mar 1990 for Donald C Cook Nuclear Plant Unit 1
ML17328A7031990-05-0707 May 1990
Forwards Monthly Operating Rept for Apr 1990 & Corrected Rept for Mar 1990 for DC Cook Nuclear Plant Unit 2
ML17328A6951990-04-30030 April 1990
Certifies That Training Programs for Initial Licensing & Requalification Training of Operators & Senior Operators at Plant Accredited by Inpo,Per Generic Ltr 87-07
ML17328A6861990-04-30030 April 1990
Forwards Annual Environ Operating Rept,Jan-Dec 1989, & DC Cook Nuclear Plant Units 1 & 2 Operational Radiological Environ Monitoring Program 1989 Rept Jan-Dec 1989.
ML17328A6801990-04-23023 April 1990
Responds to NRC Bulletin 88-004, Potential Safety-Related Pump Loss. Study Planned to Evaluate Alternative Actions to Be Taken for Protecting RHR Pumps
ML17328A6771990-04-23023 April 1990
Forwards DC Cook Nuclear Plant 1990 Annual Emergency Preparedness Exercise. Exercise Scope & Objectives & Detailed Scenario Documentation Encl
ML17334B3641990-04-11011 April 1990
Responds to NRC 900301 Ltr Re Violations Noted in Insp Repts 50-315/89-31 & 50-316/89-31.Corrective Actions:Reliability Centered Maint Program Initiated
ML17328A6601990-04-11011 April 1990
Forwards Updated QA Program Description for Cook Nuclear Plant, Incorporating Editorial,Organizational & Position Title Changes
ML17328A6551990-04-0909 April 1990
Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-316/90-08.Corrective Actions:Procedure for Steam Generator Stop Valve Operability Test Revised to Require That MSIV Be Declared Inoperable After Failing Stroke Test
ML17334B3601990-04-0606 April 1990
Forwards Revised Figures for Loss of Load Event Previously Submitted in Attachment 4,App B of Vantage 5 Reload Transition Safety Rept Supplied by Westinghouse in Jan 1990. Update Is Editorial in Nature
ML17334B3621990-04-0606 April 1990
Forwards 1989 Annual Rept & Projected Cash Flow
ML17328A6341990-03-30030 March 1990
Forwards Application for Renewal of Plant NPDES Permit,For Info,Per Section 3.2 of App B to Plant Tech Specs.W/O Encl
ML17328A6331990-03-30030 March 1990
Responds to NUMARC 900104 Request for Addl Info Re Station Blackout Submittals.Target Reliability for Emergency Diesel Generator of 0.0975 Established for 4 H ac-independent Coping Category Will Be Maintained
ML17328A6321990-03-30030 March 1990
Forwards Response to Generic Ltr 90-01, NRC Regulatory Impact Survey. Util Ack Recommendation by NUMARC
ML17328A6221990-03-27027 March 1990
Responds to 900226 Ltr Transmitting Notice of Violation & Proposed Imposition of Civil Penalty in Amount of $75,000. Corrective Actions:Flow Retention Circuitry Setpoints Set to Compensate for Missized Process Flow Orifice
ML17328A6191990-03-27027 March 1990
Forwards Payment in Amount of $75,000 for Civil Penalty Imposed Through Notice of Violation Issued in Insp Rept 50-316/89-02 on 891016-20,24-26 & 1204.Response to Violation & Corrective Actions Provided in Separate Submittal
ML17328A6161990-03-20020 March 1990
Responds to Generic Ltr 89-19 Re Safety Implications of Control Sys in LWRs (USI A-47).Both Units Have Steam Overfill Protection Sys That Would Automatically Prevent Water Carryover Into Steam Lines If Control Sys Failed
ML17328A6121990-03-13013 March 1990
Forwards Monthly Performance Monitoring Rept,Jan 1990.
ML17325B3901990-03-0606 March 1990
Forwards Annual Rept to NRC Per 10CFR50.54(W)(2) Re Nuclear Property Insurance
ML17325B3921990-03-0606 March 1990
Modifies Application for Amend to License DPR-74 Re Cycle 8, Per 900228 Telcon
ML17325B3781990-02-28028 February 1990
Forwards Response to NRC Info Notice 89-056 Re Questionable Certification of Matl Supplied to Dod by Nuclear Suppliers. Matl Capable of Performing Design Function & Acceptable for Continued Use
ML17328A5971990-02-27027 February 1990
Withdraws 890203 Application for Amend to License DPR-74, Modifying Tech Spec Table 3.2-1 Re DNB Parameters to Express RCS Flow Rate on Volumetric Rather than Mass Basis
ML17334B3511990-02-22022 February 1990
Submits Annual Rept of Changes to or Errors in Acceptable LOCA Evaluation Models or Application of Models for Plants. Wflash Analyses Will Be Superceded by Notrump Analyses for Unit 2,Cycle 8 Reload
ML17328A5881990-02-21021 February 1990
Responds to NRC 890914 Request for Addl Info Re Safe Shutdown Methodology.App R Fire Barriers Being Maintained & Surveilled Under 3/4.7.10 for Units 1 & 2
ML17328A5861990-02-16016 February 1990
Provides Revised Comments in Response to NRR Comments During 891213 Telcon Re Allowable Stresses for Piping & Piping Supports
1990-09-21
[Table view] |
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.
