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| {{#Wiki_filter:<<ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEMREGULATORY INFORMATION DISTRIBUTION SYSTEM(RIDS)ACCESSION NBR:9007170340 DOC.DATE: | | {{#Wiki_filter:<<ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9007170340 DOC.DATE: 90/07/09 NOTARIZED: |
| 90/07/09NOTARIZED: | | NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION MECREDY,R. |
| NOFACIL:50-244 RobertEmmetGinnaNuclearPlant,Unit1,Rochester GAUTH.NAMEAUTHORAFFILIATION MECREDY,R.
| | Rochester Gas 6 Electric Corp.RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. |
| Rochester Gas6ElectricCorp.RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. | | Project Directorate I-3 DOCKET 05000244 |
| ProjectDirectorate I-3DOCKET05000244
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| ==SUBJECT:== | | ==SUBJECT:== |
| Discusses designbasisofplantreAOV745,CCWIreturnvalvefromexcessletdownHX.DISTRIBUTION CODE:AOOIDCOPIESRECEIVED:LTR gENCL~SIZE:~TITLE:ORSubmittal: | | Discusses design basis of plant re AOV 745,CCWI return valve from excess letdown HX.DISTRIBUTION CODE: AOOID COPIES RECEIVED:LTR g ENCL~SIZE:~TITLE: OR Submittal: |
| GeneralDistribution NOTES:License Expdateinaccordance with10CFR2,2.109(9/19/72).
| | General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). |
| 05000244RECIPIENT IDCODE/NAME PD1-3LAJOHNSON,A COPIESLTTRENCLRECIPIENT IDCODE/NAME PD1-3PDCOPIESLTTRENCLINTERNAL:
| | 05000244 RECIPIENT ID CODE/NAME PD1-3 LA JOHNSON,A COPIES LTTR ENCL RECIPIENT ID CODE/NAME PD1-3 PD COPIES LTTR ENCL INTERNAL: NRR/DET/ECMB 9H NRR/DST 8E2 NRR/DST/SICB 7E NUDOCS-ABSTRACT OGC/HDS2 RES/DSIR/EIB EXTERNAL: LPDR NSIC NRR/DOEA/OTSB11 1 NRR/DST/SELB 8D 1 NRR/DST/SRXB 8E 1 ,OC/-LFMB E 1 REG FKE~>>=01~~ |
| NRR/DET/ECMB 9HNRR/DST8E2NRR/DST/SICB 7ENUDOCS-ABSTRACT OGC/HDS2RES/DSIR/EIB EXTERNAL: | | NRC PDR NOTE TO ALL"RIDS" RECIPIENTS: |
| LPDRNSICNRR/DOEA/OTSB11 1NRR/DST/SELB 8D1NRR/DST/SRXB 8E1,OC/-LFMB E1REGFKE~>>=01~~
| | PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!TOTAL NUMBER OF COPIES REQUIRED: LTTR 21 ENCL y |
| NRCPDRNOTETOALL"RIDS"RECIPIENTS:
| | >~f f//51/II I//I'I/t/AND ESgPII///////SS////SF'/F//F 5TAYE ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y.14649-0001 July 9, 1990 TELEPHONE AREA COOE Tld 546.2700 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Allen R.Johnson Project Directorate I-3 Washington, D.C.20555 |
| PLEASEHELPUSTOREDUCEWASTE!CONTACTTHEDOCUMENTCONTROLDESK,ROOMPl-37(EXT.20079)TOELIMINATE YOURNAMEFROMDISTRIBUTION LISTSFORDOCUMENTS YOUDON'TNEED!TOTALNUMBEROFCOPIESREQUIRED:
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| LTTR21ENCL y
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| >~ff//51/III//I'I/t/ | |
| ANDESgPII///////SS////SF'/F//F5TAYEROCHESTER GASANDELECTRICCORPORATION o89EASTAVENUE,ROCHESTER, N.Y.14649-0001 July9,1990TELEPHONE AREACOOETld546.2700U.S.NuclearRegulatory Commission DocumentControlDeskAttn:AllenR.JohnsonProjectDirectorate I-3Washington, D.C.20555
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| ==Subject:== | | ==Subject:== |
| Rochester GasandElectricCorporation R.E.GinnaNuclearPowerPlantDocketNo.50-244 | | Rochester Gas and Electric Corporation R.E.Ginna Nuclear Power Plant Docket No.50-244 |
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| ==DearMr.Johnson,== | | ==Dear Mr.Johnson,== |
| OnJune8,1990,RG&EnotifiedtheNRCunderthecriterion of10CFR550.72, ofacondition outsidethedesignbasisoftheplantassociated withAOV745,thecomponent coolingwaterisolation returnvalvefromtheexcessletdownheatexchanger.
| | On June 8, 1990, RG&E notified the NRC under the criterion of 10CFR550.72, of a condition outside the design basis of the plant associated with AOV 745, the component cooling water isolation return valve from the excess letdown heat exchanger. |
| Thisnotification wasmadeasaresultofinformation contained inTable3.6-1oftheGinnaTechnical Specifications andUFSARTable6.2-13,whichindicated thatAOV745shouldreceiveaContainment Isolation Signal(CIS).However,elementary wiringdrawingsdidnotshowthisvalveasreceiving suchasignal.Subsequent tothisnotification, RG&Edetermined thatinformation contained onTechnical Specification Table3.6-1forMOV749AandMOV749B(component coolingwaterisolation valvestothereactorcoolantpumps)wasalsopotentially misleading basedonthefactthatthesevalvesdonotreceiveaCISbutareshownwithamaximumisolation timeof60seconds.Consequently, RG&Ehasreviewedthebackground information available withrespecttoAOV745,MOV749A,andMOV749B.AllthreevalveswerereviewedbytheNRCduringimplementation oftheTMILessonsLearnedrecommendations andtheSystematic Evaluation Program.Penetration 124awasevaluated inthesetwoprogramsassumingthatitwasnonessential, andthatAOV745wasanormallyopenvalvewhichdidnotreceiveaCISbutwasclosedpost-LOCA.
| | This notification was made as a result of information contained in Table 3.6-1 of the Ginna Technical Specifications and UFSAR Table 6.2-13, which indicated that AOV 745 should receive a Containment Isolation Signal (CIS).However, elementary wiring drawings did not show this valve as receiving such a signal.Subsequent to this notification, RG&E determined that information contained on Technical Specification Table 3.6-1 for MOV 749A and MOV 749B (component cooling water isolation valves to the reactor coolant pumps)was also potentially misleading based on the fact that these valves do not receive a CIS but are shown with a maximum isolation time of 60 seconds.Consequently, RG&E has reviewed the background information available with respect to AOV 745, MOV 749A, and MOV 749B.All three valves were reviewed by the NRC during implementation of the TMI Lessons Learned recommendations and the Systematic Evaluation Program.Penetration 124a was evaluated in these two programs assuming that it was nonessential, and that AOV 745 was a normally open valve which did not receive a CIS but was closed post-LOCA. |
| Meanwhile, Penetrations 127(MOV749A)and128(MOV749B)wereconsidered essential penetrations withnormallyopenedvalvesthatalsodidnotreceiveaCISbutremainopenpost-LOCA. | | Meanwhile, Penetrations 127 (MOV 749A)and 128 (MOV 749B)were considered essential penetrations with normally opened valves that also did not receive a CIS but remain open post-LOCA. |
| InaNovember19,1979lettertotheNRCdiscussing TMILessonsLearneptopics,RG&EstatedthatPenetration 124awas"nonesbential" whilePenetrations 127and128were"essential".
| | In a November 19, 1979 letter to the NRC discussing TMI Lessons Learnep topics, RG&E stated that Penetration 124a was"nonesbential" while Penetrations 127 and 128 were"essential". |
| Inpreviouscorrespondence fromRG&Eonthissubject(datedJune22,1979),atablewasprovidedwhichlistedallvalvesreceiving aCIS;neitherAOV745,MOV749A,norMOV749Bwerelistedonthistable.9007l70340 900709PDRAGOCK05000244pplleap
| | In previous correspondence from RG&E on this subject (dated June 22, 1979), a table was provided which listed all valves receiving a CIS;neither AOV 745, MOV 749A, nor MOV 749B were listed on this table.9007l70340 900709 PDR AGOCK 05000244 ppl leap |
| /ep44I InTable1oftheFinalEvaluation ReportforSEPTopicVI-4,datedApril12,1982,itisindicated thattheNRCassumedthatAOV745wasnormallyopen,butclosedpost-LOCA. | | /e p 4 4I In Table 1 of the Final Evaluation Report for SEP Topic VI-4, dated April 12, 1982, it is indicated that the NRC assumed that AOV 745 was normally open, but closed post-LOCA. |
| However,thereisnoinformation listedbeneaththe"Actuation" columnforthisvalvewhichwouldindicatehowthevalvewasexpectedtoclose.ThistablealsoshowsMOV749AandMOV749Basnormallyopenvalveswhichremainopenpost-LOCA. | | However, there is no information listed beneath the"Actuation" column for this valve which would indicate how the valve was expected to close.This table also shows MOV 749A and MOV 749B as normally open valves which remain open post-LOCA. |
| Thesevalvesareshownas"remotemanual"under"Actuation".
| | These valves are shown as"remote manual" under"Actuation". |
| Inaddition, drawingsandtheoriginalGinnaFSARasusedbytheNRCduringtheSEPdidnotshowthattherewasaCIStothesethreevalves.Therefore, theNRChadevaluated AOV745,MOV749AandMOV749Bintheircurrentconfigurations duringboththeTMILessonsLearnedandSystematic Evaluation
| | In addition, drawings and the original Ginna FSAR as used by the NRC during the SEP did not show that there was a CIS to these three valves.Therefore, the NRC had evaluated AOV 745, MOV 749A and MOV 749B in their current configurations during both the TMI Lessons Learned and Systematic Evaluation Programs, and found them to be acceptable. |
| : Programs, andfoundthemtobeacceptable.
| | Since there was no conflicting supporting design basis information (e.g., UFSAR, elementary drawings)showing that MOV 749A and MOV 749B receive a CIS, these valves were not declared inoperable when it was determined that they did not receive a CIS, as had been done with AOV 745.Based on our review of this situation, RG&E interprets the 60 second maximum isolation time as listed in Technical Specification Table 3.6-1 to be valve closure time following a demand to close.This interpretation is consistent with Standard Review Plan 6.2.4, section II (n)and has been transmitted to the Ginna Station operation staff.'Therefore, no changes are required with respect to MOV 749A and MOV 749B.Penetration 124a is a closed system inside containment which is seismic and missile-protected. |
| Sincetherewasnoconflicting supporting designbasisinformation (e.g.,UFSAR,elementary drawings) showingthatMOV749AandMOV749BreceiveaCIS,thesevalveswerenotdeclaredinoperable whenitwasdetermined thattheydidnotreceiveaCIS,ashadbeendonewithAOV745.Basedonourreviewofthissituation, RG&Einterprets the60secondmaximumisolation timeaslistedinTechnical Specification Table3.6-1tobevalveclosuretimefollowing ademandtoclose.Thisinterpretation isconsistent withStandardReviewPlan6.2.4,sectionII(n)andhasbeentransmitted totheGinnaStationoperation staff.'Therefore, nochangesarerequiredwithrespecttoMOV749AandMOV749B.Penetration 124aisaclosedsysteminsidecontainment whichisseismicandmissile-protected.
| | The current GDC 57 only requires a manual or remote-manual valve outside containment for this configuration. |
| ThecurrentGDC57onlyrequiresamanualorremote-manual valveoutsidecontainment forthisconfiguration.
| | The safety function (i.e., containment isolation) of this penetration is met in the current configuration, even with AOV 745 normally open.RG&E has thus concluded that AOV 745 does not require a CIS to be considered operable since the containment isolation function associated with Penetration 124a is accomplished by the closed system inside containment and AOV 745 (without a CIS)outside containment. |
| Thesafetyfunction(i.e.,containment isolation) ofthispenetration ismetinthecurrentconfiguration, evenwithAOV745normallyopen.RG&Ehasthusconcluded thatAOV745doesnotrequireaCIStobeconsidered operablesincethecontainment isolation functionassociated withPenetration 124aisaccomplished bytheclosedsysteminsidecontainment andAOV745(withoutaCIS)outsidecontainment.
| | Therefore, the criteria of 10CFR550.73 are not applicable to the discovery that AOV 745 does not receive a CIS.However, because our response to item 2.1.4.3 in the above cited November 19, 1979 letter indicated that we would provide a CIS to valves in non-essential penetrations, even though such a signal goes beyond the requirements of GDC 57, RG&E commits to install such a signal to AOV 745 by the end of the 1992 refueling outage.In the meantime, in order to provide another level of assurance of containment isolation, RG&E is planning to modify the Emergency Procedures to direct the operators to remote-manually close this valve from the Control Room at an appropriate time following receipt of a CIS (e.g., following immediate actions of E-0).The addition of this procedure step will be accomplished by August 15, 1990. |
| Therefore, thecriteriaof10CFR550.73 arenotapplicable tothediscovery thatAOV745doesnotreceiveaCIS.However,becauseourresponsetoitem2.1.4.3intheabovecitedNovember19,1979letterindicated thatwewouldprovideaCIStovalvesinnon-essential penetrations, eventhoughsuchasignalgoesbeyondtherequirements ofGDC57,RG&EcommitstoinstallsuchasignaltoAOV745bytheendofthe1992refueling outage.Inthemeantime, inordertoprovideanotherlevelofassurance ofcontainment isolation, RG&EisplanningtomodifytheEmergency Procedures todirecttheoperators toremote-manually closethisvalvefromtheControlRoomatanappropriate timefollowing receiptofaCIS(e.g.,following immediate actionsofE-0).Theadditionofthisprocedure stepwillbeaccomplished byAugust15,1990. | | RGaE has also discovered several additional typographical errors and needed clarifications associated with Technical Specification Table 3.6-1.However, these anomalies do not affect the safe operation of Ginna and will be corrected in the upcoming Amendment request to remove this table from the Technical Specifications and reference the UFSAR (anticipated submittal date October 1990).Very truly yours, Robert Mecredy Division Manager Nuclear Production MDFN147 xc: Mr.Allen R.Johnson (Mail Stop 14D1)Project Directorate'-3 Washington, D.C.20555 Ms.Donna Ross New York State Energy Office U.S.Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident inspector}} |
| RGaEhasalsodiscovered severaladditional typographical errorsandneededclarifications associated withTechnical Specification Table3.6-1.However,theseanomalies donotaffectthesafeoperation ofGinnaandwillbecorrected intheupcomingAmendment requesttoremovethistablefromtheTechnical Specifications andreference theUFSAR(anticipated submittal dateOctober1990).Verytrulyyours,RobertMecredyDivisionManagerNuclearProduction MDFN147xc:Mr.AllenR.Johnson(MailStop14D1)ProjectDirectorate'-3 Washington, D.C.20555Ms.DonnaRossNewYorkStateEnergyOfficeU.S.NuclearRegulatory Commission Region1475Allendale RoadKingofPrussia,PA19406GinnaSeniorResidentinspector}}
| |
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ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
ML17265A5111999-01-11011 January 1999
Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME Bp&V Code for ISI Program.Relief Requests 37,38 & 39 Encl
ML17265A5101999-01-11011 January 1999
Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl
1999-09-30
[Table view]
Category:UTILITY TO NRC
MONTHYEARML20059L4481990-09-18018 September 1990
Forwards Responses to Rer 900709-13 Team Visit Findings, Per .Responses Withheld (Ref 10CFR73.21)
ML17309A4491990-09-13013 September 1990
Forwards Slides Presented by Facility & Westinghouse to NRC in 900724 Meeting.Encl Withheld
ML17262A1331990-09-11011 September 1990
Responds to Violations & Several Unresolved Items Noted in SSFI Rept 50-244/89-81.Update of Appropriate Unresolved Items Encl.Specific Actions Re All NRC Unresolved Items Being Tracked to Completion
ML17261B1511990-08-29029 August 1990
Forwards Semiannual Radioactive Effluent Release Rept Jan- June 1990 & Rev 3 to Process Control Program for Ginna Station, Per Tech Specs 6.9.1.4 & 6.16,respectively
ML17261B1481990-08-28028 August 1990
Lists Understanding of Issues Util Planning to Address Re Containment Integrity,Per 900718 Telcon.Any Concerns or Action Items Different from Listed Submittal Should Be Provided to Util Prior to NRC 900905 & 06 Visit to Plant
IR 05000244/19880261990-08-20020 August 1990
Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented
ML17261B1501990-08-20020 August 1990
Submits Update on Util Plans to Provide Appropriate Ventilation to Intermediate Bldg Clean Side,Per Violation Noted in Insp Rept 50-244/88-26.Extensive Program to Manually Close All External Openings Implemented
ML17261B1371990-08-17017 August 1990
Forwards Rev 0 to, Inservice Insp Rept for Third Interval (1990-1999),First Period,First Outage (1990) at Re Ginna Nuclear Power Plant.
ML17309A4481990-08-16016 August 1990
Responds to NRC 900717 Ltr Re Violations Noted in Insp Rept 50-244/90-80.Corrective Actions:Westinghouse Drawing E-2508 Approved & Issued to Central Records
ML17261B1351990-08-15015 August 1990
Provides Info & Assessment on Integrity of Connection of Containment Bldg to Foundation,Per 900808 Telcon.Util Believes That Existing Condition Not Safety Concern & No Reason Exists to Suspect Joint Will Not Perform Function
ML17261B1361990-08-14014 August 1990
Responds to Commitment Tracking Concerns Noted in Insp Rept 50-244/90-09 & Planned Corrective Actions.Util Confirms Commitments Dates for Implementation of Effective Shelf Life Program & Comprehensive Preventive Maint Program for Parts
ML17261B1331990-08-13013 August 1990
Recommits to Performing Enhanced primary-to-secondary Leak Rate Monitoring,Per NRC Bulletin 88-002
ML17261B1261990-07-30030 July 1990
Clarifies Commitment Made in 900316 RO Re Restoring Inoperable Fire Damper I-411-21-P.Util Plans to Design Removable Track Which Will Allow Charcoal Drawer to Be Manipulated.Definitive Schedule Will Be Provided in 60 Days
ML20055H9291990-07-23023 July 1990
Forwards Revised Page 6:4 of Plant Security Plan.Page Withheld Per 10CFR73.21
ML17261B1091990-07-20020 July 1990
Advises That Structural Evaluations of Containment Sys Being Performed in Response to Containment Integrity Insp
ML17309A4471990-07-17017 July 1990
Forwards Decommissioning Rept, for Plant,Per 10CFR50.33(k) & 50.75(b)
ML17261B0881990-07-13013 July 1990
Provides Update to Util 860616 Ltr Re Implementation of NUREG-0737,Item 6.2,Suppl 1, Emergency Response Capability.
ML17261B0921990-07-11011 July 1990
Responds to 900709 Request for Addl Info Re Containment Integrity Insp.Util Will Provide Preliminary Results by 900716 Re Where Groundwater Entering Annular Access Area. Meeting Proposed for Wk of 900723
ML20044B0481990-07-10010 July 1990
Discusses Review of Station Blackout Documentation,Per 10CFR50.63.Licensee Will Complete Enhancements to Station Blackout Documentation Identified in Attachment 1 as Indicated & Other Items Will Be Completed within 2 Yrs
ML17309A4461990-07-0909 July 1990
Responds to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions
ML17261B0931990-07-0909 July 1990
Responds to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Util Completed All Actions Requested in Generic Ltr & Will Retain Documentation Verification for Min of 2 Yrs
ML17261B0901990-07-0909 July 1990
Advises That Licensee Will Install Containment Isolation Signal Going to Valve AOV 745 by End of 1992 Refueling Outage,Per Util 900608 Ltr Notifying of Condition Outside Design Basis of Plant Under 10CFR50.72 Criterion
ML17250B2151990-06-29029 June 1990
Forwards Application for Amend to License DPR-18, Reformatting Auxiliary Electrical Sys Tech Specs & Action Statements for Offsite & Onsite Power Sources Available for Plant Auxiliaries
ML17261B0851990-06-28028 June 1990
Forwards LER 89-016-02 Re 891117 Failure of Safety Injection Block/Unblock Switch Which Could Render Both Trains of Safety Injection Sys Inoperable.Also Reported Per Part 21
ML17250B1991990-06-0808 June 1990
Responds to NRC 900509 Ltr Re Violations Noted in Insp Rept 50-244/89-81.Corrective Actions:Improved Battery Load Profile Developed Incorporating Calculational Improvements Contained in Current Industry Std IEEE 485-1983
ML17250B1951990-06-0505 June 1990
Responds to Generic Ltr 89-08, Erosion/Corrosion Induced Pipe Wall Thinning. Util Developed Erosion/Corrosion Program for Single & Two Phase Sys Consistent W/Requirements of NUREG-1344 & NUMARC 870611 Rept
ML17250B1851990-06-0101 June 1990
Forwards Application for Amend to License DPR-18,providing Guidance for Action Statements Associated W/Power Distribution Limit Specs
ML17261B0691990-06-0101 June 1990
Discusses Testing Frequency for Insp of Incore Neutron Monitoring Sys Thimble Tubes,Per NRC Bulletin 88-009.Thimble Tube Indicating Greatest Wear Recently Repositioned in Effort to Minimize Future Wear
ML17250B1811990-05-31031 May 1990
Forwards Addl Info Re Response to Notice of Violation from Insp Rept 50-244/90-04.Info Withheld (Ref 10CFR73.21)
ML17250B1801990-05-29029 May 1990
Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Util Does Not Need to Develop Enhanced Surveillance Program to Monitor Currently Installed Transmitter Based Upon Limited Installed Quantity
ML17250B1841990-05-29029 May 1990
Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High Hardness Type 410 Stainless Steel (SS) Internal Pre-Load Bolting.... Valves Disassembled & 410 SS Studs Removed & Visually & Liquid Penetrant Examined
ML17261B0961990-05-29029 May 1990
Responds to NRC 900427 Ltr Re Violations Noted in Insp Rept 50-244/90-07.Corrective Actions:Worker Records Immediately Corrected Indicating That Worker Received No Significant Exposure for Time While Error Occurred
ML17250B1721990-05-18018 May 1990
Forwards Rev 1 to Summary Exam Rept for 1990 Steam Generator Eddy Current Insp. Rept Summarizes Observations & Corrective Actions Resulting from Insps Performed During 1990 Outage
ML20042G8931990-05-0808 May 1990
Forwards Revised Emergency Operating Procedures,Including Rev 3 to AP-RHR.2,Rev 1 to ES-0.3 & Rev 11 to ES-1.3
ML17261B0611990-04-26026 April 1990
Advises of Changes to 890201 Commitments Made Under Programmed Enhancement Response to Generic Ltr 88-17.Changes Will Not Reduce Capability to Operate Safely in Reduced Inventory Condition or Scope of Programmed Enhancements
ML17261B0591990-04-17017 April 1990
Advises That No Interlocks Required for RHR motor-operated Valves 701 & 720 Based on Present Arrangement.Listed Failures Would Have to Occur in Order for Potential Overpressurization of RHR Sys to Occur
ML17261B0461990-04-12012 April 1990
Responds to Issues Discussed During 900323 Telcon Re Inservice Testing Program Status & Relief Request.Current Test Methodology for Seat Leakage Acceptable Based on Application of Direct Measurement Sys
ML17262A1431990-04-12012 April 1990
Responds to NRC 900202 Ltr Re Weaknesses Noted in Insp Rept 50-244/89-80.Corrective Actions:Procedure ES-0.3 Modified to Provide Guidance for Rapid Cooldown & Depressurization W/ & W/O Reactor & Vessel Instrumentation Sys
ML17250B1451990-04-0606 April 1990
Discusses Impact of SER Issuance for Inservice Insp & Inservice Testing Programs & Advises That Timing Will Not Affect Util Implementation Plans for Programs,Except as Listed
ML17250B1441990-04-0505 April 1990
Responds to NRC 900309 Ltr Re Violations Noted in Insp Rept 50-244/90-02.Corrective Actions:Positive Indicator Will Be Installed for All Check Valves to Enhance Visible Positive Position Verification Ability & to Avoid Confusion
ML17261B0401990-03-30030 March 1990
Advises That Final Results of Station Blackout Documentation Review Will Be Submitted to NRC on or About 900501
ML17261B0321990-03-28028 March 1990
Forwards Annual Rept of ECCS Model Revs as Applicable to Facility,Per 10CFR50.46.Mods to Model for Small Break LOCA Do Not Affect Calculated Peak Clad Temp
ML17261B0331990-03-28028 March 1990
Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date
ML17261B0231990-03-26026 March 1990
Responds to NRC 890222 Ltr Re Violations Noted in Insp Rept 50-244/89-17.Corrective Actions:Personnel Verified Safety Injection Block/Unblock Switch in Proper Position & Operator Procedure 0-1.1 Changed as Indicated
ML17261B0281990-03-23023 March 1990
Forwards Addl Info Re Proposed Tech Spec Amend Concerning Use of Reconstituted Fuel,Per 900315 Telcon.During Fuel Assembly Reconstitution,Failed Fuel Rods Will Be Placed W/ Filler Rods
ML17261B0221990-03-22022 March 1990
Provides Revised Test Schedule for motor-operated Valve Diagnostic Test Program,Per IE Bulletin 85-003.NRC Notification of Changes to Valve Operability Program Required by Generic Ltr 89-10,dtd 890828
ML17261B0411990-03-20020 March 1990
Forwards Summary of Onsite Property Damage Coverage Currently in Force at Plant,Per 10CFR50.54(w)(4)
ML17261B0151990-03-19019 March 1990
Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Plants (USI A-47). Overfill Protection Provided Through Trip Bistables in Reactor Protection Racks Powered from 120-volt Instrument Buses
ML17261B0101990-03-15015 March 1990
Forwards Application for Amend to License DPR-18,allowing Use of Reconstituted Fuel Assemblies
ML20012D2831990-03-14014 March 1990
Responds to Issues Discussed During 900307 Telcon W/Util & Eg&G Re Inservice Testing Program Status & Relief Request. Valves 5960A & 5960B Will Be Disassembled to Verify Forward Flow.Relief Requests PR-8 & VR-25 Encl
1990-09-18
[Table view] |
Text
<<ACCELERATED DISTRIBUTION DEMONSTRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)ACCESSION NBR:9007170340 DOC.DATE: 90/07/09 NOTARIZED:
NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION MECREDY,R.
Rochester Gas 6 Electric Corp.RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.
Project Directorate I-3 DOCKET 05000244
SUBJECT:
Discusses design basis of plant re AOV 745,CCWI return valve from excess letdown HX.DISTRIBUTION CODE: AOOID COPIES RECEIVED:LTR g ENCL~SIZE:~TITLE: OR Submittal:
General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
05000244 RECIPIENT ID CODE/NAME PD1-3 LA JOHNSON,A COPIES LTTR ENCL RECIPIENT ID CODE/NAME PD1-3 PD COPIES LTTR ENCL INTERNAL: NRR/DET/ECMB 9H NRR/DST 8E2 NRR/DST/SICB 7E NUDOCS-ABSTRACT OGC/HDS2 RES/DSIR/EIB EXTERNAL: LPDR NSIC NRR/DOEA/OTSB11 1 NRR/DST/SELB 8D 1 NRR/DST/SRXB 8E 1 ,OC/-LFMB E 1 REG FKE~>>=01~~
NRC PDR NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT.20079)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!TOTAL NUMBER OF COPIES REQUIRED: LTTR 21 ENCL y
>~f f//51/II I//I'I/t/AND ESgPII///////SS////SF'/F//F 5TAYE ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y.14649-0001 July 9, 1990 TELEPHONE AREA COOE Tld 546.2700 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Allen R.Johnson Project Directorate I-3 Washington, D.C.20555
Subject:
Rochester Gas and Electric Corporation R.E.Ginna Nuclear Power Plant Docket No.50-244
Dear Mr.Johnson,
On June 8, 1990, RG&E notified the NRC under the criterion of 10CFR550.72, of a condition outside the design basis of the plant associated with AOV 745, the component cooling water isolation return valve from the excess letdown heat exchanger.
This notification was made as a result of information contained in Table 3.6-1 of the Ginna Technical Specifications and UFSAR Table 6.2-13, which indicated that AOV 745 should receive a Containment Isolation Signal (CIS).However, elementary wiring drawings did not show this valve as receiving such a signal.Subsequent to this notification, RG&E determined that information contained on Technical Specification Table 3.6-1 for MOV 749A and MOV 749B (component cooling water isolation valves to the reactor coolant pumps)was also potentially misleading based on the fact that these valves do not receive a CIS but are shown with a maximum isolation time of 60 seconds.Consequently, RG&E has reviewed the background information available with respect to AOV 745, MOV 749A, and MOV 749B.All three valves were reviewed by the NRC during implementation of the TMI Lessons Learned recommendations and the Systematic Evaluation Program.Penetration 124a was evaluated in these two programs assuming that it was nonessential, and that AOV 745 was a normally open valve which did not receive a CIS but was closed post-LOCA.
Meanwhile, Penetrations 127 (MOV 749A)and 128 (MOV 749B)were considered essential penetrations with normally opened valves that also did not receive a CIS but remain open post-LOCA.
In a November 19, 1979 letter to the NRC discussing TMI Lessons Learnep topics, RG&E stated that Penetration 124a was"nonesbential" while Penetrations 127 and 128 were"essential".
In previous correspondence from RG&E on this subject (dated June 22, 1979), a table was provided which listed all valves receiving a CIS;neither AOV 745, MOV 749A, nor MOV 749B were listed on this table.9007l70340 900709 PDR AGOCK 05000244 ppl leap
/e p 4 4I In Table 1 of the Final Evaluation Report for SEP Topic VI-4, dated April 12, 1982, it is indicated that the NRC assumed that AOV 745 was normally open, but closed post-LOCA.
However, there is no information listed beneath the"Actuation" column for this valve which would indicate how the valve was expected to close.This table also shows MOV 749A and MOV 749B as normally open valves which remain open post-LOCA.
These valves are shown as"remote manual" under"Actuation".
In addition, drawings and the original Ginna FSAR as used by the NRC during the SEP did not show that there was a CIS to these three valves.Therefore, the NRC had evaluated AOV 745, MOV 749A and MOV 749B in their current configurations during both the TMI Lessons Learned and Systematic Evaluation Programs, and found them to be acceptable.
Since there was no conflicting supporting design basis information (e.g., UFSAR, elementary drawings)showing that MOV 749A and MOV 749B receive a CIS, these valves were not declared inoperable when it was determined that they did not receive a CIS, as had been done with AOV 745.Based on our review of this situation, RG&E interprets the 60 second maximum isolation time as listed in Technical Specification Table 3.6-1 to be valve closure time following a demand to close.This interpretation is consistent with Standard Review Plan 6.2.4, section II (n)and has been transmitted to the Ginna Station operation staff.'Therefore, no changes are required with respect to MOV 749A and MOV 749B.Penetration 124a is a closed system inside containment which is seismic and missile-protected.
The current GDC 57 only requires a manual or remote-manual valve outside containment for this configuration.
The safety function (i.e., containment isolation) of this penetration is met in the current configuration, even with AOV 745 normally open.RG&E has thus concluded that AOV 745 does not require a CIS to be considered operable since the containment isolation function associated with Penetration 124a is accomplished by the closed system inside containment and AOV 745 (without a CIS)outside containment.
Therefore, the criteria of 10CFR550.73 are not applicable to the discovery that AOV 745 does not receive a CIS.However, because our response to item 2.1.4.3 in the above cited November 19, 1979 letter indicated that we would provide a CIS to valves in non-essential penetrations, even though such a signal goes beyond the requirements of GDC 57, RG&E commits to install such a signal to AOV 745 by the end of the 1992 refueling outage.In the meantime, in order to provide another level of assurance of containment isolation, RG&E is planning to modify the Emergency Procedures to direct the operators to remote-manually close this valve from the Control Room at an appropriate time following receipt of a CIS (e.g., following immediate actions of E-0).The addition of this procedure step will be accomplished by August 15, 1990.
RGaE has also discovered several additional typographical errors and needed clarifications associated with Technical Specification Table 3.6-1.However, these anomalies do not affect the safe operation of Ginna and will be corrected in the upcoming Amendment request to remove this table from the Technical Specifications and reference the UFSAR (anticipated submittal date October 1990).Very truly yours, Robert Mecredy Division Manager Nuclear Production MDFN147 xc: Mr.Allen R.Johnson (Mail Stop 14D1)Project Directorate'-3 Washington, D.C.20555 Ms.Donna Ross New York State Energy Office U.S.Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident inspector
