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| | issue date = 04/17/1990 | | | issue date = 04/17/1990 |
| | title = 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 | | | title = 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 |
| | author name = MECREDY R C | | | author name = Mecredy R |
| | author affiliation = ROCHESTER GAS & ELECTRIC CORP. | | | author affiliation = ROCHESTER GAS & ELECTRIC CORP. |
| | addressee name = JOHNSON A R | | | addressee name = Johnson A |
| | addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) | | | addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR) |
| | docket = 05000244 | | | docket = 05000244 |
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| =Text= | | =Text= |
| {{#Wiki_filter:ACCELERATED DISTjUBUTION DEMONSHRATION SYSTEMREGULATORY INFORMATION DISTRIBUTION SYSTEM(RIDS)ACCESSION NBR:9005020083 DOC.DATE: | | {{#Wiki_filter:ACCELERATED DISTjUBUTION DEMONSHRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) |
| 90/04/17NOTARIZED: | | ACCESSION NBR:9005020083 DOC.DATE: 90/04/17 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas 6 Electric Corp. |
| NOFACIL:50-244 RobertEmmetGinnaNuclearPlant,Unit1,Rochester GAUTH.NAMEAUTHORAFFILIATION MECREDY,R.C.
| | RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. Project Directorate I-3 |
| Rochester Gas6ElectricCorp.RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. | |
| ProjectDirectorate I-3
| |
|
| |
|
| ==SUBJECT:== | | ==SUBJECT:== |
| Advisesthatnointerlocks arerequiredforRHRvalvesMOV7016MOV720atplant,per NRCSERforSEPTopicV-ll.B.DISTRIBUTION CODE,AEOIDCOPIESRECEIVED,LTR 2ENCL0SIZE.TITLE:Correspondence reSEPTopicsNOTES:License Expdateinaccordance with10CFR2,2.109(9/19/72).
| | Advises that no interlocks are required for RHR valves MOV 701 6 MOV 720 at plant,per NRC SER for SEP Topic V-ll.B. |
| DOCKET05000244S05000244,
| | DISTRIBUTION CODE, AEOID TITLE: Correspondence re COPIES RECEIVED,LTR SEP Topics 2 ENCL 0 SIZE. |
| /RECIPIENT IDCODE/NAME JOHNSON,A INTERNAL: | | S NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244, |
| NUDOCS-ABS~ | | / |
| RE~I-LE.--, | | RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL |
| 0~~EXTERNAL: | | 'D JOHNSON,A INTERNAL: NUDOCS-ABS~ OGC/HDS2 RE~I-LE.--, 0~~ |
| LPDRNSICCOPIESLTTRENCLRECIPIENT IDCODE/NAME OGC/HDS2NRCPDRCOPIESLTTRENCL'DRDSNOTETOALL"RIDS"RECIPIENTS:
| | EXTERNAL: LPDR NRC PDR NSIC R |
| PLEASEHELPUSTOREDUCEWASTElCONTACTTHE,DOCUMENTCONTROLDESK,ROOMPI-37(EXT.20079)TOELIMINATE YOURNAMEFROMDISTRIBUTION LISISFORDOCUMENTS YOUDON'TNEEDIgTOTALNUMBEROFCOPIESREQUIRED:
| | D S |
| LTTR~ENCL9,8 IIEiI4~( | | 9, NOTE TO ALL "RIDS" RECIPIENTS: |
| ,zsizzpsvze,>i,~zrurzi/pzrzizu'~YOTklizsiiriziz ROCHESTER GASANDELECTRICCORPORATION o89EASTAVENUE,ROCHESTER, NK146490001YORKiYKYRApril17,1990YCLKPHONE KDKKCOOKYYO546.2700U.S.NuclearRegulatory Commission DocumentControlDeskAttn:AllenR.JohnsonProjectDirectorate I-3Washington, D.C.20555 | | 8 PLEASE HELP US TO REDUCE WASTEl CONTACT THE,DOCUMENT CONTROL DESK, ROOM PI-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDI TOTAL NUMBER OF COPIES REQUIRED: LTTR ~g ENCL |
| | |
| | II E |
| | i I 4 ~ ( |
| | |
| | ,z sizz psvze ~ YO Tk |
| | >i, YORK zrur zi/ pzrzizu' |
| | ~ lizsiiriziz iYKYR ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N K 14649 0001 April 17, 1990 YCLKPHONE KDKK COOK YYO 546.2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555 |
|
| |
|
| ==Subject:== | | ==Subject:== |
| LackofInterlocks onRHRValvesMOV701and.MOV720R.E.GinnaNuclearPowerPlantDocketNo.50-244
| | Lack of Interlocks on RHR Valves MOV 701 and. MOV 720 R.E. Ginna Nuclear Power Plant Docket No. 50-244 |
| | |
| | ==Dear Mr. Johnson:== |
| | |
| | In the NRC's Safety Evaluation Report for SEP Topic V-11.B, "RHR Interlock Requirements", dated September 29, 1981, it was determined that the lack of interlocks on RHR MOVs 701 and MOVs 720 was acceptable. Several reasons were stated for this, as follows: |
| | ~ The RHR valves are designed such that they are physically unable to open against a differential pressure greater than 500 psi |
| | ~ |
| | By administrative procedure, the RHR valves are key-locked closed, with power removed |
| | ~ A relief valve, set at 600 psig, is available in the RHR system. |
| | In the course of RG&E's implementation of our MOV testing program, using the guidance of Generic Letter 89-10, it has been determined, that MOVs 701 and 720 will open against a differential pressure of greater than 500 psi. |
| | Nonetheless, RGGE considers that, with the present arrangement, no interlocks are required for these valves. In order for a potential overpressurization of the RHR system to occur, the following set of failures would have to occur: (Note: a. schematic of the RHR system is shown on UFSAR Figure 5.4-7.) |
| | : 1. Power would. have to be deliberately restored at the breaker, in violation of procedure A-52.2 |
| | : 2. The key-locked switch would have to be enabled, and the valve opened, in violation of normal alignment procedures |
| | : 3. The interlocked in-board valves (MOV 700 or MOV 721) would have to fail to allow significant leakage. This is considered of low probability, since these valves are in the Ginna Pump and Valve Inservice Testing Program, and are leak-tested on an annual basis usin g p rocedure PT-8. |
| | 00000244 g / ~ g~P~ |
| | PDR ADOCK P~g Pgg Pb.g 2S'8' 'CO |
|
| |
|
| ==DearMr.Johnson:==
| | J |
| IntheNRC'sSafetyEvaluation ReportforSEPTopicV-11.B,"RHRInterlock Requirements",
| | : 4. ~ ~ |
| datedSeptember 29,1981,itwasdetermined thatthelackofinterlocks onRHRMOVs701andMOVs720wasacceptable.
| | Relief valve RV-203, with a capacity of 70,000 lb/hr, would have to fail. This relief valve has been tested annually (RSSP-12) for seat leakage and setpoint verification. |
| Severalreasonswerestatedforthis,asfollows:~TheRHRvalvesaredesignedsuchthattheyarephysically unabletoopenagainstadifferential pressuregreaterthan500psi~Byadministrative procedure, theRHRvalvesarekey-locked closed,withpowerremoved~Areliefvalve,setat600psig,isavailable intheRHRsystem.InthecourseofRG&E'simplementation ofourMOVtestingprogram,usingtheguidanceofGenericLetter89-10,ithasbeendetermined, thatMOVs701and720willopenagainstadifferential pressureofgreaterthan500psi.Nonetheless, RGGEconsiders that,withthepresentarrangement, nointerlocks arerequiredforthesevalves.Inorderforapotential overpressurization oftheRHRsystemtooccur,thefollowing setoffailureswouldhavetooccur:(Note:a.schematic oftheRHRsystemisshownonUFSARFigure5.4-7.)1.Powerwould.havetobedeliberately restoredatthebreaker,inviolation ofprocedure A-52.2Theinterlocked in-boardvalves(MOV700orMOV721)wouldhavetofailtoallowsignificant leakage.Thisisconsidered oflowprobability, sincethesevalvesareintheGinnaPumpandValveInservice TestingProgram,andareleak-tested onanannualbasisusinrocedurePT-8.3.gpPDRADOCK00000244P~gg/~g~P~PggPb.g2S'8''CO2.Thekey-locked switchwouldhavetobeenabled,andthevalveopened,inviolation ofnormalalignment procedures J4.~~ReliefvalveRV-203,withacapacityof70,000lb/hr,wouldhavetofail.Thisreliefvalvehasbeentestedannually(RSSP-12) forseatleakageandsetpointverification.
| | Thus, it is considered that the capability of MOVs 701 and 720 to open against a differential pressure of greater than 500 psi, without the availability of interlocks on these valves, remains acceptable, and does not have a significant implication for public health and safety. |
| Thus,itisconsidered thatthecapability ofMOVs701and720toopenagainstadifferential pressureofgreaterthan500psi,withouttheavailability ofinterlocks onthesevalves,remainsacceptable, anddoesnothaveasignificant implication forpublichealthandsafety.Thisinformation willbereflected inthenextrevisionoftheGinnaStationUFSAR.Verytrulyyours,RobertC.MecredDivisionManagerNuclearProduction GJWN099xc:Mr.AllenR.Johnson(MailStop14D1)ProjectDirectorate I-3Washington, D.C.20555U.S.NuclearRegulatory Commission RegionI475Allendale Road".KingofPrussia,PA19406GinnaSeniorResidentInspector}} | | This information will be reflected in the next revision of the Ginna Station UFSAR. |
| | Very truly yours, Robert C. Mecred Division Manager Nuclear Production GJWN099 xc: Mr. Allen R. Johnson (Mail Stop 14D1) |
| | Project Directorate I-3 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road". |
| | King of Prussia, PA 19406 Ginna Senior Resident Inspector}} |
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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 DISTjUBUTION DEMONSHRATION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9005020083 DOC.DATE: 90/04/17 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas 6 Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. Project Directorate I-3
SUBJECT:
Advises that no interlocks are required for RHR valves MOV 701 6 MOV 720 at plant,per NRC SER for SEP Topic V-ll.B.
DISTRIBUTION CODE, AEOID TITLE: Correspondence re COPIES RECEIVED,LTR SEP Topics 2 ENCL 0 SIZE.
S NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244,
/
RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL
'D JOHNSON,A INTERNAL: NUDOCS-ABS~ OGC/HDS2 RE~I-LE.--, 0~~
EXTERNAL: LPDR NRC PDR NSIC R
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9, NOTE TO ALL "RIDS" RECIPIENTS:
8 PLEASE HELP US TO REDUCE WASTEl CONTACT THE,DOCUMENT CONTROL DESK, ROOM PI-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDI TOTAL NUMBER OF COPIES REQUIRED: LTTR ~g ENCL
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~ lizsiiriziz iYKYR ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N K 14649 0001 April 17, 1990 YCLKPHONE KDKK COOK YYO 546.2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555
Subject:
Lack of Interlocks on RHR Valves MOV 701 and. MOV 720 R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Johnson:
In the NRC's Safety Evaluation Report for SEP Topic V-11.B, "RHR Interlock Requirements", dated September 29, 1981, it was determined that the lack of interlocks on RHR MOVs 701 and MOVs 720 was acceptable. Several reasons were stated for this, as follows:
~ The RHR valves are designed such that they are physically unable to open against a differential pressure greater than 500 psi
~
By administrative procedure, the RHR valves are key-locked closed, with power removed
~ A relief valve, set at 600 psig, is available in the RHR system.
In the course of RG&E's implementation of our MOV testing program, using the guidance of Generic Letter 89-10, it has been determined, that MOVs 701 and 720 will open against a differential pressure of greater than 500 psi.
Nonetheless, RGGE considers that, with the present arrangement, no interlocks are required for these valves. In order for a potential overpressurization of the RHR system to occur, the following set of failures would have to occur: (Note: a. schematic of the RHR system is shown on UFSAR Figure 5.4-7.)
- 1. Power would. have to be deliberately restored at the breaker, in violation of procedure A-52.2
- 2. The key-locked switch would have to be enabled, and the valve opened, in violation of normal alignment procedures
- 3. The interlocked in-board valves (MOV 700 or MOV 721) would have to fail to allow significant leakage. This is considered of low probability, since these valves are in the Ginna Pump and Valve Inservice Testing Program, and are leak-tested on an annual basis usin g p rocedure PT-8.
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Relief valve RV-203, with a capacity of 70,000 lb/hr, would have to fail. This relief valve has been tested annually (RSSP-12) for seat leakage and setpoint verification.
Thus, it is considered that the capability of MOVs 701 and 720 to open against a differential pressure of greater than 500 psi, without the availability of interlocks on these valves, remains acceptable, and does not have a significant implication for public health and safety.
This information will be reflected in the next revision of the Ginna Station UFSAR.
Very truly yours, Robert C. Mecred Division Manager Nuclear Production GJWN099 xc: Mr. Allen R. Johnson (Mail Stop 14D1)
Project Directorate I-3 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road".
King of Prussia, PA 19406 Ginna Senior Resident Inspector
