Similar Documents at Ginna |
---|
Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated IR 05000244/19990081999-10-14014 October 1999 Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs IR 05000244/19992011999-09-24024 September 1999 Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 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 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 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 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 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) 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] |
Text
CATEGORY j.REGULA'I~INFORMATION DISTRIBUTIO~YSTEM (RIDS)(ACCESSION NBR:9609270295 DOC.DATE: 96/09/24 NOTARIZED:
NO , DOCKET I FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH.NAME", AUTHOR AFFIL'IATION MEXREDY,R.C.
Rochester Gas S Electric Corp.RECIP.NAME RECIPIENT AFFILIATION 5~VISSINGPG.S.
Project Directorate I-l (PD1-1)(Poet 941001)
SUBJECT:
Informs that scope of of assessment includes review of pressure locking&thermal bindi,ng evaluations that have C been previously performed.
DISTRIBUTION CODE: R056D COPIES RECEIVED:LTR j ENCL!SIZE: TITLE: Generic Ltr 95-07-Pressure Locking&Thermal B1nding of Safety Rela y NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
05000244 g RECIPIENT ID CODE/NAME NRR/DRPE/EATON INTERNAL FILE CENTER 1 EXTERNAL: NO NUDOCS ABSTRACT COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODE/NAME PD1-1 PD NRR/DE/EMEB/B NRC PDR COPIES LTTR-ENCL 1.1 1 1 1 1 D 0 NOTE TO ALL"RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT.415-2083)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!TOTAL NUMBER OF COPIES REQUIRED:.
LTTR 8 ENCL 8 4ND ROCHESTER GAS AND ELECTRIC CORPORATION
~89 EASTAVENUE, ROCHESTER, N.Y Idbf9-0001 AREA CODE716 546-2700 ROBERT C.MECREDY Vice president t Nucteor Operotions September 24, 1996 U.S.Nuclear Regulatory Commission Document Control Desk Attn: Guy S.Vissing Project Directorate I-1 Washington, D.C.20555
Subject:
Request for Additional Information
-Generic Letter 95-07,"Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," R.E.Ginna Nuclear Power Plant (TAC No.M93466)R.E.Ginna Nuclear Power Plant Docket No.50-244
Dear Mr.Vissing:
By letter dated June 18, 1996, a request was made by the NRC for RG&E to provide additional information in support of RG&E's 180 day response to NRC Generic Letter 95-07 which was submitted on February 16, 1996.RG&E is presently conducting a MOV program self assessment utilizing industry resources to determine MOV program completeness.
The scope of this assessment includes review of the pressure locking and thermal binding evaluations that have been previously performed.
As such, the information provided herein reflects changes consistent with current industry practices that have been identified as part of this assessment.
Any additional changes that may be identified by the continuing self assessment process that would affect the conclusions of any pressure locking or thermal binding information that is presented herein will be forwarded as it becomes available.
The following, based on current information, is submitted in response to your requests.I The Licensee's submittal states that these valves (857A,B,C) could experience back leakage from the RCS during the post LOCA injection phase which could pressurize the RHR system to an estimated 614.7 psig.Please discuss whether these valves could experience back leakage from the RCS during normal power operation and become pressurized to a higher pressure..";"L'0038 VSO927O2V5 9eO9a4 PDR ADOCK 05000244 P PDR
~Res onse During normal operations, MOVs 852A and 852B (core deluge)and MOVs 720 and 721 (normal RHR inlet)would remain closed and therefore RCS back leakage and pressurization through the RHR system, and pressurization to 614.7 psig, is not postulated as it is for the post LOCA scenario.The RHR system is protected from over pressurization by a relief valve (203)with a setpoint of 606 psig.Back leakage from the RCS system through the SI system is postulated for normal operation.
This pressure is limited by the SI pump suction relief valve set point of 210 psig (valve 1817).Therefore the postulated pressurization during normal operation is bounded by the pressurization during the post LOCA injection scenario.Re uest 1.b From a review of system diagrams, it appears that these valves (857A,B,C) may be potentially susceptible to thermally induced pressure locking from heat transfer from the RHR system during a design basis event.Please discuss your review of this potential susceptibility.
~Res ense MOVs 857A, 857B and 857C are not subject to thermally induced pressure locking due to heat transfer from the RHR system during a design-basis event due to the following:
The inlets to the subject MOVs are separated by an extremely long run of 6 inch piping (approximately 59 feet)from the RHR heat exchanger outlet tees where the process fluid heat source would exist during RHR recirculation following a large-break loss of coolant accident (LOCA).Since a large dead leg exists to isolate these MOVs from the heat source and since the elevation of the MOVs is below that of the heat source, it is not possible for any heat transfer to occur over that distance.~Under the analyzed case for RHR recirculation following a LOCA, the maximum temperature at the outlet of the RHR heat exchanger is approximately 180'F.This process fluid temperature is not sufficient to cause any appreciable heat transfer through 59 feet of adjacent dead leg piping.Please discuss your basis for using a friction coefficient of 0.5.~Res onse The open friction coefficient of 0.5 results from an open valve'actor of 0.5 and a disc seat angle of 0 degrees.*The open friction coefficient of 0.5 for MOVs 857A, 857B and 857C was derived, as a bounding friction coefficient, from the results of in-situ differential pressure testing of these MOVs.Based on a nominal seat diameter of 4 inches, MOV 857B had an open valve factor of Os356 and MOV 857C had an open valve factor of 0.469.Although MOV 857A had an open valve factor.of 1.18 based on the test data, this data is considered questionable due to anomalies identified during testing and since a torque-thrust cell was not employed during the test.Current methodology uses mean seat diameter (4.5 inches)and accounts for stem rejection force and torque reaction factor.The current methodology yields lower friction coefficients, thereby further enveloping the 0.5 value used in the calculation.
The use of a friction coefficient of 0.5 for all three MOVs is further justified based on the results of static testing of these MOVs performed during the 1996 outage.This testing demonstrated that the unseating thrust requirements have remained similar to previously determined requirements and were extremely low.That is, testing prior to 1994 indicated a maximum unseating thrust of 669 lbs.The most recent tests indicate the unseating thrust requirements range from 586 lbs to 770 lbs.Therefore, no indication of appreciable friction coefficient differences among the MOVs is evident nor has any significant friction coefficient increase over time occurred due to age-related degradation.
It is also noted that RG&E intends to perform additional testing on valve 857A using a torque-thrust cell, to confirm an actual friction coefficient of less than 0.5.The licensee's calculation includes an unsealing load of 669 lbf.The NRC staff believes that, due to the wedging/unwedging mechanism of Anchor-Darling double disk gate valves, this load can be considerably higher.In addition, the measured unsealing load can be different for each valve stroke.Please address these issues and provide the basis for including an unsealing load of 669 lbf.R~es ense The pressure locking and thermal binding study of record was completed in 1994 using MOV Program data available at that time.Prior to 1994, diagnostic test results of MOVs 857A, 857B and'57C indicated that the measured unsealing load was 669 lbs.The results of diagnostic testing performed during the 1996 outage validates this input and establishes that unsealing loads for these MOVs have remained extremely low and relatively constant as follows:
'h MOV Unsealin Thrust 857A 857B 857C 586 lbs.770 lbs.769 lbs.RG&E has included the highest recorded value in the latest operability calculation, Attachment 1.This calculation demonstrates that the unsealing load based on this latest data will not increase the'required thrust to a value higher than the available thrust.Please provide your actuator output capability calculations for these valves and address and provide justification for any deviations from the Limitorque guideline.
~Res ense Operability calculations based on in-situ motor/actuator capability testing are provided as Attachment 1.Present RG&E actuator sizing methodology utilizes a calculational basis and in-situ test data to verify actuator capabilities for operability.
The initial screening process compares the 89-10 program calculated available thrust, based on published actuator data, to the calculated required thrust.If it is determined that the calculated available thrust is less than the required thrust, field testing data is utilized to determine actual available thrust.The field testing consists of the performance of an"in-situ motor/actuator capability test."'his test measures actual thrust for different torque switch settings via a load cell mounted to the top of the valve stem.The testing is performed by incrementally increasing the actuator torque switch setting and measuring actual output thrust.The thrust measured is the thrust at torque switch trip and therefore any inertia developed after the trip is not erroneously included.While higher output thrust may actually be available, testing is discontinued when the required target thrust value has been It is.noted that in previous discussions with the NRC, the in-situ motor/actuator capability test was erroneously referred to as a"partial stall test" by RG&E.The in-situ motor/actuator capability test does not in fact involve motor stall.The test is also not analogous to previous industry performed motor stall testing (dynamometer) where actuator output capabilities were extrapolated from the measured motor output values.The in-situ motor/actuator capability test performed is a direct measurement of the actuator's output at the valve stem in the direction required to overcome pressure locking (i.e.opening direction).
demonstrated.
This thrust data is then adjusted via calculations to account for identified differences in valve/actuator operating characteristics (i.e.stem factors, degraded voltage).This adjustment is based on bounding values for the group of valves being qualified.
For the 857 valves, the test data compares favorably with test data from the 860 valves (unadjusted values of 10,832 lbs.[857]versus 11,016 lbs.[860])
which have a similar configuration and thus demonstrates repeatability of results.In regard to deviations from the Limitorque guidelines it is noted that the operability calculation is not based on the Limitorque actuator capacity methodology.
The one factor that is common however to both the Limitorque methodology and the in-situ motor/actuator capability testing methodology is valve factor.RG&E may use a larger valve factor than the Limitorque recommended value based on the results of in-situ tests performed at Ginna or by other licensees or based on industry data which indicates that greater values are warranted.
When a correctly assumed valve factor is utilized to establish the target differential-pressure thrust requirement and the in-situ motor/actuator capability test demonstrates that sufficient thrust is available above the required thrust, then operability is ensured.Please provide your weak link analysis and calculations for these valves.~Res onse Weak link analysis and calculations are provided as Attachment 2.The licensee's submittal discusses the susceptibility of valves 852A/B, RHR to Reactor Vessel Deluge, to pressure locking and thermal binding and states that analyses of these conditions have been performed.
Please provide these analyses for our review.~Res onse The pressure locking/thermal binding susceptibility review for MOVs 852A and 852B is provided as Attachment 3.From a review of system diagrams, it appears that the following valves may be potentially susceptible to pressure locking: 720, 721, 878A, 878C.Please provide your pressure locking, susceptibility evaluations for these valves, and include associated calculations completed for our review.
~Res onse The pressure locking susceptibility evaluations for MOVs 720 and 721 are provided as Attachment 4.However, these MOVs have no safety related function to open.Since MOVs 878A and 878C are in their safety-related position (closed)with AC power removed, the valves cannot be inadvertently opened and since these valves are not required'to be opened for any design-basis transient response (per Attachment 5), they were not evaluated for susceptibility to pressure locking.In attachment 1 to GL 95-07, the NRC staff requested that licensees include consideration of the potential for gate valves to undergo pressure locking or thermal binding during surveillance testing........
The staff stated that normally open, safety-related power-operated gate valves which are closed for test or surveillance but must return to the open position should be evaluated within the scope of GL 95-07.'lease discuss if valves which meet this criterion were included in your review, and how potential pressure locking or thermal binding concerns were addressed.
~Res onse Valves (listed below)which are normally open and are closed for surveillance testing and are required to open have been reviewed and were found to be acceptable due to one or more of the following reasons: 1~The valve is not susceptible to,pressure lock or thermal binding.2~3.Technical Specifications are followed to ensure one train is available, i.e.the tested valve is returned to its normally open position prior to declaring it operable.The valve actuator has sufficient available thrust to overcome pressure locking/thermal binding.4~Hardware or procedure modifications have been performed to prevent pressure lock or thermal binding.Valve Summar 515/516 These pressurizer power-operated relief valve block valves are double-disc gate valves modified (hole in disc)to preclude pressure locking.'
704A/B When these residual heat removal pump suction valves are tested, Technical Specifications are followed to ensure one train remains operable.813/814 These component cooling water isolation valves for the reactor support coolers have been evaluated as having sufficient available thrust to overcome pressure locking.871A/B When these Safety Injection Pump C discharge valves are tested, Technical Specifications are followed to ensure the discharge path remains operable.1815A/B When these series Safety Injection Pump C suction valves are tested, Technical Specifications are followed to ensure one safety injection pump suction path remains operable.4615/4616 When these Auxiliary Building service water isolation valves are tested, Technical Specifications are followed to ensure one train remains operable.Through review of operational experience feedback, the staff is aware of instances where licensees have completed design or procedural modifications to preclude pressure locking or thermal binding which may have had an adverse impact on plant safety due to incomplete or incorrect evaluation of the potential effects of these modifications.
Please describe evaluations and training for plant personnel that have been conducted for each design or procedural modification completed to address potential pressure locking or thermal binding concerns.R~es ense All design changes are handled in accordance with the Plant Change Record (PCR)process.This process includes both a safety review/evaluation, as applicable, and a training plan.Example of this process is the modification of the 852A and 852B valves.PCR 96-085 modified the actuator thrust capabilities for these MOVs.As part of the modification process safety evaluation (SEV 1072)was performed.
This safety evaluation addresses functional impacts as well as Technical Specification, UFSAR, Regulatory and Accident Mitigation issues.A training plan, in the form of a Training Notification Letter was also issued.
All procedure changes are handled in accordance with the Procedure Change Notice (PCN)process.This process includes a safety review/evaluation, as applicable, and the submittal of changes to training for review.Operations personnel are trained on these procedures in accordance with the operations personnel qualification program.No specific procedure changes to date have been issued to address PL/TB concerns.Very truly yours, Robert C.Mecredy xc: Mr.Guy S.Vissing (Mail Stop 14C7)Project Directorate I-1 Washington, D.C.20555 U.S.Nuclear Regulatory Commission Region'I 475 Allendale Road King of Prussia, PA 19406 US NRC Ginna Senior Resident Inspector P 4~