Similar Documents at Ginna |
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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] 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
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REGULATORY ~ ORNATION DISTRIBUTION SY'R (RIBS)
ACCESSION NBR:8012220329 DOC,DATE: 80/12/15 NOTARIZED: YES DOCKET FACIL:50-244 BYNAME Robert Emmet Ginna Nuclear Planti Unit 1~ Rochester G 05000200 AUTH AUTHOR AFFILIATION MAIERg J ~ Es Rochester Gas L Electric Corp.
RECIP ~ NAME RECIPIENT AFFILIATION CRUTCHF IELD i D ~ Operating Reactors Branch 5
SUBJECT:
,.For wards responses to NRC 801031 ltr re clarification of TMI action plan, requirements. Util intends to comply w/NRC DISTRIBUTION requirements L implementation dates,Execptions are discussed in encl'eviations will be communicated.
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' Zuu/I' ///////////I ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649 JOHN E. MA IER TCLCPHONC VICE PRESIDENT AtEC* CEDOC 7333 546.2700 December 15, 1980 C3
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A7 Director of Nuclear Reactor Regulation ri Attention: Mr. Dennis M. Crutchfield, Chief Pl cn Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commissi'on Washington, DC 50555
Subject:
Clarification of TMI Action Plan Requirements (NUREG 0737)
R. E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Crutchfield:
Darrel Eisenhut's October 31, 1980 letter provided clarifica-tion of TMI Action Plan Requirements and placed in one document all TMI-related items approved for implementation by the Commission at this time. The letter also provided implementation dates for each item and requested that we confirm that the implementation dates would be met.
In general, we intend to comply with the requirements and implementation dates as set forth in Mr. Eisenhut's letter.
Exceptions to the requirements or dates are discussed in Attachment A. Because we have had relatively of the requirements of Mr. Eisenhut's letter, we may inform you of little time to analyze some additional deviations at a later date. Our commitment to the dates in Mr. Eisenhptls October 31, 1980 letter supersedes all previous commitments which indicated Action Plan item completion dates earlier than those in NUREG-0737.
Very truly yours, J n E. Maier Subscribed and sworn to me on this 15th day of December,1980 GARY L'; R ISS ARY PUBLIC, State of N. Y. Monroe Co. \
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ATTACHMENT A Rochester Gas and Electric Corporation Response to Clarification of TMI Action Plan Requirements December 15, 1980
The following comments and commitments are stated in response to NUREG 0737, Clarification of TMI Action Plan Requirements. Item numbers correspond to those in the NUREG. Only those items for which RG&E has comments or proposes deviations from the schedule or requirements are listed.
1.A.1.1 Shift Technical Advisor RG&E's most recent. response concerning Shift Technical Advisors is contained in a letter dated August 5, 1980 from L. D. White, Jr. to Mr. Dennis M. Crutchfield, USNRC. We have implemented a program to provide trained STAs by January 1, 1981, however, because the program was redirected 'in response to a letter dated July 7, 1980 from Mr. Crutchfield, the individuals will not have one full year of training, a potential requirement implied in a letter dated September 13, 1979 from Mr. Darrell Eisenhut.
A complete description of our STA program will be provided by January 1, 1981.
Shift Manning By letter dated October 13, 1980 from L. D. White, Jr.
to Mr. Dennis M~ Crutchfield, USNRC, RGM responded to shift staffing criteria and guidelines for scheduling overtime for licensed operators. The commitments pro-vided in that letter, and proposed alternatives to some of the Staff overtime guidelines, remain unchanged.
By January 1, 1981 we will propose a policy to limit overtime work of people in addition to licensed operators who perform safety related work.
I.C.1 Guidance for the Evaluation and Development of Procedures for Transients and Accidents The Westinghouse Owners Group will submit by January 1, 1981, a detailed description of our program to comply with the requirements of Item I.C.l. The program will identify previous Owners Group submittals to the NRC, which we believe will comprise the bulk of the response.
Additional effort required to obtain full compliance with this item (with proposed schedules for completion) will also be identified, as discussed with the NRC on November 12, 1980.
Guidance on Procedures for Verifying Correct Performance of Operating Activities We intend to meet the i'ntent of'the requirements of this position as systems.
it is applied to safety related equipment and The Staff recognized in NUREG 0737 that licensed operators are qualified to perform return-to-service verification of equipment important to safety and is investigating the level of qualification for other operators to perform these functions. We intend to use individuals knowledgeable of the systems involved but not necessarily licensed operators.
I I' Several of the Action Plan items now require a human engineering review. In some cases equipment has already been procured and installed to meet an NRC deadline.
We do not intend to redesign or reinstall- equipment except, as may be required after review of the forthcoming NRC guidelines (NUREG 0700).
Reactor Coolant System Vents Requirements have been changed since we designed our system and submitted it for NRC review nearly one year ago. A good faith effort was made to meet the NRC's earlier deadline for installation of January 1, 1981 and our system is now operational. Good engineering judgment was used in the design of the system and we believe we meet all of the requirements, including those recently imposed. Because the system is installed we do not plan to make modifications to it. Tentative approval was given to our system in a letter dated February 29, 1980 from Dennis L. Ziemann, USNRC, to L. D. White, Jr. contingent upon our verification that the system met. certain requirements. That verification was provided by a letter dated June 2, 1980 from Mr. White to Mr. Ziemann. Please confirm that our installed system is acceptable.
Design Review of Plant Shielding and Environmental QuaJ.ification of Equipment for Spaces/Systems Which May Be Used in Post-Accident Operations Modifications to the plant necessary for post.-accident operations and committed to be performed. in our letter dated December 28, 1979 from L.'. White, Jr. to Mr. Dennis Ziemann, USNRC will be completed by the required date of January 1, 1982. Our December 28, 1979 response also said we would investigate a modification to the plant radwaste equipment procedures and controls.
We have performed that investigation and determined is prudent to duplicate some controls on the radwaste it panel at a remote location. As engineering has progressed and cost estimates have been developed, it is apparent that the cost of moving some controls is nearly as great as the cost of total replacement of the existing radwaste control panel with a computer-based digital control system. A major advantage to implementing the computer-based system would be that control of the Radwaste System could be from several remote locations.
This would eliminate the need for an operator to be stationed in the area of the existing panel, even during normal operation, and thus eliminate radiation exposure. The computer-based system would therefore contribute to reducing the radiation exposure of the plant operators and contribute to our ALARA effort.
Since the design of a computer-based control system
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involves current state-of-the-art technology and system development, the total estimated project. length is longer than that of the remote duplication effort. For this reason RG6E will not be able to complete this modification by January 1, 1982. Our present, schedule, however, projects completion by end of the Spring 1982 Refueling Outage.
The design of our shielding modifications is ongoin gq h owever not 'complete. Thus all design details will not be available by January 1, 1981.
aey Performance Testing of BWR and PWR Relief and Safet Valves As a sponsor of the EPRI PWR Safety and Relief Valve Test Program, Rochester Gas and Electric Corporation intends to comply with the requirements of NUREG 0578, Item 2.1.2. By letter dated December 15 I 1980 R. C. Youngdahl of Consumers Power Company has provided the current PWR Utilities'ositions on NUREG 0737, Item II.D.1 clarifications. Briefly those positio si ions are:
A. Safety and Relief Valves and Piping the EPRI "Program Plan for Performance Testing of PWR Safety and Relief Valves", Revision 1, dated July 1, 1980, does provide a program that satisfies the NRC requirements. Discussion with the NRC staff and their consultants is resolving specific detailed issues.
B. Block Valves - The EPRI Program has not formally included the testing of block valves. However, a small number of block valves have been tested at the Marshall Steam Station Test Facility. The PWR Utilities and EPRI cannot provide a detailed block valve test program until results of the Wyle and CE relief valve tests are available. Therefore, a block valve test program will not be provided before July, 1981. The PWR Utilities and EPRI believe that the proper operation of the TMI-2 and Crystal, River block valves and other operational experience, plus knowledge of the Marshall tests, support a less hurried and more rational approach to block valve testing.
C. ATWS Testing PWR Utilities will not support additional efforts for ATWS valve testing until regulatory issues are resolved. The major safety and relief valve test facility (CE) is nearing completion and some measures were taken to provide additional test capability beyond the current, program requirements. Results from the current
program are likely to provide most of the information necessary to address ATWS events (i;e., relief capability at high pressures).
II.E.4.2 Containment Isolation Dependability The purge and vent. system at Ginna consists of four 48 inch isolation valves. The Staff s interim position on containment purging (now called Position 6) was im-plemented by our December 14, 1979 and May 29, 1980 letters. During a recent review of Position 6, it was postulated 'that two 6 inch valves on our containment depressurization line may be interpreted as falling under Position 6 requirements. These valves are not used for containment purge and vent operations but are used periodically to equalize pressure between inside and outside containment. If these valves are not allowed to be opened during normal operation, unacceptable containment pressures may result. We are reviewing the operability of these valves under accident conditions.
In addition we are reviewing the effects on the containment and the containment, isolation pressure setpoint.
line is not used periodically to minimize containment if this pressure fluctuation. We concur with the basic requirement that the containment pressure setpoint that initiates containment isolation of nonessential penetrations should be the minimum setpoint compatible with normal operating conditions. However, pending additional evaluation, we cannot. concur with the clarification statement that 1 PSI above normal operation pressure is an appropriate minimum pressure setpoint. This value may .not be sufficient to account for instrumentation drift and accuracy as well as abnormal operation scenarios which may increase containment pressure above normal but for which isolation of nonessential penetrations is not desirable. It should be noted that the same pressure setpoint that actuates containment isolation also actuates safety injection and trips the reactor. By January 1, 1981 we will address how the intent of Position 6 wj.ll be applied to the depressurization valves. I II.F.1.1 Noble Gas Effluent Monitor The possible release paths for noble gases from the plant secondary system have been identified as the main steam safety and relief valves, turbine driven auxiliary feedwater pump (TAFWP) exhaust, and main condenser air ejector. An adjacent steam fane monitor having a detection range of 10 - 10 pCi/cc is proposed on each main steam line upstream of the safety valves and TAFWP supply lines. Any releases from the safeties, TAFWP exhaust, or condenser air-ejector can then be quantified by integrating the flow rate times the steam activity r concentration with respect to time.
Condenser exhaust gases are diluted by a factgr of approximately 100. A low range detector (10 - 10 3 pCi/cc) presently monitors the condenser air ejector exhaust as a means of early detection of primary to secondary leaks. We propose to ipcreasq3the range of this detector to overlap with the 10 10 pCi/cc range of the steam line monitors, thus providing a means for following a qgntinuous increase in steam activity from 10 to 10 pCi/cc.
A l'ow range detector on the TAFWP exhaust is believed unnecessary since the air ejector monitor will be viewing steam activity levels under normal operations.
The TAFWP is not normally in use during power operation when a low range monitor on the steam exhaust might be used to detect steam generator tube leaks. Tube leaks will normally be detected by the air ejector monitor.
A detailed system description including detector models, data recording methods, and calculational procedures will be submitted by February 1, 1981.
Clarification 4.a.(iv) requires a capability to obtain readings at least every 15 minutes during and following an accident. Table II.F.1-1 requires the display to be continuous and recording as equivalent Xe-133. These
-requirements are inconsistent, however, we assume that the Eberline Sping-4's purchased and now on site for monitoring the containment vent and the plant vent are acceptable. We previously documented our intention to use these monitors on December 28, 1979. The Sping-4's are capable of automatically logging activity levels every ten minutes when in alarm or at any time when manually requested by the operator.
II.F.1.3 Containment High Range Radiation Monitor A Victoreen Model 875 High Range Containment. Area Monitor System has been purchased for installation by January 1, 1982. The system is currently being qualified to IEEE-323 and Regulatory Guides 1.97 and 1.89, with test reports expected to be completed by March 1981.
Until those tests are complete, however, we cannot commit that the installed system will meet all of the NRC requirements.
II.F.1.4 Containment Pressure Monitor The Staff position presently calls for "continuous recording" of containment pressure; it is felt that this would result in a waste of paper and unnecessary wear on the recorder mechanism. A system is proposed, however, that will start recording whenever a safety injection or containment isolation signal is present.
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a This proposed system will provide adequate recording of signals.
II.F.1.5 Containment Water Level Monitor The narrow range containment water level instrument.
(Sump A) will provide useful information only for leaks or small fluid line breaks. It will provide no useful information for most. LOCA break sizes. Thus, the narrow range instrumentation should be designed and qualified for those postulated accident. conditions for which operability of the instrumentation can provide meaningful indication. Regulatory Guide 1.89 is an appropriate reference for this instrumentation, however, the qualification envelope is based upon the environment anticipated for small leaks and breaks.
The wide range containment, water level instrument (Sump B) is designed to monitor water depths corresponding to at least 500,000 gallons. Because we could not provide documentation of the qualification of the existing instrumentation, we committed in our Environmental Qualification of Electrical Equipment letter dated October 31, 1980 from John E. Maier to Mr. Darrell G.
Eisenhut to replace this instrumentation. It will be replaced by June 30, 1982.
II.F.2 Instrumentation for Detection of Inadequate Core Cooling A letter dated July 2, 1980 from L. D. White, Jr. to Mr. Dennis M. Crutchfield provided RG&E's position concerning installation of additional instrumentation for the purpose of determining reactor vessel water level. Our position is unchanged since that time. In essence, we do not believe that a reliable, easy-to-interpret, unambiguous indication of vessel level has been demonstrated to exist in any of the devices available for installation at, this'ime. We will continue to encourage and support the development of such a device, however, and when one has been successfully demonstrated to function properly over the range of conditions for which it is intended to operate, and shown to provide useful information to the operator, we will install additional instrumentation. Until that time we do not intend to install a so-called water level device.
Additional information concerning our position is contained in our July 2 letter. No response to our letter has been received. Information received at a recent NRC sponsored seminar on water level devices in Idaho Falls reinforces our belief that much, testing and confirmation work must, be completed before water level devices should be installed in operating reactors.
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Since the TMI accident, improved operator training, revised operating procedures based upon extensive new
.,analyses, and installed subcooling meters have provided additional safety margin for adequate core cooling.
Plant procedures for recognizing and mitigating in-adequate core cooling coupled with existing instrumentation provide for all the mitigating actions available to the plant operator. If the core is determined to be inadequately cooled, knowledge of the specific water level (if something other than froth exists under transient conditions) will not affect the actions to be taken by the operator.
Therefore, continued operation of the R. E. Ginna Nuclear Power Plant with existing instrumentation and without the addition of a reactor vessel water level device will not adversely affect public health and safety.
II.K.2.13 Thermal Mechanical Report -- Effect of High-Pressure Injection on Vessel Integrity for Small-Break Toss-of-Coolant Accident. with No Auxiliary Feedwater To completely address the NRC requirements of detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater, a program will be completed and documented to the NRC by the Westinghouse Owners Group by January 1, 1982. This program will consist of analysis for generic Westinghouse PWR plant groupings.
Following completion of this generic program, additional plant specific analyses, if required, will be provided.
A,schedule for the plant, specific analysis will be determined based on the results of the generic analysis.
II.K.2.17 Potential for Voiding in the Reactor Coolant Systems during Transient The Westinghouse Owners Group is currently addressing the potential for void formation in the Reactor Coolant System (RCS) during natural circulation cooldown condi-tions, as described in W Letter NS-TMA-2298 (T. M.
Anderson, W. to P. S. Check, NRC). We believe the results of this effort will fully address the NRC require-ment for analysis to determine the potential for voiding in the RCS during anticipated transients. A report, describing the results of this effort will be provided to the NRC before January 1, 1982.
iI c-II.K.2.19 Sequential Auxiliary Feedwater Flow Analysis The Transient Analysis Code, LOFTRAN, and the present small break evaluations analysis code, WFLASH, have both undergone benchmarking against, plant information or experimental test facilities. These codes under appropriate conditions have also been compared with each other. The Westinghouse Owners Group will provide on a schedule consistent with the requirement, of Task II.K.2.19, a report addressing the benchmarking of these codes.
II.K.3.1 Installation and Testing of Automatic Power-Operated Relief Valve Isolation System And II.K.3.2 Report on Overall Safety Effect of Power-Operated Relief Valve Isolation System The Westinghouse Owners Group is, in the process of developing a report (including historical valve failure rate data and documentation of actions, taken since the TMI-2 event to decrease the probability of a stuck-open PORV) to address the NRC concerns of Item II.K.3.2.
However, due to the time-consuming processing of data gathering, breakdown, and evaluation, this report is scheduled for submittal to the NRC on March 1, 1981. As required by the NRC, this report will be used to support a decision on the necessity of incorporating an automatic PORV isolation system as specified in Task Action Item II.K.3.1.
II.K.3.5 Automatic Trip of Reactor Coolant Pump During Loss of Coolant Accident The Westinghouse Owners Group resolution of this issue has been to perform analyses using the Westinghouse Small Break Evaluation Model WFLASH to show ample time is available for the operator to trip the reactor coolant.
pumps following certain size small breaks (See WCAP-9584). In addition, the Owners Group is supporting a best estimate study using the NOTRUMP computer code to demonstrate that tripping the reactor coolant pump at the worst trip time after a small break will lead to acceptable results.
For both of these analysis efforts, the Westinghouse Owners Group is performing blind post-test predictions of LOFT experiment L3-6. The input data and model to be used with WFLASH on LOFT L3-6 has been submitted to the Staff on December 1, 1980 (NS-TMA-2348). The information to be used with NOTRUMP on LOFT L3-5 will be submitted prior to performance of the L3-6 test as stated in Westinghouse Owners Group letter OG-45 dated December 3, 1980.
The LOFT prediction from both models will be submitted to the Staff on February 15, 1981 given that the test is performed on schedule. The best estimate study is scheduled for completion by April 1, 1981.
Based on these studies, the Westinghouse Owners Group believes that resolution of this issue will be achieved without any design modifications. In the event that, this is not the case, a schedule will be provided for potential modifications.
II.K.3.30 Revised Small Break LOCA Methods to Show Compliance with 10 CFR 50, Appendix K and II.K.3.31 Plant Specific Calculations to Show Compliance with 10 CFR 50.46.
Our response to these items was provided by a letter dated November 13, 1980 from John E. Maier to Mr. Dennis M. Crutchfield, USNRC.
III.A.2 Improving Licensee Emergency Preparedness Long Term At this time we believe we will be able to comply with the implementation schedule established for this item.
However, we plan to comply with the requirement for a prompt notification system primarily with the installa-tion of sirens. We do not yet have a commitment for supply of the sirens because field work necessary to establish sound levels, siren locations and the number of sirens required is not yet completed. If it becomes necessary to request an extension of the implementation date as this work proceeds, we will notify you promptly.
III.D.3.4 Control Room Habitability Requirements The information requested in Attachment 1 to item III.D.3.4 will not be submitted January 1, 1981 for the reasons given in a letter dated November 24, 1980 from L. D. White, to Mr. Dennis M. Crutchfield, USNRC.
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