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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11rl r ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE. JR. TKr.KPHONK VICK PAK5IOKNT ARKA COOK Tld 546.2700 July 27, 1979 Mr. Boyce H. Grier, Director U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 631 Park Avenue King of Prussia, PA 19496
Subject:
IE Bulletin No. 79-13 - Cracking in Feedwater Piping Thirty (30) Day Report on the Examination, Evaluation and Corrective Action at R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Grier:
In accordance with IE Bulletin 79-13 and request from Nuclear Regulatory Commission DOR Staff during a meeting on July 24, 1979 we are submitting with this letter a report of the examination, evaluation and corrective actions associated with the examination of R.E. Ginna Nuclear Power Plant's steam generator feedwater nozzle weld examinations. This transmittal is to be considered a thirty (30) day report and documentation of the presentations made at the July 24, 1979 'meeting in Bethesda, Maryland.
Very truly yours, L. D. White, Jr.
Enclosure xc: U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D.C. 20555 90823o69'Q'r
Response to IE Bulletin No. 79 Cracking in Feedwater Piping and Request by DOR Staff During a July 24, 1979 Meeting Bethesda,, Maryland Thirty (30) Day Report on Examination, Evaluation, and Corrective Action at R. E. Ginna Nuclear Power Plant Docket No. 50-244 Jul 27, 1979 In accordance with the requirements of I.E. Bulletin 79-13, Rochester Gas and Electric examined the feedwater nozzle-to-elbow welds of Ginna Station's two steam generators, on July 9 and 10, 1979. These examinations consisted of both radiographic and ultrasonic inspections. Both the ultrasonic and radiographic examination data revealed small linear inside diameter indications in the elbows consistent with those reported in Bulletin 79-13 for the plants listed.'fter confirming the feedwater'ozzle cracking problem, the examination scope was expanded as required by the Bulletin to include other feedwater piping welds inside containment upstream from the nozzle-to-elbow welds. These examinations consisted of both visual and radiographic inspection of 18 additional welds. This includes the welds between the steam generator nozzles and one weld past the first support.
Included also was the containment feedwater penetration piping welds, which represents the other terminal, ends inside containment.
These examinations did not reveal any linear indications and confirmed the continued structural integrity of these additional 18 welds. Information on design, materials, fabrication and operation requested by Mr. Victor Stello, Jr. in his May 25, 1979 letter was submitted by our Mr. L. D. White, Jr.'s letter of June 18, 1979.
After removal of the'lbows from the feedwater system, liquid penetrant and magnetic particle examinations were performed which revealed some pitting type and cracking indications in the nozzle bore and thermal sleeve machined taper and counterbore.
These indications were found in both nozzles with those in the A nozzle being more pronounced than those in the B nozzle.
The cracks in the A and B elbows were determined to be adjacent to the feedwater nozzle-to-elbow welds, in the elbow base material counterbore-taper intersection approximately 1/2 and 3/4 of an inch, respectively from the center line of the weld. Interpretations of the UT and RT data were made as follows:
A Steam Generator maximum depth of approximately 1/16 inch wall penetration, 300'round the circum-ference of the elbow.
B Steam Generator maximum depth of approximately 3/32 inch wall penetration, 360'round the circum-ference of the elbow.
4 The preliminary metallurgical analysis has confirmed that these cracks are a result of' corrosion fatigue mechanism with maximum depth in the areas'investigated to date as foll'ows:
A Steam Generator two cracks noted of 0.065 inch 'and 0.043 inch in depth.
.. B Steam Generator one crack noted of 0.107 inch in depth.
This analysis corresponds very well with the ultrasonic examination flaw sizing data. Also noted were multiple cracks along the area of the weld prep counterbore and taper approximately 0.015 inch in depth. Most of these cracks originated in the groove of deep machine situations, approximately 0.010 inch deep.
Due to the appearance of the cracks with the blunt crack tip ends and the fact that they are completely filled with corrosion products, it is postulated that these cracks are old and have not grown recently. Attachments A and B cover the results of pre-liminary metallurgical analyses of the A and B steam generator
elbows, respectively. The A steam generator elbow was analyzed by RG&E materials laboratory personnel while the B steam generator elbow was analyzed by Westinghouse R and D personnel. When a final report on the metallurgical analysis of these cracks is available, the Nuclear Regulatory'Commission will be sent a copy.
Based on these analyses, the cause of the base material cracking cannot be definitively known at this time. Potential causes include feedwater chemistry, original heat treatment, thermal fatigue, and synergistic effects of these with normal operating stresses. Feedwater chemistry control has been reviewed and we do not believe that it has contributed to the cracking.
Although corrosion pitting was found in the nozzle bore along with the cracking, oxygen control at Ginna has been very good.
During normal power operation, oxygen concentrations are less than 1.0 ppb while during hot standby conditions, the feedwater oxygen concentration is controlled to approximately 100-300 ppb.
Other mechanisms are being evaluated. A report on the normal operating stresses is provided at Attachment C. As part of our evaluation to defemine the factors involved in the cracking phenomenon, we are participating in a pipe cracking owners group to study and analyze the cause. This work will be performed by Westinghouse guided by a technical advisory committee from the utilities involved and will consist of both analytical and experi-mental approaches. Recommendations for permanent corrective actions will be developed as part of the owners group effort.
The corrective actions taken at Ginna have been to build up the nozzle end preps from schedule 60 to schedule 80 and to replace the carbon steel (P1) elbows with schedule 80 chrome-moly (P4) elbows. The replacement has been performed utilizing qualified repair procedures and very precise preheat and post-veld heat
treatment procedures. All pitting, cracks and surface checking inside the nozzle bore were removed by mechanical means. Any
'reas where minimum wall was encroached were repaired by repair welding in accordance with qualified procedures. These corrective actions are considered a repair and not a. modification to the existing configuration of the plant. The replacement chrome-moly (P4) elbows are manufactured to the same Standard Specification, ASTM A-234, and provide greater assurance for the continued structural integrity of the steam generator feedwater nozzle to elbow connection. Therefore, the repair of the cracks found in
.the, nozzles and replacement of the P1 elbows with a P4 material does not represent an unreviewed safety question and does not require a- change in technical specifications.
After completing the repair program on the replacement of the feedwater piping to steam generator elbows we will do a final radiographic and ultrasonic baseline examination. This baseline examination data will be used when an inservice examination is made on both nozzle-to-elbow welds and base metal adjacent to the welds at our next refueling outage.
As part of a longer-term effort, we have installed thermo-couples on both the A and B steam generator replacement elbows to monitor both feedwater temperature and elbow material temperatures.
These thermocouples are installed as shown on Attachments D and E. Four are located on the ID and OD of the A steam generator elbow in the area of the weld prep counterbore and taper transition.
The B steam generator has thermocouples located in the same position as the A steam generator but in addition has twelve (12) thermocouples located upstream of the nozzle to elbow weld at strategic locations. Temperature data will be given to Westinghouse to contribute more information to the owners group effort. These data will be utilized to complement data from other units that also have similar instrumentation. We understand that Westinghouse will be providing these data to the NRC.on a regular basis.
On July 24, 1979 we met with members of the Nuclear Regulatory Commission DOR and IE Staff in Bethesda, Maryland. During this meeting we discussed Ginna's feedwater system background, the nondestructive examinations performed, the preliminary metallurgical analysis to date, our repair program and the cause investigation into the cracking mechanism. This Report documents the information provided by our personnel during the meeting as requested by NRC staff personnel. Attachment F shows the agenda for this meeting as well as the visual aids that were used during the presentations.
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Attachment A Rochester Gas and Electric Co oration Inter-Ofhce. Correspondence Ginna. Station July 15, 1919
SUBJECT:
Materials Laboratory Analysis of "A" Steam Generator Feedwater Piping Cracks TQo B.A., Snow Contained in this report are the initial metallographic of a specimen taken from the "A" Steam Generator
'bservations Feedwater Nozzle to the elbow weld area following the detection of linear- indications by ultrasonic and radiographic non-destructive examination methods.
Figure 1 is a composite photograph of a macrosection taken from the Feedwater System elbow and a drawing representation of'he nozzle and weld configuration. The two cracks shown on the elbow side of the weld are located from the point of the weld prep counterbore and the transition to the nominal wall thickness of the elbow..
Figure 2, taken looking down on the inside Surface of the elbow, is a photograph of the counterbore transitioning to the elbow nominal thickness. Near the bottom of the photo a distinct line can be seen. It is at this point that the cracks begin. The roughness of .the machined transition surface should also be noted.
Figures 3 and 4 are forty magnification cross-sectional views of the two detected cracks. Note the roughness of the machined surface and the existence of the cracks at the bottom of the machined cuts where maximum stress can be expected. The depth of the deeper crack measured approximately .065 inches. Note also in Figure 3 the presence of a very small crack (approximately
.015 inches) in the upper right corner.
Materials Labor July 15, 1979 Page 2 Figure. 5 is a 500 magnification photomicrograph of'he larger crack.'s tip. This photo clearly shows the mechanism behind these cracks to be corrosion.
Figure 6 is a view of the internal surface of the crack after the specimen was separated. Note the layered appearance which indicates the gradual progression of the crack:
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M; J.. Saporito Welding and NDE Specialist xc:" I.D. White, Jr.
J.E.. Arthur T;R Schuler Approved:
Albert E. Curtis III Welding and NDE Engineer
Nozzle Qf Cracks Ginna Sta S/6 Feedwater Elbow Cracks
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Inside Surface of Elbow '5X Figure- 2.
As Polished 40X Figure 3
5X Natl Etch 40X, Figure ~-
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5X Hital Etch. 500X Figure 5
Cracks Internal Surface n
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"5'igure, 6 30X,
Attachement B Robert Emmett Ginna, Loop B -.
Summary of" Metallurgical Evaluations We have examined the cracked elbow section of. the Loop B feedwater. pipe of the Robert Emmett Ginna Nuclear Station of the Rochester Gas and Electric Corporation using nondestructive inspection techniques, metallography and fractography. ,We also checked the material for chemistry, strength, and ductile-brittle transition behavior. Following are some of the results obtained:
1.. The chemistry of the steel conforms to ASTM specification A 106 Grade B with carbon being 0.28%%d and Mn 0.86X.
2.. The tensile strength was 72 ksi at room temperature and 67 ksi at= 440'F with corresponding yield strengths of 39 ksi and 34 ksi.
3 The ductile-brittle transition temperature was near O'P.
Typical Charpy impact values. were 7 ft-lbs at -50'F, 174 ft-lbs at room temperature and 179-ft-lbs at 440'F.';
Multiple cracks were found at a machined section change of the elbow near the elbow-to-nozzle weld joint. Cracks started from deep machining marks and progressed to various depths in the elbow.
5'. Maximum crack depth occurred at the 'knee'f the tapered elbow section at the 8:35 o'lock position. The depth here was 0.107 inches.
- 6. The cracks were filled with oxide scale, were transgranular in nature, and showed horizontal branching along pearlite bands.
7.. Opened cracks showed a dark, banded oxide structure.
- 8. Energy dispersion analysis of- X-ray showed some surface deposits containing Cl, Na, K and Cu.
- 9. Cleaned fracture surfaces showed a multitude of arrest lines (beach marks) corresponding to cracks running at 90'o the fracture surface.
- 10. Fractographic examination using the scanning electron and trans-mission electron microscopes showed a relatively flat, corroded topography with microstructural features formed by pearlite.
We conclude that cracking was- due. to stress concentrations in the form of a section change and deep machining grooves. A nearby weld most likely added, to-the stress concentration effect by introducing a residual stress in the notches. Progressive cracking then started by stress corrosion or corrosion fatigue.
L.. Albertin Application Metallurgy July; 23, 1979
Attachment C RG&E NOZZLE STRESS EVALUATION (Nozzle/Elbow Juncture)(Ksi)
Old Arrangement Dead Pressure Norm Op Hot Shut Loop lA 195' 4 ..898 4.555 4 ..890 Loop lB 190 4. 898 5". 290 4. 181 New Arrangement Loop lA ..262, 4 ..432 6.225 '.,699.
Loop 1B .256 4..432 7.,19 5.693'EMP HOT STANDBY THERMAL TRANSIENT DEFINITION
( F) 550 Steady State 60 I
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hment E
hment F CRACKING IN FEEDWATER SYSTEM PIPING RGScE/NRC MEETING BETHESDA, MD JULY 24,, 1979 A'ntroduction J; C:. Hutton D.C. Cook Stello letter Westinghouse recommendation IE: 79-13 B. S stem Back round J. C. Hutton design materials fabrication operation stress analysis C; Nondestructive Examination A E. Curtis
'nozzle welds piping welds nozzle ID D., Metallur ical Evaluation A. E.. Curtis S/G 1A S/G 1B characterization E. ~Re air A. E. Curtis design materials fabrication examination F. Cause Investi ation A. E. Curtis owners group instrumentation analytical experimental Revision 0 J. Hutton 7/23/79
Sheet 1, 7/23/79 A.. E. Curtis III Reason for Feedwater System Inspection A. NRC IE Bulletin 79-13 B Westinghouse Recoraaendation.
II. Feedwater Design Information A. Conf iguration Loop A Figure B-12 Loop B Figure B-13 B Materials
- 1. The feedwater ring and thermal sleeve material in ASTM A 106 Grade B.
- 2. Steam Generator nozzles are SA 336 Code Case 1332, Para. 5a-5d later incorporated in SA 508, Class 2.
I
- 3. The feedwater piping is ASTM A-106, Grade C, seamless pipe with ASTM A-234, Grade WPB fillings.
C. Welding
- 1. Nozzle to elbow J-groove, 1/16" land with backing ring CXtà 1st pass fill with SMAW E7018.
- 2. Piping welds 37-1/2' 10'ompound bevel 1/16" land K insert GTAW fillwith E7018.
SMAW
- 3. Heat treatment 1150 +- 25 F
e Sheet 2 7/23/79 A. E., Curtis III III.. Nondestructive Examination A. Nozzle to Elbmr VT, RP, UZ B.. Bulletin Extention Welds VT, RT C. Nozzle bore , PT, MT D.. Under thermal sleeve RT Feedwater Chemistry A., Chemistry History
- 1. Phosphate type mntrol 12/69 10/74
- a. pH Control only emphasis b., Marey/Halstead Ratio 2.8 - 6.0 Range 2.. 71 72 M/H Ratio <2.6
- 3. 73 M/H Ratio ~ 2.1
- 4. Late 73 M/H Ratio 2.3 2.6
- a. pH 8.5 10.6 b., PO4 10 80 ppm
- c. Free OH=0
- d. CI <75ppm
- 5. 10/74 12/77 AVT
'6. 12/77 AVT with Condensate Polishers (Deep Bed)
- a. Excellent Chemistry
- b. Cat. Cond. 0.2 0.3 m mhos B. Oxygen Control
- 1. Normal Operation <5 ppb
- 2. Maximum Excursion <40ppb 1 week
- 3. 7/78 present <1 ppb
Sheet 3 7/23/79 A. E. Curtis III V. Corrective Actions A. Fitting Replacement
- 1. Cr Mo (P4) Material SA-234, NP-ll 2., Schedule 80 Welding Ends on Nozzle
- 3. Repair Procedures VI.. Longterm Actions A. PWR Feedwater Pipe Cracking Owners Group B. Refueling Outage Inservice Inspections of Areas
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