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
[Table view] |
Text
..4C CE1Z RATED D1PIR1BUTION DEMONSTRATION SYSTEM REGULATORIa{FORMATION DISTRIBUTION S'EM {RIDE)
ACCESSION NBR:8901100076 DOC.DATE: 88/12/27 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas s Electric Corp.
RECIP.NAME RECIPIENT AFFILiATION STAHLE,C. Document Control Branch (Document Control Desk)
SUBJECT:
Forwards info re upcoming 10 yr inservice insp.New inservice insp techniques will be demonstrated on 890110 & 11.
DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR i ENCL ( SIZE:
TITLE: OR Submittal: Inservice Inspection/Testang/Relief from ASME C ode D NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 0 PD1-3 PD 5 5 STAHLE,C 1 1 INTERNAL: AEOD/DSP/TPAB 1 1 ARM/DAF/LFMB 1 0 NRR/DEST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1 NUD BSTRACT 1 1 OGC/HDS2 1 0 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: EGGG ROCKHOLD,H 1 1 LPDR 1 1 NL 007 HEMMING 1 1 NRC PDR 1 1 NSIC 1 1 h
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//dt's//gfli ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.K 14649.0001 December 27, 1988 TKLKPHONK ARKA COOK 7ld 546.2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Mr. Carl Stahle PWR Project Directorate No. 1 Washington, D.C. 20555
Subject:
Demonstration of ISI Technique at Southwest Research Institute R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Stahle:
Enclosed is information concerning the upcoming 10-year ISI of the Ginna Reactor Vessel. This also includes historical information from the 1979 Ginna reactor vessel inspection and the 1986 Point Beach reactor vessel inspection.
The new In-Service Inspection techniques to be employed for the 1989 inspection will be demonstrated at the Southwest Research Institute facilities in San Antonio on January 10 and 11, 1989. The NRC has expressed interest in this topic, and are invited to this demonstration. An agenda for January 10 is included. The 11th will be set aside to discuss any questions arising from the demonstration.
Very truly yours, 8901l00076 88l227 Robert C. credy PDR ADOCK pNU,'"- ',
05000244 ,,
'eneral M
Manager Production
uclear GJWN011 Enclosures xc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector Robert Winters, Region I Bob Hermann, NRR
January 10, 1989 Enclosure 1 Introduction RG&E>> John Smith Description of Reactor Vessel Ultrasonic Examination - SwRI Bill Clayton Review of the Results of the 1979 R.E. Ginna RG&E>> Mike Saporito Reactor Vessel Ultrasonic Examination Review of the Results of the 1986 Point Beach WE Jim Kohlwey/Gary Sherwood Unit 1 Reactor Vessel Ultrasonic Examination Discuss Indication Sizing Techniques Used SwRI Bill Clayton in 1979 R.E. Ginna and 1986 Point Beach Unit 1 Examinations Discussion of Heavy Wall Sizing Techniques NDEC ~
Frank Ammirato Discussion of Nozzle Flaw Sizing Program
- 1) Overview SwRI Bill Clayton
- 2) Mockups SwRI Bill Clayton
- 3) Techniques and Transducers SwRI George Gruber/Glen Light Buffet Lunch Demonstration
- 1) Presentation of Data Format SwRI Bill Clayton
- 2) Description of Demonstration SwRI Bill Clayton
- 3) Preselected Flaw, One or More Flaws to be Selected by NRC
- 4) Scanning of Selected Flaws SwRI Bill Clayton
- 5) Evaluation of Selected Flaws SwRI Bill Clayton Program Results SwRI Bill Clayton Closing Comments RG&E< John Smith
nclosure 2 1989 REACTOR VESSEL EXAMINATIONS The 1989 Ginna Station Reactor Vessel Examination will consist of two phases, a visual examination of the internal surfaces, nozzles and internal components of the reactor vessel and an ultrasonic examination of the vessel welds. These examinations will satisfy the requirements of the Ginna Second Interval Inspection Plan, ASME Section XX (1974 Edition including Addenda thru Summer 1975) and Regulatory Guide 1.150, Rev.l.
The visual examination will utilize the Benthos Mark II Mini-Rover Submersible Camera System', which has been used successfully by Surry and Point Beach for reactor vessel examinations. The examination plan will allow 100% inspection of all the internal surfaces, nozzles and internals. The internal surfaces will be examined by moving the submersible in a vertical motion and then indexing around the vessel circumference until all surfaces have been covered. The inlet and outlet nozzles will be inspected to the extent feasible by driving the submersible into the nozzle and tilting the TU Camera to view the nozzle ID surface and its weld.
Reactor vessel internals will be examined after they have been removed from the vessel, but while still under water in the reactor t
vessel cavity. The entire examination will be performed in accordance with written RG&E Visual Inspection Procedures.
The ultrasonic examination will be performed on the reactor vessel welds and nozzle to piping welds from the inside surface utilizing The Southwest Research Institute (SWRI) Programmed and Remote (PAR) ISI-2 Device and Fast PAR Equipment. In addition, the
.SWRI Enhanced. Data Acquisition System (EDAS), will be utilized for data collection, correlation and analysis of ultrasonic data. EDAS provides color graphics and data analysis and presents the data in easy to interpret displays. The ultrasonic examination will include 1004 examination of all the reactor vessel welds as shown in Attachment 1. The Ginna examination will include 1004 of the weld length even though the ASME Code requires only 5% of the weld length to be examined. Ultrasonic transducers will be utilized that satisfy the requirements of Section XI of the ASME Code.
Described in Attachment 2 are the results of the Reactor Vessel Exam performed during the 1979 (end of the first inspection interval) outage. This exam included both a visual and ultrasonic examination. The visual exam was performed utilizing cameras mounted on poles and on the PAR device. The ultrasonic examination was done utilizing the PAR device and the SWRI Standard Data Acquisition System (SDAS). This examination identified an indication in the "N2B" Inlet Nozzle to Vessel Weld. This indication was detected from the nozzle bore utilizing the 15 degree angle beam longitudinal wave technique. Based on beam spread correction the indication was sized as approximately Ac=
.160" X Lc = 4.52", as described in Figure 3 of Attachment 3. This size indication is allowable per the ASME Code Table IWB-3510.
A Fracture Mechanics Evaluation of the indication was also performed to demonstrate its acceptability and is described in Attachment. 4.
In conjunction with the two previously discussed evaluations, a review of the original fabrication radiographs confirmed the presence of slag at the identified location.
As a result of these evaluations it is believed that this indication is a code allowable thin planar slag inclusion located mid-wall of the vessel to nozzle weld.
During the 1989 examination of the Ginna Reactor Vessel new techniques and technology will be applied to better resolve any indications identified. To accomplish this the following actions are being taken:
- A joint flaw sizing program has been initiated by Wisconsin Electric (WE) and Rochester Gas and Electric
{RGE) utilizing SWRI and EPRI for support. The purpose of the program is to develop optimized techniques for flaw sizing.
- Two full size mockups have been being fabricated; one representative of the inlet/outlet configuration and the other representative of the safety inspection nozzle.
The mockups have built-in defects similar to indications in the RGE/WE nozzles.
- Focused Transducers and Time-of-Flight Techniques are
/
being investigated for more accurately sizing indications. These technologies will be integrated with the Southwest Research Xnstitute EDAS System.
- A demonstration has been scheduled for mid-January 1989.
The purpose of the demonstration will be to compare 1979 flaw sizing techniques with the enhanced techniques developed through this program. Representatives of RG&E and WE will witness this demonstration and representatives of the NRC 'and the, Authorized, Nuclear Insurers will be invited to attend.
More specific information regarding the Nozzle Sizing Program can be found in Attachment, 5.
M.J. Saporito Rev.0 11/28/88
The following is a listing of mechanized ultrasonic emminations of the reactor pressure vessel welds and adjacent piping wells. These examinations willinclude 1/2T base material for vessel welds and 1/4 inch base material for piping welds. Also shown are the anticipated examination angles and the direction of the beam component.
The lower head is forged and has no meridional welds and the shell courses are ring sections with no longitudinal welds. In all cases the goal is to examine 100% of the weld plus 1/2T each side of the weld. Examination of 100% of the weld length is the goal also for the circumferential vessel welds even though 74/S75 Section XI only requires 5%. Interference from other vessel components may limitthe desired examination coverage. Ifthis was the case in previous examinations, it has been noted. A complete discussion of the individual examination area coverage willbe provided in the Qnal report of" the examinations as required by Regulatory Guide 1.150 Rev. 1.
Mech UT examinations willbe performed on the reactor vessel welds and selected reactor coolant piping welds from the inside surface utilizing the PaR ISI-2 Device and SwRI Fast PaR equipment. Examination areas include vessel circumferential, nozzle-to-shell, and nozzle piping welds.
The Mech UT examinations of the RPV willbe performed in accordance with the requirements of the 74/S75 Section XI and Regulatory Guide 1.150, Rev. 1.
a) RPV Shell and Head Welds
- 1) 0-degree longitudinal wave (UTOL) emminations will be performed for detection of laminar reflectors which might acct interpretation of angle-beam results.
- 2) 0-degree longitudinal wave (UTOW) examinations willalso be performed for detection of reflectors in the weld and base material.
- 3) 45-degree and 60-degree shear wave (UT45 and UT60) examinations willbe performed for detection of reflectors in the weld and base material oriented parallel to the weld.
- 4) 45-degree and 60-degree transverse shear wave (UT45T and UT60T) examinations willbe performed for detection of reflectors in the weld and base material oriented transverse to the weld.
- 5) In the case of the RPV, welds, SwRI 50/70 tandem search units willbe used to examine to a depth of approximately 2.25 inches for detection of reflectors in the clad-to-base metal interface area and also'in the volume between the examination surface and the depth of the first Code calibration reflector.
These dual-element tandem search units develop an interactive beam with longitudinal wave propagation and produce an examination with significantly improved signal-to-noise ratio over conventional near-surface techniques.
ATTACHMENT SCOPE OF ULTRASONIC EXAMINATIONS OF THE REACTOR PRESSURE VESSEL WELDS
b) RPV Nozzle Areas i The inlet, outlet, and safety injection nozzle-to-vessel welds willbe examined from the bore utilizing 15-degree (for inlet nozzles), 10 degree (for outlet nozzles) 10-degree (for safety injection nozzle) and 45-degree beams for detection of reflectors in the weld and base material. In addition, UT45T and UT60T examinations willbe performed from the shell inside surface for detection of reflectors oriented transverse to the weld and base material. These transverse examinations willutilize a computer to control the X-Y-Z movements of the PaR device to assure accurate positioning around the nozzle during examinations. 50/70 tandem search units willbe utilized from the bore and shell inside surfaces for detection of reflectors located in the clad-to-base metal interface region and also the volume between the examination surface end the first Code calibration reflector for the purpose of satisfying the requirements in Section XI.
c) Piping Welds Nozzle Pi in Welds
'or the inlet safe end-to-nozzle welds, a UTOL scan willbe used for detection of reflectors which might aQ'ect interpretation of the angle-beam results. UT45 and UT 60 s'cans willbe used for detection of reflectors parallel to the weld from both sides of the weld. A UT45T scan willbe used for detection of reflectors oriented transverse to the weld. The acoustic properties of the inlet elbows preclude lamination from the elbow side; therefore, a UTOW scan willbe performed in addition to the scans identified above..
Limitations are expected around the vessel support lugs, safety injection and inlet nozzles due to the proximity of these components. Other limitations are listed.
I. Circumferential welds Estimated time - (2.5 shifts)
Ring forging-to-lower head weld (RPV-E)
Examination area Angle Beam Component 0- 360 0,45,60,50/70 up/dn 0- 360 0,45T,60T,50/70T av/ccw Lower shell-to-ring forging weld (RPV-D)
Examination area Beam Component 0- 360 Angle',45,60,50/70 up/dn 0-360 0,45T,60T,50/70T cw/ccw Limitations due to proximity of core support lugs 0 from (344.20 - 15.90) CG-1 I
90 from (74.20 - 105.80) CG-2 180 from (164.20 - 195.80) CG-3 270 from (255.25 - 284.75) CG-4 Intermediate shell-to-lower shell weld (RPV-C)
Examination area Angle Beam Component 0- 360 0,45,60,50/70 up/dn 0- 360 0,45T,60T,50/70T cw/cd
D. Upper shell-to-intermediate shell weld (RPV-B)
Examination area Angle Beam Component 0- 360 0,45,60,50/70 up/dn 0- 360 0,45T,60T,50/70T cw/ccw II. Upper shell region area (A) Estimated time - (3.0 Shifts)
A. Flange-to-upper shell weld (RPV-A) from shell Examination area Angle Beam Component 0- 360 0,45,60,50/70 up 0- 360 0,45T,60T,50/70T cw/ccw B. Outlet nozzle-to-shell welds (N1A), (N1B) from sheH Emmination area Angle Beam Component nozzle (0 - 860) 0,45T,GOT,50/70T nv/cnv C. Inlet nozzle-to-shell welds (N2A), (N2B) from shell Examination area Angle Beam Component nozzle (0 - 860) 0,45T,GOT,50/70T cw/ccw D. Safety injection nozzle-to-shell weld (AC-1002), (AC-1003) from shell Examination area ~
Angle Beam Component nozzle (0 - 860) 0,45T,GOT,50/70T cw/cnv III. Upper shell region area (B) Estimated Time - (1.5 shifts)
A. Flange-to-upper shell weld (RPV-A) from seal surface Examination area Angle Beam Component 0- 360 18, 11,4 dn B. Vessel support lugs Examination area Angle Beam Component Vessel support (RPV-VSL-1) 0,45,60,50/70 up/dn Vessel support (RPV-VSL-1) 0,45T,60T,50/70T cw/ccw Vessel support (RPV-VSL-2) 0,45,60,50/70 up/dn Vessel support (RPV-VSL-2) 0,45T,60T,50/70T cw/ccw
IV. Nozzle inner radius, integral extention and nozzle bore Estimated time - (3.5 shifts)
A. Outlet nozzle inner radius section integral extension region and nozzle bore.
Examination area Angle Beam Component Outlet A (N1A-IHS) 10,45,50/70 To Vessel C/L cw/ccw Outlet B (NlB-IRS), 10,45,50/70 To Vessel C/L cw/ccw Outlet A (NlA-IE) 50/70 To Vessel C/L
'utlet B (N1B-IE) 50/70 To Vessel C/L Inlet nozzle inside radius region Examination area Angle Beam Component Inlet A (N2A-IRS) 50/70 cw/cv,v Inlet B (N2B-IRS) 50/70 cw/ccw C. Nozzle-to-shell welds from nozzle bore Examination area Angle Beam Component Inlet A (N2A) 15,45,50/70 To Vessel C/L cw/ccw Inlet B (N2B) 15,45,50/'70 To Vessel C/L cw/ccw D. Safety injection inside radius region and nozzle bore Examination area Angle Beam Component Safety injection A(AC-1003-IHS) 0,10 To Vessel C/L Safety injection B(AC-1002-IHS) 0,10 To Vessel C/L E. Safety injection nozzle integral mtension Examination area Angle Beam Component Safety injection A(AC-1003-IE) 70 cw Safety injection B(AC-1002-IE) 70 cw V. Nozzle-to-piping welds Estimated Time - (3.5 Shifts)
Elbow-to inlet nozzle welds Examination area Angle Beam Component Inlet A (PL-PW-V) 0,45,60 Away from Vessel C/L Inlet B (PL-FiV-VII) 0,45,60 Away from Vessel C/L Inlet A (PL-FW-V) 45RLT cw/cnv Inlet B (PL-FW-VII) 45RLT cw/cd Inlet A (PL-FW-V) 45RL To Vessel C/L Inlet B (PL-FW-VII) 45RL To Vessel C/L
B. Nozzle-to piping welds Examination area Angle Beam Component Outlet A (PL-FW-II) 0,45,60,45T,60 Away from Vessel C/L Outlet A (PL-FW-Il) 0,45,60,45T,GO To Vessel C/L Outlet B (PL-FW-IV) 0,45,60,45T,60 Away from Vessel C/L Outlet B (PL-FW-IV) 0,45,60,45T,60 To Vessel C/L C. Safe end-to-nozzle welds Examination area Angle Beam Component Safety injection A(AC-1003-1) 0,45>45T,60 Away from Vessel C/L Safety injection A(AC-1003-1) 0,45,45T,60 To Vessel C/L Safety injection B(AC-1002-1) 0,45,45T,60 Away from Vessel C/L Safety irjection B(AC-1002-1) 0,45,45T,60 To Vessel C/L D. Piping-to-safe end welds Examination area Angle Beam Component Safety injection A(AC-1003-2) 0,45,45T,60 Away from Vessel C/L Safety injection A(AC-1003-2) 0,45,45T,60 To Vessel C/L Safety injection B(AC-1002-2) 0,45,45T,60 Away from Vessel C/L Safety injection B(AC-1002-2) 0,45,45T,60 To Vessel C/L A1-5
SCHEDULE OF HECHANIZED EXAMINATIONS FOR R ~ E ~ GINNA RPV Days On Day 1 Day 2 Day 3 Day 4 Day 5 Day 6 Day 7 <==Vessel xaeination Areas 2 2 1
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Outlet 8 (N18-IRS,-IE)
Inlet A (N2A-IRS)
Inlet 8 (H28-IRS)
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ATTACHMENT 2 RESULTS OF 1979 REACTOR VESSEL EXAMINATIONS
ATTACHMENT2 INTRODUCTION The 1979 ISI of the R.E. Ginna Reactor Pressure Vessel (RPV) and associated nozzle piping welds was performed in accordance with Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, 1974 Edition with Addenda through Summer 1975.
The SwRI Data Acquisition System (now referred to as the Standard Data Acquisition System or SDAS) was used for recording and processing of ultrasonic test information obtained during mechanized examinations. This system included Sonic Mark II ultrasonic instruments. These instruments were also used for thickness gauging and as an aid in determining the acoustic characteristics and attenuation of the materials. Various brands, sizes and frequencies of ultrasonic transducers were used to perform the examinations.
In general, information obtained from the examinations was processed by the SwRI Data Acquisition System in the following manner. Signal information from each of the ultrasonic instruments was displayed on the analog cathode ray tube (CRT) for each instrument.
Amplitude and time analog information was channeled through electronic gates and examined
.according to preselected levels.:For the signals which exceeded these levels, calibrated voltages were generated which were analogous to the signal amplitude and the sound beam distance to the reflector. The time analog and amplitude analog voltages corresponded to the first triggering signal encountered along the sound beam and the signal with the largest amplitude, respectively. These data were combined with location information and sent to a six-channel strip chart recorder and an analog tape recorder. The SwRI Data Acquisition System handled information from three ultrasonic instruments simultaneously.
This versatile system produced the following types of examination records for this ISI:
Video Tape Strip Chart Analog Tape The PaR Examination Device, fabricated by Programmed and Remote Systems Corporation, with technical design input from SwRI, was used to perform ultrasonic examinations of the RPV welds, components, or areas. The system was utilized during this ISI for lamination from the interior of the nozzles and from the vessel inside surface.,The device was mounted on the top flange of the vessel by three legs which clamped to the guide studs for accurate reproducible positioning. Attached to the boom assembly were various SwRI accessories which enabled examination of the nozzle-to-shell welds, nozzle-to-safe end welds, the nozzle high-strain areas, the vessel shell-to-flange weld, the vessel circumferential welds, and the vessel support welds.
A2- I
Page 2 The results of the 1979 mechanized examinations of the R.E. Ginna RPV were put into four categories:
The term "No Recordable" was applied when no indications were observed greater than the recording level and no indications were observed between 50 percent of the Distance
'Amplitude Correction (DAC) 'curve and the recording level.
"Insignificant" was applied when (1) the amplitude of any indication observed was equal to or greater than the recording level, but less than 100 percent of the DAC curve, and not resolved as being relevant, or (2) nonrelevant indications, such as refiections due to standing waves, water multiples, etc., were observed.
"Geometric" indications were (1) those indications which had an amplitude equal to or greater than 100 percent of the DAC curve, and were resolved and documented to be geometric in nature, or (2) indications that are less than 100 percent of the DAC curve but were suspected by the Level II examiners to be other than geometric in nature, and were then resolved and documented to be geometric in nature.
"Other" ihdications include those indications evaluated to be other than "Insignificant" or "Geometric" indications, as described above.
A2-2
SUMMARY
OF RESULTS The following Table list all MECH UT indications; Insignificant, Geometric and Other, along with the type of search unit used, and how the indications were resolved.
Examination Area ~Insi iTicant Geometric Other Resolution Circumferential Welds RPV-3 Upper Shell to 45o Water Multiple Intermediate Shell 45o Redirected Shear due to Clad RPV-C Intermediate Shell to No recordable Indications Lower Shell RPV-D Lower Shell to Ring Forging OOL Oow 45o T 60o T 45'ater Multiple
'ater Multiple Water Multiple Water Multiple Redirected Shear V-Z Ring Forging to Lower o'L Multiple ead onw Water Multiple 45o T Water Multiple 60o T 45'o'ater Water Multiple Redirected Shear to Outside Surface (OD)
Redirected Shear to OD RPV-A Vessel to Flange, o'L Water Multiple onW Water Multiple 45o Water Multiple 45o Redirected Shear to OD 4oRL} }Vessel OD Between Stud 11oRL} Holes 18 RL}
Vessel Su orts VSL-1 No Recordable Indications VSI 2 No Recordable Indications
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. amination Area ~insi ificsnt Geometric Other Resolution Loo B Outlet Nozzle N-1B Nozzle to Vessel 45o Water Multiple 15o Water Multiple, Integral Extension OD Geometry 60oT 'Water Multiple Nozzle Inside Radius and 60oT Water Multiple Integral Extension Safe End to Nozzle No Recordable Indications Loo B Inlet Nozzle N2B Nozzle to Vessel OoL Water Multiple 45o Water Multiple 45oT Water Multiple 15'RL Water Multiple 15oRL Slag like indications at 310 (A)
Nozzle Inside Radius SW Surface Irregularities (B)
Safe End to Nozzle OoL Water Multiple OoW Water Multiple 45o Water Multiple 45oT Water Multiple Denotes Appendix containing resolution information for that indication.
A2-4
Page 3 amination Area ansi ificant Geometric Other Resolution Loo A Outlet Nozzle NlA Nozzle to Vessel 1SoRL Water Multiple, Integral Pension Water Multiple, Inte'gral 4SO Extension Code Allowable (C) 4So Code AHowable (D) lsoRL Water Multiple 4SoT OD Geometry 60oT Nozzle Inside Radius and 60'T Water Multiple Integral Extension Nozzle to Safe End No Recordable Indications Loo A Inlet Nozzle N2A
. ozzie to Vessel 4So Water Multiple 1S'RL Water Multiple Nozzle Inside Radius SW Surface Irregularities (E)
Safe End to Nozzle 4so Water Multiple OoL ooW Water Multiple 4So Water Multiple 4SoT Water Multiple Water Multiple Denotes Appendix containing resolution information for that indication.
AZ-5
Page 4 xamination Area ~Insi ifiean< Other Resolution Loo B Safe In ection Nozzle
~AC-1002 Nozzle to Vessel 60'eometric Water Multiple 4So OD Surface 60'0 Redirected Shear to OD Integral Extension 45'oL 10o Integral Extension onL Metallurgical Conditions (F) 10o RL Metallurgical Conditions (F)
Nozzle Inside Radius No Recordable Indications Safe End to Nozzle Water Multiple 4s's'T Water Multiple Water Multiple Water Multiple oo A Safe In ection ozzie
~AC-10M 0'. Water Multiple Nozzle to Vessel 0oW Water Multiple Water Multiple 45oT Outside Surface OoL Integral Extension, Metallurgical 45'or, Conditions (F) 10o RL Integral Extension Nozzle Inside Radius Water Multiple, OD Surface 10oRL Water Multiple Safe End to Nozzle OoL Water Multiple Water Multiple 4s's'0o Water Multiple Water Multiple Denotes Appendix containing resolution information for that indication.
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ATTACHMENT 3 ULTRASONIC EVALUATION OF INLET NOZZLE INSERVICE INSPECTION INDICATION