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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~89 AP.EAST AVENUE, ROCHESTER, N.Y.14649 JOHN E.MAIER Vice Prosldont TEI.EPHOHE AREA CODE 7le 546-2700 March 2, 1983 Director of Nuclear Reactor Regulation Attention:
Mr.Dennis M.Crutchfield, Chief Operating Reactors Branch 55 U.S.Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
"Control of Heavy Loads" Supplemental Report Addressing Techni'cal:
Ev'a1ua'tio'n'eo'rt', Dat'ed'u ust9','982
Dear Mr.Crutchfield:
Enclosed please find our response to your letter of August 19E 1982E"Control of Heavy Loads-NUREG 0612." A part two response,"NRC Request for Additional Information on Control of Heavy Loads," will be submitted December 1, 1983.AVery truly yours, E.Maier 8303080333 05000244 P~R~D~C.PDR'-,',;.'.,-
l rl i4 i I', I b l' SUPPZEMENTAL REPORT ON CONTROZ OF HEAVY LOADS RESPONSE TO THE DRAFT TECHNICAL EVALUATION REPORT PREPARED BY FRANKLIN RESEARCH CENTER CONCERNING CONTROL OF HEAVY LOADS-NUREG 0612AT R.E.GINNA NUCZEAR POWER STATION REFERENCED LETTERS 2.'3.NRC GENERIC LETTER 81-07 DATED DECEMBER 22~'980 NRC GENERIC LETTER 81-07 DATED FEBRUARY 3, 1981 LETTER TO MR.DARRELL G.EISENHUT (NRC)FROM MR.JOHN E.MAIER (RG&E)DATED FEBRUARY 1, 1982 LETTER TO MR.JOHN E.MAIER (RGScE)FROM MR.DENNIS M.CRUTCHFIELD (NRC)DATED AUGUST 19~1982 (TECHNICAL=EVALUATION REPORT)R.E.GINNA NUCLEAR POWER STATION ROCHESTER GAS AND EZECTRIC CORPORATION DOCKET NO.50-244 MARCH 1, 1983 MF3A H~Clif I$IY g I, E t dl I I I t I lt I II It,~y\+, It, lt I~f I II ,r H'I yr, tl f I I f i,(I/I I I ,I I r INTRODUCTION AND BACKGROUND On February 1, 1982 Rochester Gas and Electric Corporation submitted a report stating the extent.of compliance with general load-handling policies and procedures at the R.E.Ginna Nuclear Power Plant with NUREG-0612,"Control of Heavy Loads at Nuclear Power Plants", Section 5.1.1.A Technical Evaluation Report (TER)was prepared and returned to Rochester by the NRC and its consultants.
Enclosed in this report are responses to open items in this TER, schedules for implementation of procedures and training programs to assure compliance with NUREG-0612 and justification for exceptions taken to NUREG 0612.MF3A RESPONSE TO THE DRAFT TECHNICAL EVALUATION REPORT PREPARED BY FRANKLIN RESEARCH CENTER CONCERNING"CONTROL OF HEAVY LOADS-NUREG 0612" AT R.E.GINNA NUCLEAR POWER STATION TER Section 2.1.lc"Overhead Hea Load Handlin S stems" Conclusions and Recommendatxons The Licensee should identify overhead handling systems excluded from compliance with NUREG-0612 and verify that such exclusion was based on one or more of the following:
1.No safety-related equipment or irradiated fuel is located in close proximity or sufficient physical separation exists.2.The systems are sole-purpose systems and are used only when the equipment is out of service.3.Heavy loads are not carried by the excluded system.(A heavy load for R.E.Ginna Nuclear Power Station is taken as 1500 lbs.)RG&E Res onse In response to"Conclusions and Recommendations" listed for this section RG&E has provided an enclosed list of all Overhead Handling Systems'at R.E.Ginna Nuclear Power Station (Attachment 1).The list of overhead handling systems within the scope of NUREG-0612 provided in RG&E's February 1, 1982 submittal has been reduced by using the exclusion criteria shown above.Attachment 2 shows the physical location of each overhead handling system identified.
TER Section 2.1.2"Safe Load Paths Guideline 1, NUREG-0612, Sects.on 5.1.1 1 Conclusz.ons and Recommendatxons The Licensee partially complies with Guideline 1 for the Ginna plant.In order to comply fully with the criteria of this guideline, the Licensee should perform the following:
2.3.Provide suitable visual aids to assist the crane operator in the areas where loads.are handled by all equipment that.is not excluded from compliance with NUREG-0612, as specified in 2.1.1.c.F Define all safe load paths in procedures.
Incorporate safe load paths into equipment layout drawings.Verify that deviations from established load paths require written alternatives to be approved by the plant safety review committee.
MF3A
~~I I li 4 dl l'At l'A n 1' RG&E Res onse In response to"Conclusions and Recommendations" listed for this section RG&E will: 1)define specific heavy load safe load paths, 2)define"generic" heavy load safe load paths, 3)incorporate safe load paths into load handling procedure A-1305, 4)establish'a deviation of safe load path control procedure, 5)provide suitable visual aids to assist crane operators properly handle heavy loads where practical.
The above procedures will be implemented for the following Overhead Handling Systems: Crane No.Overhead Handlin S stem 7 29 32 100(2)Ton O.H.Crane in CTMT 40(5)Ton O.H.Crane Auxiliary Bldg.7 1/2 Ton Screenhouse Crane (East)The remaining 7 overhead handling systems within the scope of NUREG-0612 (Attachment 1, Page 6 of 6)are restricted by the physical capabilities of the system and therefore require no additional work regarding safe load paths.RG&E will have these procedures in place by September 1, 1983.TER Section 2.1.3"Load Handlin Procedures" Guideline 2, NUREG-0612, Section 5.1.1 2 Conclusions and Recommendations The Ginna plant does not comply with Guideline 2.Load-handling procedures should clearly identify inspection and acceptance criteria, steps, and proper sequence, and clearly define the safe load paths for the various heavy loads listed.RG&E Res onse In response to"Conclusions and Recommendations" listed for this section RG&E will develop Load Handling Procedures for the 10 Load Handling Systems within the scope of NUREG 0612.The program will identify specific and"generic lifts", insure the proper equipment is utilized for the lift and that the load is handled properly.This development of Load Handling Procedures will be done in administrative procedure A-1305 accompanied by a qualifying rigger program.RG&E will have these procedures in place by December 31, 1983.TER 2.1.5"S ecial Liftin Devices" Guideline 4, NUREG-0612 Section 5.1.1 4 Conclusions and Recommendations A conclusion regarding compliance with this guideline must be deferred until the Licensee completes the analysis of their special lifting devices.However, this analysis should address the specific items identified in this independent evaluation.
NF3A
A comparison analysis for the Special Lifting Devices used at R.E.Ginna Nuclear Power Plant to determine the compliance with NUREG-0612 has been completed by Westinghouse Corp.This analysis encompasses the following special lifting devices: Reactor Head Lifting Rig, The Upper and Lower Internals Lifting Assembly and The Reactor Coolant Pump Motor Sling.The evaluation shows that the stress limit criteria of ANSI N14.6-1978 associated with certain stress design factors for tensile and shear stresses are adequately satisfied.
Enclosed is justification for dynamic load factors considered in the analysis of these special lifting devices.(Attachment 3)RG&E has implemented an inspection,, testing and maintenance program to insure continued compliance with ANSI N14.6-1978.
This program will require visual inspections annually (not to exceed 15 months, consistent with permissible extensions for Technical Specification surveillance).
Nondestructive examination of critical welds will also be done every ten years.At.this time the analysis is based on the original design drawings.During our spring outage 1983"as built," sketches of any modifications or additions to the special lifting devices will be made, reviewed and analyzed to determine if the evaluation performed is effected.RGGE will submit this information by August 1, 1983.TER 2.1.6"Liftin Devices" Not S eciall Desi ned Guideline 5, NUREG-0612 Section 5.1.1 5 Conclusions and Recommendatxons Determination of compliance with Guideline 5 must be defer'red until RG&E provides specific information concerning:
1.Installation and use of slings (in accordance with ANSI B30.9-1971) 2.Selection of slings (base'd upon the sum of the static and maximum dynamic loads)3.Sling markings and sling restrictions, where appropriate In response to"conclusions and recommendations" listed for this section RGSE will implement a rigger qualification program to insure that slings are selected and used in accordance with ANSI B30.9-1971.At this time sling selection and use at R.E.Ginna Station is the responsibility of each individual rigger.The"Handbook for Riggers" by W.G.Newberry, Calgary, Alberta, Canada is currently used and believed to be equivalent to ANSI B30.9-1971.
All slings used at Ginna Station have load ratings marked on them.MF3A 3 Q~~l k)'J'h g l, C I~>~'4 V I>v P~f Pp~*4 1 I L 1 l RG&E believes that adding additional factors to consider dynamic loads above the 5:1 safety factors for slings already considered is impractical and provides no justifiable cushion of safety.Dynamic loads are negligible and only static loads are considered.
RG8E will have the qualified rigger program in place by December 31, 1983.TER 2.1.7"Cranes" Ins ection, Testin and Maintenance Guideline 6, NUREG-0612, Section 5.1.1 6 Conclusions and Recommendations The Ginna plant will comply with Guideline 6 upon implementation of proposed revisions to inspection, testing, and maintenance programs.RGSE Res onse Revisions are being made to the existing RGSE inspection, testing, and maintenance programs listed in the February 1, 1982 submittal to insure full compliance with ANSI B30.2-1976, Chapter 2 2~Additional programs are being developed for jibs and monorails to insure that inspection, testing, and maintenance procedures comply with ANSI B30.11-1980"Monorails and Underhung Cranes" Chapter 11-2.All procedures will be in place by September 31, 1983.In response to a question concerning the Auxiliary Building overhead crane and a"rated load test", the February 1, 1982 submittal stated that this crane was load tested prior to lifting a spent fuel cask in 1973.This is correct.The 40 ton auxiliary crane was load tested using a 40 ton test load.TER 2.1.8"Crane Desi n" Guideline 7, NUREG 0612, Section 5.1.1 7 Conclusion The Ginna Auxiliary Building crane complies with Guideline 7.The Reactor Building substantially complies with this guideline; however, the Licensee should be ready to provide suitable documentation to justify the location and moment of inertia of the actual longitudinal stiffeners.
RG&E Res onse Justification for the location and moment of inertia of the longitudinal stiffeners used on the crane bridge of the Containment.
Overhead Crane is provided in Attachment 4.In response to a question concerning the horsepower rating for the main hoist of the 100 ton containment overhead crane it is a 5 H.P.motor.This capacity was omitted in error in the February 1, 1982 submittal.
MF3A P h Ph P TER Interim Protection Measures Conclusions and Recommendations Insufficient information has been provided to clearly determine compliance with this interim measure.The Licensee should review and determine whether current, technical specifications meet, the intent of this guideline.
RG&E Res onse RG&E has reviewed the technical specifications which limit travel of the Auxiliary Building overhead crane over the spent fuel area and has found them to be adequate interim protection.
No other overhead handling systems can physically move over the spent fuel area except the spent fuel manipulator crane.This crane has undergone analysis for a single spent fuel assembly handling accident and has been deemed not to require further evaluation.
TER S ecial Review for Hea Loads Handled Over the Core Interim Protection Measure 6 NUREG-0612, Section 5.3 6 Evaluation Conclusion and Recommendatz.ons The Licensee has made no statements or conclusions regarding this interim protection measure.The Licensee should report the completion of the special review identified in ths interim measure.RG&E Res onse Safe load paths for heavy load lifts made in containment at Ginna Station are physically restricted.
Procedure A-1305 is currently the controlling load-handling procedure, the crane is inspected according to ANSI and the crane operators are trained according to ANSI.Further analysis of handling of heavy loads in this area is required to satisfy NUREG-0612 and shall be included in the load-drop analysis submittal.
MF3A d v It g I ATTACHMENT 1 TOTAL LIST OF OVERHEAD HANDLING SYSTEMS AT R.E.GINNA NUCLEAR POWER STATION 1 2 3 No.10 11 12 I 13 15,16,17 18 24 25,26 27 28 29 30 31 32 33 34 35 36 37 19,20,21,22 23~Bui1 din Containment Containment, Containment Containment Containment Containment, Containment Containment, Containment H Containment",'ontainment.
Turbine Turbine Turbine Turbine Turbine Turbine Turbine Auxiliary Auxiliary Auxiliary Auxiliary Auxiliary*Screenhouse
- Screenhouse Service Service Service Service Service Overhead Handlin S stem 3 Ton Jib 1 Ton Jib (E)1 Ton Jib (W)1 1/2 Ton Fuel Manipulator Bridge Aux.Trolley on Fuel Manipulator Bridge 10 Ton Jib 100(20)Ton Overhead Crane 2 Ton Jib 1/4 Ton Jib on OverheadCrane, 3 Monorail Hoists'on Reactor Head Monorail'-'ib on Overhead (2)Monorails Over.Feed-water Pumps (Basement)-
2 Ton Jib--Turbine Oil Part (3)Monorails-Condenser Parts (1)Monorail-Feed-water Heater Parts (4)Monorails--Moisture Separator Parts Turbine Overhead 125 Ton (25 Ton)3 Ton Movable Gantry (2)Nonorails RHR Pit (Basement)
(1)Monorail (Basement) 2 Ton Spent Fuel Crane 40(5)Ton Overhead Crane 5 Ton Drumming Crane 7 1/2 Ton Screenhouse (W)7 1/2 Ton Screenhouse (E)2 Ton--Steamfitting Shop 2 Ton--Hot Shop 5 Ton--Machine 6 Elect.Shop 2 Ton--Machine S Elec.Shop 2 Ton--Meter Shop MF3B Page 1 of 6 J a P I,, tI J~'8~f r
.No.38,39 40 41 42 43 44 45 46 47 48 50 51 52 B~nildin Intermediate Intermediate Intermediate Intermediate Intermediate Standby Auz F.W.Pump Bldg.Diesel Gen.Diesel Gen.AVT Bldg.Oil Storage Building Plant Exterior Containment Containment Containment Overhead Handlin S stem (2)Monorails--MDAFWP (1)Monorail--TDAFWP (1)Monorail--M.S.Header 3 Ton Monorail Upper Level (1)Monorail A F.W.Line (1)Monorail (1)Monorail DG1 (1)Monorail.DG2, 1 Ton Stairwell Monorail'1)Monorail 18 Ton Mobile Crane Reactor Vessel Stud Cleaning Crane Base Mounted Pillar Jib Crane Jib at Equipment Hatch (On Wall)*Screenhouse crane is classified as two different cranes, however, there is only one crane for two areas which are separated by a track switch MF3B Page 2 of 6 v i1~3 l OVERHEAD HANDLING SYSTEMS EXCLUDED FROM'COMPZ IANCE WITH NUREG 0612 BASED ON>>NO SAFETY-RELATED E UIPMENT OR IRRADIATED FUEL IS ZOCATED IN CLOSE PROXIMITY OR'SUFFICIENT PHYSICAL SEPARATION EXISTS No.12, 13 15,16,17 18 19,20,21,22 e 23 24 30 31 33 34 35 36 37 47 48 50~Buildin Turbine Turbine Turbine Turbine,i Turbine,..
Turbine Turbine Auxiliary Screenhouse Service Service Service Service Service, AVT Bldg.Oil Storage Building Plant Exterior Containment Overhead Handlin S stem (2)Monorails Over Feed-water Pumps (Basement) 2 Ton Jib--Turbine Oil Part (3)Monorails--
Condenser Parts (1),Monorail
--Feed-water Heater Parts', (4)Monorails-Moisture Separator Parts 125(25)Turbine O.H.Crane 3Ton,,Gantry 5 Ton Drumming'Crane 7 1/2 Ton Screenhouse (W)2 Ton--Steamfitting Shop 2 Ton--Hot Shop 5 Ton-Machine 6 Elect.Shop 2 Ton-Machine 6 Elec.Shop 2 Ton--Meter Shop 1 Ton Stairwell Monorail (1)Monorail 18 Ton Mobile Crane Reactor Vessel Stud Cleaning Crane TOTAL=24 MF3B Page 3 of 6 Ob h'1 I~4 t OVERHEAD HANDLING SYSTEMS EXCLUDED FROM COMPI IANCE WITH NUREG 0612 BASED ON: "SOLE PURPOSE SYSTEMS AND ARE USED ONLY WHEN THE E UIPMENT IS OUT OF SERVICE" No.10 25, 26 38, 39 40 41 TOTAL=11~Buildin Containment Auxiliary Intermediate Intermediate Intermediate Intermediate Standby Aux F.W.Pump Bldg.Diesel Gen.Bldg.A Diesel Gen.Bldg.B Overhead Handlin S stem (3)Monorail Hoists on Reactor Head Monorail (2)Monorails RHR Pit Basement-(2)Monorails--K)AFWP (1)Monorail--TDAFWP Monorail Over Main Steam Header Monorail Over"A" F.W.Line (1)Monorail (1)Monorail ,(1)Monorail All of the above overhead handling systems are used only when the safety related equipment below them or within their proximity is out of service.Each monorail will be posted to prohibit its use while the systems below are in operation and each will be locked to insure that this restriction is adhered to.H MF3B Page 4 of 6 I I I J.gf Y'I OVERHEAD HANDLING SYSTEMS EXCLUDED FROM COMPLIANCE WITH NUREG 0612 BASED ON: "HEAVY LOADS ARE NOT CARRIED BY THE EXCLUDED SYSTEM" (A HEAVY LOAD AS DEFINED BY NUREG 0612 FOR R.E.GINNA STATION IS 1500 LBS.)11 51*52 No.~Buildiu Containment Containment Containment Containment Containment Containment Containment Overhead Handlin S stem 1 Ton Jib (E)1 Ton Jib (W), Aux.,Trolley on Fuel.Manipulator Bridge 1/4 Ton Jib on Overhead Crane Jib on Overhead CTMT Crane Base Mounted Pillar'ib Crane Jib at Equipment Hatch (On Wall)TOTAL=7*Designed, not installed, on hold.All of the overhead handling systems above will be derated and posted so as to prevent.the lifting of any heavy loads.(>1500 lbs.)MF3B Page 5 of 6 k 4 V'h li Total identified Overhead Handling Systems 52 Total eliminated by"Physical Separation" 24 Total eliminated by"Sole Purpose" 11 Total eliminated by"No Heavy Load Lifted" 7 TOTAL OVERHEAD HANDLING SYSTEMS THAT MUST COMPLY WITH NUREG 0612 10 6 7 8 27 28 29 32 42 No.~Bnildin Containment Containment Containment Containment Containment Auxiliary Auxiliary Auxiliary Screenhouse Intermediate Overhead Handlin S stem 3 Ton Jib 1 1/2 Ton Fuel Manipulator Bridge'10 Ton Jib 100(20)Ton O.H.Crane'2 Ton Jib (1)Monorail (Basement) 2 Ton Spent Fuel Crane 40(5)Ton O.H.Crane 7 1/2 Ton Screenhouse (E)3 Ton Monorail-Upper Level Total=10 MF3B Page 6 of 6 I v L j, lt'I g I-i
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I~1 SPECIAL LXPTING DEVICES Attachment 3 1/2 ITEM 1=STRESS DESXGN FACTOR PROM NUREG 0612 The stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device based on character-istics of the crane which will be used.This is in lieu of the guideline in Section 3.2.1.1 of ANSI N14.6 which bases the stress design factor on only the weight (static load)of the load and of the intervening components of the special handling device.DISPOSITION OP DYNAMIC LOADS XN WESTINGHOUSE ANALYSIS OP LIFTING DEVXCES I It can be inferred from this paragraph that the stress design factors specified in Section 3.2.1.1 of ANSI N14.6 (3 and 5)are not all inclusive.
Also, it can be inferred that the specified ANSI N14.6 stress, design factors should be increased by any amount based on the crane dynamic characteristics.
The dynamic characteristics of the crane would be based on the main hook and associated wire ropes holding the hook.Most, main containment'ranes use sixteen (16)or more wire ropes to handle the load.Should the crane hook suddenly stop during the lifting or lowering of a load, a shock load could be transmitted to the connected device.Because of the elasticity of the sixteen or more wire ropes, the dynamic factor for a typical containment crane is:not much larger than 1.0.The maximum design factor that is recommended by most design texts I 7, 8, 9j is a factor of 2 for loads that are suddenly applied.The stress design factors required in Section 3.2.1.1 of ANSI N14.6 are: 3 (weight)<Yield Strength 5 (weight)<Ultimate Strength The factor of 3 specified, certainly, includes consideration of suddenly applied loads for cases where the dynamic impact factor may be as high as 2.0~Thus, we feel that the use of the design criteria in ANSI N14.6 satisfies the NUREG requirement.
CMAA f70 IMPACT LOAD CRXTERIA 3~3~2~1~1.3 Impact Allowance:
For cranes operating on runways as described'n Section 1.4, the impact allowance of the rated capacity shall be taken as 1/2%of the load per foot per minute of hoisting speed, but not less than 158 nor more than 50%, except for bucket and magnet cranes for which the impact value shall be taken as 50%of the rated capacity.
~I ay~RG&E DISPOSITION Since all three of our special lifting devices are operated only with the main hoist of the containment, overhead crane, which has a hoist speed of 3"/min., the following maximum impact load has been calculated:
Rated Capacity=100 Tons Hoist Speed=3 inches/minute Impact Allowance=I I=1/2S (100 tons)/3 inches per minute I=4000 or 2%(158 equals 30,000 lbs.)Westinghouse has determined that the three special lifting devices analyzed are within the 3 and 5 strength factors for static loading.Based on this analysis, the nature of the crane used with these devices, and their actual impact allowance we have determined that pur special lifting devices conform to NUREG 0612.
ATTACHMENT 4 CRANE DESIGN CTMT OVERHEAD ITEM 1: LONGITUDINAL STIFFENERS FROM NUREG 0612 SECTION 5.1.1 (7)The crane should be designed to meet the-applicable criteria and guidelines of Chapter 2-1 of ANSI B30.2-1976,"Overhead and Gantry, Cranes" and of CMAA-70,"Specifications for Electric Overhead Travelling Cranes." An alternative to a specification in ANSI B30.2 or CMAA-70 may be accepted in lieu of specific compliance if the intent of the specification is satisfied.
CRANE COMPARISON ANALYSIS RESULTS A comparison analysis was done on the containment overhead crane to determine its extent of compliance to CMAA-70.The longitudinal stiffeners are not located as specified in CMAA-70.When two longitudinal stiffeners are used, they shall be located as stated in Section 3.3.3.1.$.2 of CMAA-70.Each shall have a moment of inertia of 2407 in.(calculated) the existing moments of.inertia are.68 in.as stated in the"Whiting" crane comparison analysis.TECHNICAL EVALUATION REPORT COMMENTS Although not in verbatim compliance, with CMAA-70 requirements concerning location and moment, of inertia, the use of longitudinal stiffeners, in the containment, building crane is judged to meet the intent of this guideline.
Longitudinal stiffeners are used, in conjunction with transverse stiffeners or diaphragms, to allow the use of thin web plates (i.e., web plates with large h/t ratios where h=web depth, and t=web thickness.)
CMAA-70 allows for h/t ratios of up to 188 for girders with no longitudinal stiffener, and of up to 376 for girders with a single longitudinal stiffener.
Were such a single stiffener used, in the Ginna containment crane, a moment of inertia about the web face of approximately 1.8 in.would be required.The Ginna design employs two longitudinal stjffeners, each with a moment of inertia of approximately 0.46 in.about the web face, in conjunction with a web h/t ratio of 236.Although judged to meet the intent of this guideline, the Licensee should provide suitable documentation to justify the location and moment of inertia of the installed stiffeners.
RGScE DISPOSITION The location of the installed stiffeners in the containment building crane are shown on a Westinghouse Shop Drawing, U-55964 which is a permanent RGB record;(See attached sketch for stiffener location).
MF3D 4 II 4 I,~I Ik I 4 I~I'l tlrl r I 4 4 J~'4k r*4 Calculations for the moment of inertia of the stiffeners are shown on pages 15 and 16 of the Whiting Comparison Report.(February 1, 1982 submittal, Attachment 5).The crane girder was designed under EOCI-Cl specifications using sound engineering practices.
There is simply'o indication that the crane is in jeopardy of structural failure or any overstress due to the location and size of the existing longitudinal stiffeners.
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