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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217D2901999-10-13013 October 1999 Forwards SER Accepting Licensee 990305 Proposed Changes to Edwin I Hatch Nuclear Plant Emergency Classification Scheme to Add Emergency Action Levels Related to Operation of Independent Spent Fuel Storage Installation ML20217G0401999-10-0707 October 1999 Forwards Insp Repts 50-321/99-09 & 50-366/99-09 on 990607-11 & 0823-27.One Violation Occurred Being Treated as NCV ML20217G2631999-10-0707 October 1999 Informs That on 990930,NRC Staff Completed mid-cycle PPR of Hatch Plant & Did Not Identify Any Areas Where Performance Warranted More than Core Insp Program.Regional Initiative Insps to Observe Const Activities Will Be Conducted ML20216G0251999-09-24024 September 1999 Concludes That All Requested Info of GL 98-01 & Supplement 1 Provided & Licensing Action for GL 98-01 & Supplement 1 Complete for Plant ML20216H3641999-09-20020 September 1999 Forwards NRC Form 536 in Response to AL 99-03, Preparation & Scheduling of Operator Licensing Exams HL-5839, Forwards Three New & One Revised Relief Requests for Third 10-year Interval ISI Program for Plant,Developed to Clarify Documentation Requirements,To Propose Alternate Exam Requirements IAW ASME Code Cases N-598 & N-5731999-09-16016 September 1999 Forwards Three New & One Revised Relief Requests for Third 10-year Interval ISI Program for Plant,Developed to Clarify Documentation Requirements,To Propose Alternate Exam Requirements IAW ASME Code Cases N-598 & N-573 ML20217B5271999-09-16016 September 1999 Forwards Insp Repts 50-321/99-05 & 50-366/99-05 on 990711-0821.No Violations Noted ML20212A6411999-09-13013 September 1999 Forwards Safety Evaluation of Relief Request RR-V-16 for Third Ten Year Interval Inservice Testing Program HL-5837, Informs That Recent Evaluation of Inservice Testing Program Activities Has Resulted in Requirement for Snoc to Revise Two Existing Requests for Relief,Withdraw One Request for Relief & Revise One Existing Cold SD Justification1999-09-13013 September 1999 Informs That Recent Evaluation of Inservice Testing Program Activities Has Resulted in Requirement for Snoc to Revise Two Existing Requests for Relief,Withdraw One Request for Relief & Revise One Existing Cold SD Justification HL-5832, Submits Comments Concerning Reactor Vessel Integrity Database (Rvid),Version 2 for Plant Hatch,Per NRC1999-09-0101 September 1999 Submits Comments Concerning Reactor Vessel Integrity Database (Rvid),Version 2 for Plant Hatch,Per NRC ML20211Q4801999-09-0101 September 1999 Informs That on 990812-13,Region II Hosted Training Managers Conference on Recent Changes to Operator Licensing Program. List of Attendees,Copy of Slide Presentations & List of Questions Received from Participants Encl HL-5825, Forwards Response to Informal NRC RAI Re Proposed Emergency Actions Levels Associated with Independent Spent Fuel Storage Operations1999-08-20020 August 1999 Forwards Response to Informal NRC RAI Re Proposed Emergency Actions Levels Associated with Independent Spent Fuel Storage Operations HL-5827, Forwards Fitness for Duty Performance Data for six-month Reporting period,Jan-June 1999,as Required by 10CFR26.71(d)1999-08-19019 August 1999 Forwards Fitness for Duty Performance Data for six-month Reporting period,Jan-June 1999,as Required by 10CFR26.71(d) HL-5788, Forwards Owner Activity Repts,Form OAR-1 for Ei Hatch Nuclear Plant for First Period of Third 10-yr Interval ISI Program.Repts Are for Unit 2 Refueling Outages 13 & 141999-08-19019 August 1999 Forwards Owner Activity Repts,Form OAR-1 for Ei Hatch Nuclear Plant for First Period of Third 10-yr Interval ISI Program.Repts Are for Unit 2 Refueling Outages 13 & 14 HL-5824, Requests Exemption from Expedited Implementation Requirements of Paragraph 10CFR50.55a(g)(6)(ii)(B) as Applicable to Containment General Visual Exams of Subsection Iwe,Table IWE-2500-1,Category E-A,Item E1.111999-08-18018 August 1999 Requests Exemption from Expedited Implementation Requirements of Paragraph 10CFR50.55a(g)(6)(ii)(B) as Applicable to Containment General Visual Exams of Subsection Iwe,Table IWE-2500-1,Category E-A,Item E1.11 ML20210T6421999-08-17017 August 1999 Discusses Licensee 950814 Initial Response to GL 92-01, Rev 1,Supp 1, Rv Structural Integrity (Rvid), Issued on 950519 to Plant.Staff Revised Info in Rvid & Being Released as Rvid Version 2 ML20210V3311999-08-13013 August 1999 Provides Synposis of NRC OI Report Re Alleged Untruthful Statements Made to NRC Re Release of Contaminated Matl to Onsite Landfill.Oi Unable to Conclude That Untruthful state- Ments Were Provided to NRC ML20210Q4821999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Authorized Representative of Facility Must Submit Ltr,As Listed,Identifying Individual to Take Exam,Thirty Days Before Exam Date ML20210L7581999-08-0404 August 1999 Forwards Insp Repts 50-321/99-04 & 50-366/99-04 on 990530-0710.One Violation of NRC Requirements Occurred & Being Treated as Ncv,Consistent with App C of Enforcement Policy ML20210J9501999-08-0202 August 1999 Forwards SER Finding Licensee Established Acceptable Program to Verify Periodically design-basis Capability of safety-related MOVs at Edwin I Hatch Nuclear Plant,Units 1 & 2 ML20210J9021999-08-0202 August 1999 Forwards SER Finding Licensee Adequately Addressed Actions Requested in GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Edwin I Hatch Nuclear Plant,Units 1 & 2 HL-5814, Forwards New Relief Requests for Third 10-year Interval Inservice Insp Program for Ei Hatch Nuclear Plant.New Relief Requests Were Developed to Propose Alternate Insps & to Allow Use of ASME Code Cases N-605,N-508-1,N-323-1 & N-5281999-07-30030 July 1999 Forwards New Relief Requests for Third 10-year Interval Inservice Insp Program for Ei Hatch Nuclear Plant.New Relief Requests Were Developed to Propose Alternate Insps & to Allow Use of ASME Code Cases N-605,N-508-1,N-323-1 & N-528 05000366/LER-1999-007, Forwards LER 99-007-00 Re Personnel Error & Inadequate Corrective Action Causing Automatic Reactor Shutdown1999-07-27027 July 1999 Forwards LER 99-007-00 Re Personnel Error & Inadequate Corrective Action Causing Automatic Reactor Shutdown ML20210E1601999-07-20020 July 1999 Forwards Insp Repts 50-321/99-10 & 50-366/99-10 on 990616-25.One Violation Noted Being Treated as Ncv.Team Identified Lack of Procedural Guidance for Identification & Trending of Repetitive Instrument Drift & Calibr Problems HL-5810, Forwards Rev 17B to Ei Hatch FSAR & Rev 14B to Fire Hazards Analysis & Fire Protection Program, for Plant.Encls Reflect Changes Made Since Previous Submittal.With Instructions1999-07-15015 July 1999 Forwards Rev 17B to Ei Hatch FSAR & Rev 14B to Fire Hazards Analysis & Fire Protection Program, for Plant.Encls Reflect Changes Made Since Previous Submittal.With Instructions HL-5808, Forwards Ei Hatch Nuclear Plant,Unit 1 Extended Power Uprate Startup Test Rept for Cycle 19. Rept Summarizes Startup Testing Performed on Unit 1 Following Implementation of Extended Uprate During Eighteenth Refueling Outage1999-07-15015 July 1999 Forwards Ei Hatch Nuclear Plant,Unit 1 Extended Power Uprate Startup Test Rept for Cycle 19. Rept Summarizes Startup Testing Performed on Unit 1 Following Implementation of Extended Uprate During Eighteenth Refueling Outage HL-5807, Estimates That Seventeen (17) Submittals Will Be Made During Fy 2000 & Two (2) Submittals Will Be Made During Fy 2001,in Response to Administative Ltr 99-02, Operating Reactor Licensing Action Estimates1999-07-14014 July 1999 Estimates That Seventeen (17) Submittals Will Be Made During Fy 2000 & Two (2) Submittals Will Be Made During Fy 2001,in Response to Administative Ltr 99-02, Operating Reactor Licensing Action Estimates ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants HL-5804, Informs NRC That on or After 991018,SNOC Plans to Begin Storing Spent Fuel in Ei Hatch ISFSI IAW General License Issued,Per 10CFR72.2101999-07-0909 July 1999 Informs NRC That on or After 991018,SNOC Plans to Begin Storing Spent Fuel in Ei Hatch ISFSI IAW General License Issued,Per 10CFR72.210 HL-5796, Forwards Response to NRC 990331 Request for Supplemental Info Re SNC Earlier GL 95-07 Responses.Gl 95-07 Evaluation Sheets for Units 1 & 2 RHR Torus Spray Isolation Valves Are Encl1999-07-0909 July 1999 Forwards Response to NRC 990331 Request for Supplemental Info Re SNC Earlier GL 95-07 Responses.Gl 95-07 Evaluation Sheets for Units 1 & 2 RHR Torus Spray Isolation Valves Are Encl HL-5801, Forwards New Relief Requests for Third 10-Yr Interval ISI Program for Ei Hatch Nuclear Plant.New Relief Requests RR-25 & RR-26 Were Developed to Propose Alternate Repair Techniques IAW ASME Code N-562 & N-561,respectively1999-07-0909 July 1999 Forwards New Relief Requests for Third 10-Yr Interval ISI Program for Ei Hatch Nuclear Plant.New Relief Requests RR-25 & RR-26 Were Developed to Propose Alternate Repair Techniques IAW ASME Code N-562 & N-561,respectively ML20209E4801999-06-30030 June 1999 Confirms 990630 Telcon Between M Crosby & DC Payne Re Arrangements Made for Administration of Licensing Exam at Plant During Weeks of 991018-1101 ML20196H8811999-06-25025 June 1999 Forwards Insp Repts 50-321/99-03 & 50-366/99-03 on 990418- 0529.No Violations Occurred.Conduct of Activities at Plant Generally Characterized by safety-conscious Operations & Sound Engineering & Maint Practices HL-5790, Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants1999-06-21021 June 1999 Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20212J0541999-06-17017 June 1999 Responds to Requesting That NRC Staff ...Allow BWR Plants Identified to Defer Weld Overlay Exams Until March 2001 or Until Completion of NRC Staff Review & Approval of Proposed Generic Rept,Whichever Comes First ML20207E7561999-06-0303 June 1999 Informs of Completion of Review & Evaluation of Info Provided by Southern Nuclear Operating Co by Ltr Dtd 980608, Proposing Changes to Third 10-Yr Interval ISI Program Plan Requests for Relief RR-4 & R-6.Requests Acceptable HL-5763, Informs That Util Is Changing Responsibility for Periodic Concerns Program Review,Per Insp Repts 50-321/95-12 & 50-366/95-12.Reviews Will Now Be Performed Under Direction of Vice President & Corporate Counsel1999-05-20020 May 1999 Informs That Util Is Changing Responsibility for Periodic Concerns Program Review,Per Insp Repts 50-321/95-12 & 50-366/95-12.Reviews Will Now Be Performed Under Direction of Vice President & Corporate Counsel HL-5785, Forwards New Relief Requests RR-V-16 for Third 10-yr Interval Inservice Testing Program for Ei Hatch Nuclear Plant.Relief Request Was Developed to Propose Alternate Schedule for Replacement of HPCI Rupture Disks1999-05-18018 May 1999 Forwards New Relief Requests RR-V-16 for Third 10-yr Interval Inservice Testing Program for Ei Hatch Nuclear Plant.Relief Request Was Developed to Propose Alternate Schedule for Replacement of HPCI Rupture Disks ML20206Q0751999-05-0606 May 1999 Forwards Insp Repts 50-321/99-02 & 50-366/99-02 on 990307-0417.No Violations Noted ML20206G1611999-05-0404 May 1999 Forwards SER Approving Util 990316 Revised Relief Request RR-P-14,for Inservice Testing Program for Pumps & Valves Pursuant to 10CFR50.55a(a)(3)(ii) ML20206P6921999-04-27027 April 1999 Discusses 990422 Public Meeting at Hatch Facility Re Results of Periodic Plant Performance Review for Hatch Nuclear Facility for Period of Feb 1997 to Jan 1999.List of Attendees & Copy of Handouts Used by Hatch,Encl HL-5777, Notifies NRC That Exam Coverage for One Weld Exceeded Criteria for Plant.Reasons for Exam Coverage Being Less than Initially Estimated as Listed1999-04-26026 April 1999 Notifies NRC That Exam Coverage for One Weld Exceeded Criteria for Plant.Reasons for Exam Coverage Being Less than Initially Estimated as Listed HL-5758, Forwards Response to NRC 990129 RAI Re GL 96-05 Program at Ei Hatch Nuclear Plant1999-04-0909 April 1999 Forwards Response to NRC 990129 RAI Re GL 96-05 Program at Ei Hatch Nuclear Plant ML20205T1831999-04-0909 April 1999 Informs That on 990316,S Grantham & Ho Christensen Confirmed Initial Operator Licensing Exam Schedule for Ei Hatch NPP for FY00.Initial Exam Dates Are 991001 & 2201 for Approx 12 Candidates.Chief Examiner Will Be C Payne ML20205M3181999-04-0707 April 1999 Confirms Telcon Between D Crowe & Ph Skinner Re Mgt Meeting Scheduled for 990422 in Conference Room of Maint Training Bldg.Purpose of Meeting to Discuss Results of Periodic PPR for Plant for Period of Feb 1997 - Jan 1999 ML20205M3011999-04-0202 April 1999 Forwards Insp Repts 50-321/99-01 & 50-366/99-01 on 990124-0306.Non-cited Violation Identified HL-5761, Forwards Revised Ei Hatch Nuclear Plant Psp,Effective 990331,per 10CFR50.54(p)(2).Justification & Detailed Instructions Encl.Plan Withheld,Per 10CFR73.211999-03-31031 March 1999 Forwards Revised Ei Hatch Nuclear Plant Psp,Effective 990331,per 10CFR50.54(p)(2).Justification & Detailed Instructions Encl.Plan Withheld,Per 10CFR73.21 HL-5750, Forwards Status of Decommissining Funding,Per Requirements 10CFR50.75(f)(1),on Behalf of Licensed Owners of Ei Hatch Nuclear Plant1999-03-30030 March 1999 Forwards Status of Decommissining Funding,Per Requirements 10CFR50.75(f)(1),on Behalf of Licensed Owners of Ei Hatch Nuclear Plant ML20205H1491999-03-25025 March 1999 Forwards Info for OLs DPR-7 & NPF-5 Re Financial Assurance Requirements for Decommissioning Nuclear Power Reactors,Per 10CFR50.75(f)(1).Municipal Electric Authority of Georgia Is One of Licensed Owners of Ei Hatch,Owning 17.7% of Facility ML20205D3211999-03-24024 March 1999 Informs That Safety Sys Engineering Insp Previously Scheduled for 990405-09 & 19-23,rescheduled for 990607-11 & 21-25 1999-09-24
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20216H3641999-09-20020 September 1999 Forwards NRC Form 536 in Response to AL 99-03, Preparation & Scheduling of Operator Licensing Exams HL-5839, Forwards Three New & One Revised Relief Requests for Third 10-year Interval ISI Program for Plant,Developed to Clarify Documentation Requirements,To Propose Alternate Exam Requirements IAW ASME Code Cases N-598 & N-5731999-09-16016 September 1999 Forwards Three New & One Revised Relief Requests for Third 10-year Interval ISI Program for Plant,Developed to Clarify Documentation Requirements,To Propose Alternate Exam Requirements IAW ASME Code Cases N-598 & N-573 HL-5837, Informs That Recent Evaluation of Inservice Testing Program Activities Has Resulted in Requirement for Snoc to Revise Two Existing Requests for Relief,Withdraw One Request for Relief & Revise One Existing Cold SD Justification1999-09-13013 September 1999 Informs That Recent Evaluation of Inservice Testing Program Activities Has Resulted in Requirement for Snoc to Revise Two Existing Requests for Relief,Withdraw One Request for Relief & Revise One Existing Cold SD Justification HL-5832, Submits Comments Concerning Reactor Vessel Integrity Database (Rvid),Version 2 for Plant Hatch,Per NRC1999-09-0101 September 1999 Submits Comments Concerning Reactor Vessel Integrity Database (Rvid),Version 2 for Plant Hatch,Per NRC HL-5825, Forwards Response to Informal NRC RAI Re Proposed Emergency Actions Levels Associated with Independent Spent Fuel Storage Operations1999-08-20020 August 1999 Forwards Response to Informal NRC RAI Re Proposed Emergency Actions Levels Associated with Independent Spent Fuel Storage Operations HL-5827, Forwards Fitness for Duty Performance Data for six-month Reporting period,Jan-June 1999,as Required by 10CFR26.71(d)1999-08-19019 August 1999 Forwards Fitness for Duty Performance Data for six-month Reporting period,Jan-June 1999,as Required by 10CFR26.71(d) HL-5788, Forwards Owner Activity Repts,Form OAR-1 for Ei Hatch Nuclear Plant for First Period of Third 10-yr Interval ISI Program.Repts Are for Unit 2 Refueling Outages 13 & 141999-08-19019 August 1999 Forwards Owner Activity Repts,Form OAR-1 for Ei Hatch Nuclear Plant for First Period of Third 10-yr Interval ISI Program.Repts Are for Unit 2 Refueling Outages 13 & 14 HL-5824, Requests Exemption from Expedited Implementation Requirements of Paragraph 10CFR50.55a(g)(6)(ii)(B) as Applicable to Containment General Visual Exams of Subsection Iwe,Table IWE-2500-1,Category E-A,Item E1.111999-08-18018 August 1999 Requests Exemption from Expedited Implementation Requirements of Paragraph 10CFR50.55a(g)(6)(ii)(B) as Applicable to Containment General Visual Exams of Subsection Iwe,Table IWE-2500-1,Category E-A,Item E1.11 HL-5814, Forwards New Relief Requests for Third 10-year Interval Inservice Insp Program for Ei Hatch Nuclear Plant.New Relief Requests Were Developed to Propose Alternate Insps & to Allow Use of ASME Code Cases N-605,N-508-1,N-323-1 & N-5281999-07-30030 July 1999 Forwards New Relief Requests for Third 10-year Interval Inservice Insp Program for Ei Hatch Nuclear Plant.New Relief Requests Were Developed to Propose Alternate Insps & to Allow Use of ASME Code Cases N-605,N-508-1,N-323-1 & N-528 05000366/LER-1999-007, Forwards LER 99-007-00 Re Personnel Error & Inadequate Corrective Action Causing Automatic Reactor Shutdown1999-07-27027 July 1999 Forwards LER 99-007-00 Re Personnel Error & Inadequate Corrective Action Causing Automatic Reactor Shutdown HL-5810, Forwards Rev 17B to Ei Hatch FSAR & Rev 14B to Fire Hazards Analysis & Fire Protection Program, for Plant.Encls Reflect Changes Made Since Previous Submittal.With Instructions1999-07-15015 July 1999 Forwards Rev 17B to Ei Hatch FSAR & Rev 14B to Fire Hazards Analysis & Fire Protection Program, for Plant.Encls Reflect Changes Made Since Previous Submittal.With Instructions HL-5808, Forwards Ei Hatch Nuclear Plant,Unit 1 Extended Power Uprate Startup Test Rept for Cycle 19. Rept Summarizes Startup Testing Performed on Unit 1 Following Implementation of Extended Uprate During Eighteenth Refueling Outage1999-07-15015 July 1999 Forwards Ei Hatch Nuclear Plant,Unit 1 Extended Power Uprate Startup Test Rept for Cycle 19. Rept Summarizes Startup Testing Performed on Unit 1 Following Implementation of Extended Uprate During Eighteenth Refueling Outage HL-5807, Estimates That Seventeen (17) Submittals Will Be Made During Fy 2000 & Two (2) Submittals Will Be Made During Fy 2001,in Response to Administative Ltr 99-02, Operating Reactor Licensing Action Estimates1999-07-14014 July 1999 Estimates That Seventeen (17) Submittals Will Be Made During Fy 2000 & Two (2) Submittals Will Be Made During Fy 2001,in Response to Administative Ltr 99-02, Operating Reactor Licensing Action Estimates ML20210J1091999-07-10010 July 1999 Submits Suggestions & Concerns Re Y2K & Nuclear Power Plants HL-5801, Forwards New Relief Requests for Third 10-Yr Interval ISI Program for Ei Hatch Nuclear Plant.New Relief Requests RR-25 & RR-26 Were Developed to Propose Alternate Repair Techniques IAW ASME Code N-562 & N-561,respectively1999-07-0909 July 1999 Forwards New Relief Requests for Third 10-Yr Interval ISI Program for Ei Hatch Nuclear Plant.New Relief Requests RR-25 & RR-26 Were Developed to Propose Alternate Repair Techniques IAW ASME Code N-562 & N-561,respectively HL-5796, Forwards Response to NRC 990331 Request for Supplemental Info Re SNC Earlier GL 95-07 Responses.Gl 95-07 Evaluation Sheets for Units 1 & 2 RHR Torus Spray Isolation Valves Are Encl1999-07-0909 July 1999 Forwards Response to NRC 990331 Request for Supplemental Info Re SNC Earlier GL 95-07 Responses.Gl 95-07 Evaluation Sheets for Units 1 & 2 RHR Torus Spray Isolation Valves Are Encl HL-5804, Informs NRC That on or After 991018,SNOC Plans to Begin Storing Spent Fuel in Ei Hatch ISFSI IAW General License Issued,Per 10CFR72.2101999-07-0909 July 1999 Informs NRC That on or After 991018,SNOC Plans to Begin Storing Spent Fuel in Ei Hatch ISFSI IAW General License Issued,Per 10CFR72.210 HL-5790, Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants1999-06-21021 June 1999 Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants HL-5763, Informs That Util Is Changing Responsibility for Periodic Concerns Program Review,Per Insp Repts 50-321/95-12 & 50-366/95-12.Reviews Will Now Be Performed Under Direction of Vice President & Corporate Counsel1999-05-20020 May 1999 Informs That Util Is Changing Responsibility for Periodic Concerns Program Review,Per Insp Repts 50-321/95-12 & 50-366/95-12.Reviews Will Now Be Performed Under Direction of Vice President & Corporate Counsel HL-5785, Forwards New Relief Requests RR-V-16 for Third 10-yr Interval Inservice Testing Program for Ei Hatch Nuclear Plant.Relief Request Was Developed to Propose Alternate Schedule for Replacement of HPCI Rupture Disks1999-05-18018 May 1999 Forwards New Relief Requests RR-V-16 for Third 10-yr Interval Inservice Testing Program for Ei Hatch Nuclear Plant.Relief Request Was Developed to Propose Alternate Schedule for Replacement of HPCI Rupture Disks HL-5777, Notifies NRC That Exam Coverage for One Weld Exceeded Criteria for Plant.Reasons for Exam Coverage Being Less than Initially Estimated as Listed1999-04-26026 April 1999 Notifies NRC That Exam Coverage for One Weld Exceeded Criteria for Plant.Reasons for Exam Coverage Being Less than Initially Estimated as Listed HL-5758, Forwards Response to NRC 990129 RAI Re GL 96-05 Program at Ei Hatch Nuclear Plant1999-04-0909 April 1999 Forwards Response to NRC 990129 RAI Re GL 96-05 Program at Ei Hatch Nuclear Plant HL-5761, Forwards Revised Ei Hatch Nuclear Plant Psp,Effective 990331,per 10CFR50.54(p)(2).Justification & Detailed Instructions Encl.Plan Withheld,Per 10CFR73.211999-03-31031 March 1999 Forwards Revised Ei Hatch Nuclear Plant Psp,Effective 990331,per 10CFR50.54(p)(2).Justification & Detailed Instructions Encl.Plan Withheld,Per 10CFR73.21 HL-5750, Forwards Status of Decommissining Funding,Per Requirements 10CFR50.75(f)(1),on Behalf of Licensed Owners of Ei Hatch Nuclear Plant1999-03-30030 March 1999 Forwards Status of Decommissining Funding,Per Requirements 10CFR50.75(f)(1),on Behalf of Licensed Owners of Ei Hatch Nuclear Plant ML20205H1491999-03-25025 March 1999 Forwards Info for OLs DPR-7 & NPF-5 Re Financial Assurance Requirements for Decommissioning Nuclear Power Reactors,Per 10CFR50.75(f)(1).Municipal Electric Authority of Georgia Is One of Licensed Owners of Ei Hatch,Owning 17.7% of Facility ML20205H1411999-03-24024 March 1999 Forwards Info for OLs DPR-7 & NPF-5 Re Financial Assurance Requirement for Decommissioning Nuclear Power Reactors,Per 10CFR50.75(f)(1).Oglethorpe Power Corp Is One of Licensed Owners of Ei Hatch,Units 1 & 2,owning 30% of Facility HL-5754, Forwards Ei Hatch Nuclear Plant Unit 2 Extended Power Uprate Startup Test Rept for Cycle 15, IAW Regulatory Commitments.Testing Identified No Major Problems.Startup Testing for Unit 1 Is Scheduled for Spring 1999 RFO1999-03-22022 March 1999 Forwards Ei Hatch Nuclear Plant Unit 2 Extended Power Uprate Startup Test Rept for Cycle 15, IAW Regulatory Commitments.Testing Identified No Major Problems.Startup Testing for Unit 1 Is Scheduled for Spring 1999 RFO ML20205H1381999-03-22022 March 1999 Forwards Info for OLs DPR-7 & NPF-5 Re Financial Assurance Requirements for Decommissioning Nuclear Power Reactors,Per 10CFR50.75(f)(1).Georgia Power Is One of Licensed Owners of Ei Hatch,Units 1 & 2,owning 50.1% of Facility ML20205H1581999-03-16016 March 1999 Forwards Info for OLs DPR-5 & NPF-7 Re Financial Assurance Requirements for Decommissioning Nuclear Power Reactors,Per 10CFR50.75(f)(1).Dalton Utilities Is One of Licensed Owners of Ei Hatch,Units 1 & 2,owning 2.2% of Facility HL-5753, Submits Response to Five Criteria Described in NUREG/CR-6396 Section 3.3.2.Attachment Contains Relief Request RR-P-14, Which Has Been Revised & Enhanced to Include Addl Info to Justify Proposed Alternative1999-03-16016 March 1999 Submits Response to Five Criteria Described in NUREG/CR-6396 Section 3.3.2.Attachment Contains Relief Request RR-P-14, Which Has Been Revised & Enhanced to Include Addl Info to Justify Proposed Alternative HL-5757, Informs That SNC Withdraws SNC Proposed Rev to Plant Hatch QA Program Submitted 9901271999-03-15015 March 1999 Informs That SNC Withdraws SNC Proposed Rev to Plant Hatch QA Program Submitted 990127 HL-5756, Submits Summary Rept of Present Level & Source of Onsite Property Damage Insurance Coverage for Ei Hatch Nuclear Plant,Units 1 & 21999-03-12012 March 1999 Submits Summary Rept of Present Level & Source of Onsite Property Damage Insurance Coverage for Ei Hatch Nuclear Plant,Units 1 & 2 HL-5751, Forwards Change to Hatch EP for Review & Approval,Per Requirements of 10CFR50,App E,Section Iv.B.Bases for Proposed EALs & Statements of Agreement from State & Local Governmental Authorities1999-03-0505 March 1999 Forwards Change to Hatch EP for Review & Approval,Per Requirements of 10CFR50,App E,Section Iv.B.Bases for Proposed EALs & Statements of Agreement from State & Local Governmental Authorities HL-5735, Provides NRC with Update Status of Progress Util Is Making Toward Development of Product That Was Originally Identified in to NRC1999-03-0202 March 1999 Provides NRC with Update Status of Progress Util Is Making Toward Development of Product That Was Originally Identified in to NRC HL-5737, Forwards Part of Table (Table 1) Util Provided in 980728 Response to RAI for GL 92-01,Rev 1,Suppl 1,reflecting Lower Shell Axial Welds Identified as C-4-A Through C-4-C & Lower Intermediate Axial Welds Identified as C-3-A Through C-3-C1999-02-0505 February 1999 Forwards Part of Table (Table 1) Util Provided in 980728 Response to RAI for GL 92-01,Rev 1,Suppl 1,reflecting Lower Shell Axial Welds Identified as C-4-A Through C-4-C & Lower Intermediate Axial Welds Identified as C-3-A Through C-3-C HL-5733, Forwards Rev 17A to Ei Hatch FSAR & Rev 14A to Fire Hazards Analysis & Fire Protection Program, for Plant.Encl Reflects Changes Made Since Previous Submittal.With Instructions1999-01-29029 January 1999 Forwards Rev 17A to Ei Hatch FSAR & Rev 14A to Fire Hazards Analysis & Fire Protection Program, for Plant.Encl Reflects Changes Made Since Previous Submittal.With Instructions HL-5729, Submits Proposed Rev to Plant QA Program as Described in Chapter 17 of Unit 2 Fsar,Per 10CFR50.54(a)(3).Rev Would Relocate to Controlled Document Other than Fsar,Units 1 & 2 Lists of SR SSC Comprising Items Covered Under QA Program1999-01-27027 January 1999 Submits Proposed Rev to Plant QA Program as Described in Chapter 17 of Unit 2 Fsar,Per 10CFR50.54(a)(3).Rev Would Relocate to Controlled Document Other than Fsar,Units 1 & 2 Lists of SR SSC Comprising Items Covered Under QA Program HL-5728, Requests That RPV Shell Weld Ultrasonic Exams Be Performed Using Techniques & Procedures,Iaw ASME Section Xi,App Viii, 1992 Edition,1993 Addendum,In Lieu of ASME Section XI,1989 Edition Exam Techniques & Procedures1999-01-19019 January 1999 Requests That RPV Shell Weld Ultrasonic Exams Be Performed Using Techniques & Procedures,Iaw ASME Section Xi,App Viii, 1992 Edition,1993 Addendum,In Lieu of ASME Section XI,1989 Edition Exam Techniques & Procedures HL-5712, Forwards Ehnp Intake Structure License Rept, to Provide Example of Technical Content & Level of Detail Planned for Application for License Renewal,For NRC Review1999-01-0707 January 1999 Forwards Ehnp Intake Structure License Rept, to Provide Example of Technical Content & Level of Detail Planned for Application for License Renewal,For NRC Review HL-5725, Informs That SNC Intends to Examine Approx Dozen Weld Overlays During Unit 1 Spring 1999 Outage That to Date Have Not Been Examined Three Times Since Weld Overlay Was Supplied1999-01-0707 January 1999 Informs That SNC Intends to Examine Approx Dozen Weld Overlays During Unit 1 Spring 1999 Outage That to Date Have Not Been Examined Three Times Since Weld Overlay Was Supplied 05000366/LER-1998-004, Forwards LER 98-004-01 Re Personnel Error Which Resulted in Condition Prohibited by Ts.Rev Is Necessary Because Original LER Reported Inadvertent Criticality Could Have Occurred1999-01-0404 January 1999 Forwards LER 98-004-01 Re Personnel Error Which Resulted in Condition Prohibited by Ts.Rev Is Necessary Because Original LER Reported Inadvertent Criticality Could Have Occurred HL-5710, Requests Approval of Alternative Reactor Vessel Weld Exam for Ehnp,Unit 1,per 10CFR50.55a(g)(6)(ii)(A)(5) for Permanently Excluding Exam of RPV Circumferential Shell Welds1998-12-0202 December 1998 Requests Approval of Alternative Reactor Vessel Weld Exam for Ehnp,Unit 1,per 10CFR50.55a(g)(6)(ii)(A)(5) for Permanently Excluding Exam of RPV Circumferential Shell Welds HL-5708, Provides Updated Response to RAI for GL 96-06, Waterhammer in Containment Coolers1998-11-20020 November 1998 Provides Updated Response to RAI for GL 96-06, Waterhammer in Containment Coolers HL-5573, Withdraws Requesting Exemption from Requirements of GDC 56 with Respect to Ei Hatch,Unit 2 Reactor bldg-to- Suppression Chamber Vacuum Relief Sys Design Due to Listed Reasons1998-10-19019 October 1998 Withdraws Requesting Exemption from Requirements of GDC 56 with Respect to Ei Hatch,Unit 2 Reactor bldg-to- Suppression Chamber Vacuum Relief Sys Design Due to Listed Reasons HL-5687, Forwards Required 120-day Response to GL 98-04, Potential for Degradation of ECCS & Containment Spray Sys After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment1998-10-19019 October 1998 Forwards Required 120-day Response to GL 98-04, Potential for Degradation of ECCS & Containment Spray Sys After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment HL-5686, Informs That Corrective Actions Relative to Thermo-Lag 330-1 Issues Are Complete1998-10-16016 October 1998 Informs That Corrective Actions Relative to Thermo-Lag 330-1 Issues Are Complete HL-5697, Requests Delay of Insp of RPV Circumferential Shell Welds for Two Operating Cycles as Outlined in NRC Info Notice 97-63,Suppl 1 & That Proposed Alternative Be Authorized Per 10CFR50.55a(a)(3)(i),per1998-10-16016 October 1998 Requests Delay of Insp of RPV Circumferential Shell Welds for Two Operating Cycles as Outlined in NRC Info Notice 97-63,Suppl 1 & That Proposed Alternative Be Authorized Per 10CFR50.55a(a)(3)(i),per HL-5689, Forwards Request for NRC Finding of Exigent Circumstance Re License Amend for Extended Power Update Operation.Changes Would Be Made While Plant Is Operating,Which as Stated,Has Potential to Create Unnecessary Plant Transients1998-09-30030 September 1998 Forwards Request for NRC Finding of Exigent Circumstance Re License Amend for Extended Power Update Operation.Changes Would Be Made While Plant Is Operating,Which as Stated,Has Potential to Create Unnecessary Plant Transients HL-5673, Informs NRC That Util Intends to Load Four Lead Use Assemblies Into Unit 2 Core as Part of Reload 14/Cycle 15 During Fall 1998 Refueling Outage.Proprietary Info Encl. Proprietary Info Withheld,Per 10CFR2.7901998-09-18018 September 1998 Informs NRC That Util Intends to Load Four Lead Use Assemblies Into Unit 2 Core as Part of Reload 14/Cycle 15 During Fall 1998 Refueling Outage.Proprietary Info Encl. Proprietary Info Withheld,Per 10CFR2.790 HL-5680, Informs NRC of Util Plans & Schedule for Exam of Unit 1 RPV Shell Welds.Util Anticipates Having Final Relief Request Submitted to NRC for Review by 9907011998-09-18018 September 1998 Informs NRC of Util Plans & Schedule for Exam of Unit 1 RPV Shell Welds.Util Anticipates Having Final Relief Request Submitted to NRC for Review by 990701 1999-09-20
[Table view] Category:UTILITY TO NRC
MONTHYEARHL-1278, Responds to NRC Re Violations Noted in Insp Repts 50-321/90-15 & 50-366/90-15.Corrective Actions: Mispositioned Valves 1E21-F025B & 1E21-F027B Placed in Correct Positions & Technicians Disciplined1990-09-12012 September 1990 Responds to NRC Re Violations Noted in Insp Repts 50-321/90-15 & 50-366/90-15.Corrective Actions: Mispositioned Valves 1E21-F025B & 1E21-F027B Placed in Correct Positions & Technicians Disciplined HL-1176, Forwards Updated Listing of Outstanding Licensing Requests for Plant,Tabulated Chronologically by Util Submittal Date1990-09-12012 September 1990 Forwards Updated Listing of Outstanding Licensing Requests for Plant,Tabulated Chronologically by Util Submittal Date HL-1237, Requests Permission to Use Facility Reactor Bldg Crane to Move Large Shipping Casks,Per Tech Spec 3.10.F.1 & 4.10.F.11990-09-0404 September 1990 Requests Permission to Use Facility Reactor Bldg Crane to Move Large Shipping Casks,Per Tech Spec 3.10.F.1 & 4.10.F.1 HL-1250, Forwards Post-Refueling Outage Startup Test Rept,Unit 1 Cycle 13, Per Tech Spec 6.9.1.1.Rept Presents Static & Dynamic Functional Core Tests Performed During Startup from Spring 1990 Maint/Refueling Outage1990-08-27027 August 1990 Forwards Post-Refueling Outage Startup Test Rept,Unit 1 Cycle 13, Per Tech Spec 6.9.1.1.Rept Presents Static & Dynamic Functional Core Tests Performed During Startup from Spring 1990 Maint/Refueling Outage ML20059C6551990-08-27027 August 1990 Informs of Intention to Transfer Right of Way for Road 451 to Appling County So Road Can Be Straightened & Paved. Transfer Will Have No Significant Impact on Use of Road & Site Emergency Plan ML20028G8441990-08-27027 August 1990 Forwards Owners Data Rept for Inservice Insp Ei Hatch Nuclear Plant Unit 1 Feb-June 1990. HL-1245, Forwards Fitness for Duty Performance Data for First Six Month Reporting Period,Per 10CFR26.71d1990-08-23023 August 1990 Forwards Fitness for Duty Performance Data for First Six Month Reporting Period,Per 10CFR26.71d ML20056B3011990-08-20020 August 1990 Forwards Revised Ei Hatch Nuclear Plant,Units 1 & 2 Inservice Insp Program Second 10-Yr Interval, for Review & Approval.Program Will Be Implemented While Awaiting SER HL-1215, Informs of Implementation of Amend 169 to Facility Tech Specs1990-07-26026 July 1990 Informs of Implementation of Amend 169 to Facility Tech Specs HL-1035, Forwards Nuclear Decommissioning Funding Plan for Plant,Per 10CFR50.75(b) & 33(k).Reasonable Assurance That NRC Prescribed Min Funding Will Be Available to Decommission Each Unit on Current Expiration Date Exists1990-07-25025 July 1990 Forwards Nuclear Decommissioning Funding Plan for Plant,Per 10CFR50.75(b) & 33(k).Reasonable Assurance That NRC Prescribed Min Funding Will Be Available to Decommission Each Unit on Current Expiration Date Exists HL-1158, Forwards Rev 0 to SIR-90-039, Flaw Evaluation & Weld Overlay Designs for Ei Hatch Unit 1 Spring 1990 Refueling Outage. Rept Details IGSCC Exam Results,Weld Overlay Design & Evaluation of Weld Shrinkage Stresses1990-06-29029 June 1990 Forwards Rev 0 to SIR-90-039, Flaw Evaluation & Weld Overlay Designs for Ei Hatch Unit 1 Spring 1990 Refueling Outage. Rept Details IGSCC Exam Results,Weld Overlay Design & Evaluation of Weld Shrinkage Stresses ML20043E6691990-06-0707 June 1990 Forwards Rev 0 to Core Operating Limits Rept for Operating Cycle 13, Per Amend 168 to License DPR-57 ML20043C8621990-05-31031 May 1990 Submits Certification That Operator Licensing Simulation Facility Located at Plant Meets NRC Requirements ML20043A8081990-05-0707 May 1990 Forwards Response to NRC 900410 Ltr Re Violations Noted in Insp Repts 50-321/90-07 & 50-366/90-07.Encl Withheld (Ref 10CFR73.21) ML20042F3331990-05-0101 May 1990 Provides Response to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Power Plants. Plant Procedures Address Possibility of Vessel Overfill Events & Training Alert Operators to Potential Overfills ML20012C6351990-03-14014 March 1990 Responds to Generic Ltr 89-19 Re Safety Implementation of Control Sys in LWR Nuclear Power Plants,Per 890920 Request & Understands That NRC Has Agreed to Extend Response Deadline Until 900504 ML20012B7291990-03-0707 March 1990 Forwards Owners Data Rept for Inservice Insp Ei Hatch Nuclear Plant,Unit 2 Sept-Dec 1989 & Metallurgical Evaluation of Four Inch Pipe to Elbow Weld from Plant Hatch, Unit 2. ML20012B1161990-03-0707 March 1990 Forwards Results of Circuit Breaker Testing,Per Bulletin 88-010,per Telcon W/Lp Crocker ML20012A1261990-03-0101 March 1990 Forwards Completed Questionnaire in Response to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey. ML20012A9051990-02-27027 February 1990 Forwards Summary Rept of Present Level & Source of Onsite Property Damage Insurance Coverage for Plant ML20012B4101990-02-22022 February 1990 Discusses NRC 900221 Granting of Discretionary Enforcement to Continue Shutdown Cooling Operation Until Reactor Level Instrument 1B21-N080A Can Be Returned to Svc.Replacement Expected to Be Completed by 900222 ML20006F4561990-02-20020 February 1990 Responds to Request for Addl Info Re Generic Ltr 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Matl & Impact on Plant Operations. RTNDT Value for Unit 2 Closure Flange Region Addressed ML20006D7481990-02-0606 February 1990 Forwards Final Technical Rept, Edwin I Hatch Nuclear Plant Unit 2 Reactor Containment Bldg 1989 Integrated Leakage Rate Test for Fall 1989 Maint/Refueling Outage,Per IE Notice 85-071 ML20006C9481990-01-31031 January 1990 Responds to NRC 900102 Ltr Re Violations Noted in Insp Repts 50-321/89-28 & 50-366/89-28.Corrective Actions:Deficiency Card Documenting Event Initiated as Required by Plant Procedures ML20006A8911990-01-23023 January 1990 Responds to Generic Ltr 89-13 Re Svc Water Sys Problems Affecting safety-related Equipment.Util Plans to Augment Existing Programs or Implement New Programs to Meet Intent of Generic Ltr ML20005F9341990-01-10010 January 1990 Offers No Comments Re SALP Repts 50-321/89-22 & 50-366/89-22 Dtd 891205 ML20005E6491990-01-0202 January 1990 Responds to NRC 891208 Ltr Re Violations Noted in Insp Repts 50-321/89-30 & 50-366/89-30.Corrective Actions:Util Personnel Documented Engineering Judgment Used as Basis for Use of Agastat Relays in Question ML20005E5621989-12-28028 December 1989 Certifies That fitness-for-duty Program Meets 10CFR26 Requirements.Util Screens for Two Addl Substances Not Required by Rule,Benzodiazepine & Barbiturates.List Re Panel & Cutoff Levels Encl ML20005E1411989-12-28028 December 1989 Responds to Generic Ltr 89-10, Motor-Operated Valve Testing & Surveillance. Thermal Overloads on Most safety-related motor-operated Valves Are Jumpered During Operation.Epri Developing Program to Calculate Valve Thrust Requirements ML20005D9611989-12-22022 December 1989 Forwards Rev to Physical Security Plan.Rev Withheld (Ref 10CFR73.21) ML20011D8721989-12-21021 December 1989 Responds to NRC Bulletin 89-002, Stress Corrosion Cracking of High-Hardness Type 410 Stainless Steel Internal Preloaded Bolting in Anchor-Darling S350W.... Review of Sys Drawings Determined That No Subj Valves Installed at Facilities ML19332G0371989-12-13013 December 1989 Summarizes Util Plans to Recaulk & Seal Plant Refueling Floor Precast Concrete Panel Walls,Per 891129 Telcon. Special Purpose Procedure Developed to Ensure That Containment Integrity Maintained During Recaulking ML19332G0201989-12-12012 December 1989 Forwards Addl Info Re Use of Code Case N-161 for Upgrading Ultrasonic Insp & Testing Instrument Calibr Blocks ML19332F3571989-12-0707 December 1989 Provides Feedback on NRC Pilot Project Involving Electronic Distribution of NRC Generic Communications.Sys Found to Be Most Useful Re Generic Ltrs & Bulletins Where Timely Receipt Critical ML19332E1521989-11-29029 November 1989 Responds to NRC 891101 Ltr Re Violations Noted in Insp Repts 50-321/89-19 & 50-366/89-19.Corrective Actions:Procedure 31GO-INS-001-OS Revised to Include Requirements to Record & Compare Valve Stroke Times Following Valve Maint ML19332D0921989-11-22022 November 1989 Responds to Generic Ltr 89-21 Re Status of Implementation of USI Requirements.Closure Plan for USI A-10 Will Be Submitted in 1990.Response to USI A-47 Re Safety Implications of Control Sys Will Be Submitted in Mar 1990 ML19332E4451989-11-21021 November 1989 Certifies That Initial & Requalification License Operator Training Programs at Plant Accredited & Based on Sys Approach to Training,Per Generic Ltr 87-07 ML19327C2451989-11-13013 November 1989 Forwards Amend 13 to Indemnity Agreement B-69 ML19332B9461989-11-10010 November 1989 Forwards Updated Chronological Tabulated List of Outstanding Licensing Requests for Plant.List Identifies Priority Items for Early NRC Approval ML19327C0321989-11-0606 November 1989 Advises That No Corrections Necessary Re 890331 Response to NRC Bulletin 88-010,Suppl 1.Documentation Available at Plant Site for Review ML19325E8821989-11-0101 November 1989 Responds to Generic Ltr 89-07, Power Reactor Safeguards Contingency Planning for Surface Vehicle Bombs. Contingency Plan Developed Which Has Been Added to Security Plan to Include short-term Actions Against Attempted Sabotage ML19324B8741989-10-27027 October 1989 Transmits Proposed Program for Completing Individual Plant Exam Process,Per Generic Ltr 88-20 & NUREG-1335.Program Should Identify Method & Approach Selected for Performing Exam ML19325E5491989-10-27027 October 1989 Submits Update on Lighting Observed During NRC Insp on 891002-06.All Temporary Lighting Reinstalled.Mfg of Four Cluster Lights,Holophane,Has Been Onsite & Will Give Recommendations for Permanent Lighting ML19327B6151989-10-24024 October 1989 Responds to Generic Ltr 89-16, Hardened Vent, by Encouraging Licensees to Voluntarily Install Hardened Vent Under 10CFR50.59 ML19327B3001989-10-23023 October 1989 Documents NRC Agreement W/Util Justification for Use of Pathway Corp as Replacement Bellows Vendor,Based on 891004 Telcon.Util Proceeding W/Procurement of Replacement Bellows ML19327B1551989-10-17017 October 1989 Forwards Revs 0 to Corporate Emergency Implementing Procedures,Including HNEL-EIP-01,HNEL-EIP-02,HNEL-EIP-03, HNEL-EIP-04,HNEL-EIP-05,HNEL-EIP-06,HNEL-EIP-07,HNEL-EIP-08, HNEL-EIP-10 & HNEL-EIP-11 ML19325C7451989-10-11011 October 1989 Advises That Effective 890913 Th Hunt No Longer Employed by Util.Operator License Terminated ML20248H3061989-10-0404 October 1989 Forwards Revised Tech Specs to Util 890622 Application for Amends to Licenses DPR-57 & NPF-5,per NRC Request,Re cycle- Specific Parameter Limits ML20247G4631989-09-14014 September 1989 Responds to NRC Re Violations Noted in Insp Repts 50-321/89-08 & 50-366/89-08.Corrective Actions:Procedure Revised to Include Periodic Analysis of Fuel Oil Parameters & Change Sampling Methodology ML20246D4541989-08-22022 August 1989 Forwards Corrected Tech Spec Changes Re Reactor Protection Sys Instrumentation Surveillance Requirements,Per NRC Request 1990-09-04
[Table view] |
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,o Georgia Power t ,ec o ,. o ,p vi m nt SL-4431c 1995C X7GJ17-H110 May 4, 1988 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Hashington, D. C. 20555 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 REQUEST FOR ADDITIONAL INFORMATION ON ILJLILLETIN 85-03 Gentlemen:
By letters dated October 2, 1986, and March 12, 1987 Georgia Power Company (GPC) submitted its response to IE Bulletin (IEB) 85-03, "Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings," for Plant Hatch Units 1 and 2. On April 4, 1988, the NRC transmitted a request for additional information (RFAI) on GPC's IEB 85-03 program. Our responses are contained in Enclosure 1, and a detailed program description is contained in Enclosure 2.
During the recently completed Unit 2 Refueling / Maintenance outage, GPC completed static HOV testing on all the applicable High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) valves. (High delta pressure tests are planned.) Per the bulletin requirements, GPC had planned to submit a written request within 60 days of completion of the Unit 2 program. However, as a result of the RFAI, GPC is providing this interim submittal. In addition, GDC would like to schedule a meeting with appropriate NRC and Region II personnel to discuss the details of our program. As you know, we have elected to use a Motor Actuator Characterizer (MAC) system from Limitorque for signature analysis rather than the Motor-Operated Valve Actuation Testing System (HOVATS). Additionally the torque switch settings have been determined utilizing the Limitorque valve equations. GPC realizes that the NRC may be more familiar with MOVATS, and a discussion of our experience with the limitorque equipment and methodology can be included in the meeting agenda,
'f8 I\
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~eso5160001 880504 \
PDR ADOCK 05000321 0 DCD
GeorgiaPower A U.S. Nuclear Regulatory Commission May 4, 1988 Page Two He hope the enclosed material will be helpful in your review. You may contact this office if you have questions. I i
Sincerely, i 2 $5 R. P. Mcdonald Executive Vice President Nuclear Operations GKM/ac/lc
Enclosures:
- 1. Response to RFAI on IE Bulletin 85-03.
- 2. IE Bulletin 85-03 Program Description.
c: Georgia Power Comoany Mr. J. T. Beckham, Jr., Vice President - Plant Hatch '
Mr. L. T. Gutwa, Manager Nuclear Safety and Licensing GO-NORMS U . S . N u c 1 e aLRegulltory_Commi s s i on . Huhing t o n . D . C1 Mr. L. P. Crocker, Licensing Project Manager - Hatch U.S. Nuclear Rf9211 tory Commistion. Region _H Dr. J. N. Grace, Regional Administrator Mr. P. Holmes-Ray, Senior Resident Inspector - Hatch 1995C wn
L
. GeorgiaPower A i
ENCLOSURE 1 l
PLANT HATCH - UNITS 1, 2 !
NRC DOCKETS 50-321, 50-366 !
OPERATING LICENSES DPR-57, NPF-5 l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RFAI) - IE BULLETIN 85-03 The following is a response to the NRC request for additional information !
on IEB 85-03 (Alan R. Herdt to George F. Head) dated April 4, 1988. !
i Question 1. ;
Revise the summary tables of the response dated March 12, 1987, to I include values of differential pressures for opening the following MOVs, 1 or justify exclusion of these pressures. As required by Action Item "a" l of the bulletin, assume inadvertent equipment operations. ;
(a) HPCI MOV F004 is shown normally open in Zone 0-9 of Drawing H-16332, Revision 21, and as H0V 3 on Page 68 of BHROG Report :
NEDC-31322 dated September 1986. How would suction from the !
CST be ensured if this H0V were to be (a) actuated i inadvertently to the closed position upon intended initiation !
of the system or (b) left closed inadvertently? .
(b) RCIC HOV F010 is shown normally open in Zone 0-0 of Drawing '
H-16334, Revision 16, and as H0V 3 on Page 72 of the BHROG Report. The question in Item 1(a) above applies here also. .
(c) HPCI HOV F007 is shown normally open in Zone E-5 of Drawing :
H-16332, Revision 21, and as MOV 8 on Page 68 of the BHROG i Report. How would discharge to the reactor vessel be ensured !
if this MOV were to be (a) actuated inadvertently to the closed [
position upon intended initiation of the system or (b) left '
closed inadvertently? r (d) RCIC MOV F012 is shown normally open in Zone E-6 of Drawing ,
H-16334, Revision 16, and as HOV 8 on Page 72 of the BWROG (
Report. The question in Item 1(c) above applies here also, j l
Resoonse to Question 1: !
i The differential pressure calculations for the four valves which consider l' valve mispositioning errors are not available to provide to the NRC at this time. The BWR Owner's Group (BWROG) report (Reference 1) concluded i that the inadvertent valve operation information was beyond the design !
basis of the plant. As such, the BHROG methodology for calculation of '
1995C El-1 05/04/88 SL-4431 t
""6 , - - - - . . - . - - _ . _ - . _ , , _ ._ - _ - _ - . . - _ . . .
Georgia Powerkh ENCLOSURE 1 (Continued)
BESPONSE TO RFAI - IE BULLETIN 85-03 differential pressure was not developed. If required. Georgia Power Company would use this methodology to calculate the Hatch-specific differential pressure as we did for the valves in our March 12, 1987, submittal.
He understand that the question of whether or not valve mispositioning should be considered under this program is still under discussion and that a revision to IEB 85-03 may be issued shortly. Although GPC may elect to follow the BWROG "position" delineated in the Reference 2 letter, we strongly disagree with any precedent which causes us to consider events beyond the plant design basis in this IEB 85-03 program.
Question 2:
Revise the RCIC summary table of the response dated March 12, 1987, to include Trip and Throttle Valve F524 leading to the RCIC Turbine, or justify its exclusion.
This HOV is shown in Zone D -3 of Drawing H-16335, ,
l Revision 11 for Unit 1, and as HOV X on Page 74 of the BWROG Report. Is this MOV meant to be identified with a number in Zone C-9 of Drawing H-26024, Revision 13 for Unit 2? Assume inadvertent equipment operations, as described in Item 1 above.
E11ponse to Question 2:
Valve number X in Reference 1 functions as the RCIC turbine trip and throttle valve. The active safety function of the RCIC turbine trip and throttle valve is to trip closed when required to protect the turbine and ,
l pump. The closure of the valve, when tripped, is spring actuated. The motor operator on this valve is only used to reset the valve to the open position following a turbine trip.
The differential pressure across the RCIC turbine trip and throttle valve during opening is negligible. The basis for this is that, prior to resetting the RCIC turbine trip and throttle valve, the RCIC system steam admission valve located upstream of the trip and throttle valve would first be closed. This action resets the RCIC system startup logic (i.e.,
the ramp generator for the RCIC turbine). The RCIC turbine trip and throttle valve above the seat drain upstream of the valve will vent steam that is trapped between the closed steam admission valve and the trip and throttle valve to the turbine exhaust line drain pot. This will reduce the differential pressure across the turbine trip and throttle valve to a negligible value prior to valve opening.
Unit 2 drawing H-26024 does not show the master part list (HPL) number for MOV X and will be corrected.
1995C El-2 05/04/88 SL-4431 m,
Georgia Powerkh ENCLOSURE I (Continued)
RESEORSE TO RFAI ___IE_ BULLETIN 85-03 Question 3:
Revise the summary tables of the response dated March 12, 1987, to include values of differential pressure for opening and closing the following MOVs, or justify exclusion of these pressures. According to Pages 55 and 59 of the BWROG Report, these CST test return valves have no safety action; however, utilities are expected to report differential pressures for testing, per Note "o" on Page 66.
(a) HPCI MOVs F008 and F0li are shown normally closed as MOVs 5 and 6 on Page 68 of the BWROG Report.
(b) RCIC MOV F022 is shown normally closed as MOV 5 on Page 72 of the BHROG Report, gesponst_t2_0y1511on_3:
The plant design basis evaluations for transient and accident responses analyzed in the FSAR assume the HPCI and RCIC systems are in their normal standby condition at the start of the event. This assumption is made because of the low probability of the system being in a test mode or out of service during the occurrence of an abnormal event. Based on this assumption, the HPCI MOVs F008 and F0li and the RCIC MOV F022 perform no active safety function during FSAR design basis events. These valves would not be included as part of GPC's IEB 85-03 program, especially since they are not among the nine valves of question relative to the mispositioning issue. (See response to Question 1.) Nevertheless, the Unit 2 summary tables from the March 12, 1987, submittal have been revised to includt these calculations and appear at the end of this enclosure. Unit I results will be similar and will be available at a l
later date. It should be noted that operation of these valves during system flow tests demonstrates their capability to operate against differential pressures that occur during testing.
Question 4:
The proposed program for action items b, c and d of the bulletin is incomplete. Provide tha following details as a minimum:
(a) commitment to justify continued operations of a valve determined to be inoperable, 1995C El-3 05/04/88 l
SL-4431 m'
7 i 4 Georgialbwer A ENCLOSURE I (Continued)
RESPONSE TO RFAI - IE BULLETIN 85_03 (b) description of a method possibly needed to extrapolate valve stem thrust determined by testing at less than maximum differential pressure, (c) justification of a possible alternative to testing at maximum differential pressure at the plant.
(d) consideration of pipe break conditions as required by the bulletin, and (e) description of program for selection of switch settings (i.e.,
torque bypass,. position limit, overload) for valve operation.
Resoonse to Ouestion 4:
(a) Much of the static HOV testing of the HPCI and RCIC valves was performed during the Unit 2 Refueling / Maintenance outage and similar testing is planned for Unit I during the Fall 1988 cutage. During the outage, HPCI and RCIC would not be required; therefore the valve could be instrumented and tested in the as-found condition and set up within IEB 85-03 program specifications without having to enter a limiting condition - for operation (LCO) for the applicable system.
For bulletin valves tested prior to shutdown (or for selected valves which will be tested at high differential pressure), the HPCI or RCIC system was (will be) declared inoperable and the LCO entered.
The valve was (will be) adjusted. If necessary, so the as-left condition was within program specifications and the system returned to service. In the very unlikely event that the valve could not be mace operable within the allowable outage time (A0T), GPC would probably pursue alternate paths of justifying continued operation (i.e., by analysis, repair, or replacement). If these efforts were unsuccessful and the LCO expired, appropriate actions per Technical Specifications would be taken.
(b) See Enclosure 2.
(c) See Enclosure 2.
(d) Consideration of pipe break conditions was included in the differential pressure calculations reported in our NRC submittals.
They were also factored in the establishment of permissible torque and thrust settings. (See Er. closure 2.) Georgia Power has no plans 1995C El-4 05/04/88 SL-4431 rwn _
Georgialbwer b ENCLOSURE 1 (Continued)
RESf0NSE TO RFAI - IE BULLETIN 15-03 to test these valves at full differential pressure, but will perform a static MOV test and adjust the required settings per the Limitorque equations relative to the maximum differential pressure expected during accident conditions.
(e) See Enclosure 2.
l l
1995C El-5 05/04/88 SL-4431 rxm
Georgia Power d ENCLOSURE 1 (Continued)
RESEQMSE TO RFAI - ILMLLETIN 85-03
REFERENCES:
- 1. NEDC-31322, "8HR Owners Group Report on the Operational Design P sis of Selected Safety-Related Motor Operated Valves," September, 198(.
- 2. Letter, R. F. Janecek (BWROG) to J. H. Sniezek (NRC), same subject, dated March 28, 1988.
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GeorgiaPower d ENCLOSURE 2 PLANT HATCH - UNITS 1, 2 NRC DOCKETS 50-321, 50-366 OPERATING LICENSES DPR-57, NPF-5 LE BULLETIN 85-03 PROGRAM DESCRIPTION Introduction IE Bulletin 85-03 addresses potential problems associated with switch settings on certain safety-related motor-operated valves. The bulletin was issued in response to several events at operating nuclear plants in which safety-related valves failed to operate due to improper switch settings. The bulletin requests each plant to implement a program to set and maintain certain MOV switches to ensure valve operability under maximum differential pressure conditions.
The program developed for Plant Hatch can basically be divided into four sections which correspond to Action Items a, b, c, and d of tha Bulletin. These sections include the following:
- 1. Identification of Valves and Determination of Maximum Differential Pressure.
- 2. Establishment of Correct Switch Settings.
- 3. Switch Adjustment and Demonstration of Operability at Maximum Differential Pressure.
- 4. Development of Procedures to Ensure that Switch Settings are Maintained for the Life of the Plant.
Each section of the program, along with its current status, is outlined t,el ow.
I. Identi fi rJtion of Valves and Determination of Maximum Differential Pressure The Hatch response to Action Item "a" is based on the "BHR Owners Group Report on the Operational Design Basis of Selected Safety-Related Motor-0perated Valves." This report identified the BHR valves covered by the bulletin and outlined a methodology for calculating the maximum differential pressure for each valve. For Plant Hatch, the bulletin covers a total of 22 valves on Unit I and 23 valves on Unit 2, all contained within the HPCI and RCIC systems.
1995C E2-1 05/04/88 SL-4431c 700775
GeorgiaPower A ENCLOSURE 2 (Continued)
IE BULLETIN 85-03 PROGRAM DESCRIPTION This section of the program has been ctopleted, and an initial submittal identi fying the bulletin valves and their maximum differential pressure was transmitted to the NRC on October 2, 1986. This submittal was subsequently amended on March 12, 1987, to revise the maximum differential pressure for certain valves.
II. Establishment of Correct Switch Settinas This section of the program can be divided into three parts, including: A. Switch Functional Review, B. Calculation of Required Opening and Closing Thrust, and C. Engineering Evaluation of Operator Capability.
A. Switch Functional Review The elementary wiring diagrams for each of the valves was reviewed to determine the design function of each of the switches contained in the Limitorque operator. In addition, the applicable maintenance procedures were reviewed to determine the current method for setting each of the switches. The results of this review are summarized below.
- 1. Open Torque Switch - This switch is not utilized in the control circuit of any bulletin valves.
- 2. Close-to-Open Torque Bypass Switch - This switch is not required because the open torque switch is not utilized in the control circuits of any bulletin valves.
- 3. Open Limit Switch - This switch is utilized to trip the operator in the opening direction. The switch is adjusted to ensure that inadvertent backseating due to inertia does not occur.
- 4. Close Torque Switch - This switch is utilized to trip the operator in the closing direction. The switch setting is currently based on manufacturer recommendations.
- 5. Open-to-Close Torque Bypass Switch -
This switch is utilized to bypass the close torque switch during the initial 1/8 in. of travel in the closing direction.
1995C E2-2 05/04/88 SL-4431C
""5
GeorgiaPower A ENCLOSURE 2 (Continued)
IE BULLETIN 85-03 PROGRAM DESCRIPTION 6.
Close controlLimit Switch circuit of any- bulletin This switch is not utilized in the valves.
- 7. Motor Overloads - The motor overloads are jumpered out during normal plant operation on all bulletin valves.
The motor overloads are in service only durinp routine surveillance and post maintenance testing.
The switch functional review indicated that the design '
philosophy of the operator trip scenario was sound and that it afforded the valve the maximum opportunity to perform its intended function. The significant switches, with regard to valve operability, are the open limit switch and the close torque and torque bypass switches. The setpoints for the open limit switch and the close torque switch are delineated in the applicable maintenance procedures. The close torque i switch settings are currently based on information provided by the valve and operator manufacturers and are in terms of switch position rather than engineering units.
\
In order to evaluate the adequacy of the torque swit h '
settings and operator sizing, it was necessary to know the minimum required thrust to open and close each valve under the maximum differential pressure. Standard Limitorque equations were used to revalidate the sizing and settings for full differential pressure. Although documentation on the valve and operator manufacturers' original torque switch setpoints was available, the detailed engineering l
' calculations sometimes were more difficult to locate. In addition, the origiral thrust calculations were based on i system design pressures rather than the actual expected maximum differential pressures. It was determined that the required thrusts to open and close each valve under its maximum differential pressure should be recalculated to 3
verify operator sizing and torque switch setpoints.
B. Calculation of Required Opening and Closing Thrust In order to calculate the required opening and closing thrusts, it was necessary to obtain data on the physical characteristics of the valves. This information was not always available on drawings or in vendor manuals and required going back to the original equipment manufacturer.
Each valve vendor was contacted and requested to provide the following information:
1995C E2-3 05/04/88
- SL-4431c 4
700776
7
- Georgia Power d ENCLOSURE 2 (Continued)
IE BULLETIN 85-03 PROGRAM DESCRIPTION
- Seat ring diameter.
- Disk or plug coefficient of friction.
- Stem efficiency. ,
j
- Stem diameter, pitch, and lead.
Maximum allowable torque.
These data were used to calculate the required opening and closing thrusts for each valve. These calculations were performed utilizing the standard Limitorque empirical equations for gate and globe valves. These calculations provide the minimum required thrusts to open and close each valve. The next step was to perform an engineering evaluation to determine whether each operator was capable of {
providing the required thrust.
C. Engineering Evaluation of Operator Capability A complete review of the Limitorque operator for each valve was performed to determine the maximum torque rating of each operator.
This included a review of operator capability at derated voltage (90-percent voltage for AC and 84-percent voltage for DC). This review verified that the original operator setting was sufficient to operate the valves against the maximum differential pressure and also established an upper limit above which the valve or operator could sustain mechanical damage.
With this section of the program complete, information became available regarding the minimum torque required to open and close each valve and the maximum allowable torque to avoid damage to the valve or operator. This information (in the form of "target" ranges of permissible torque and thrust) has been utilized to evaluate switch settings in the field and to adjust switches, as required to ensure operation at maximum differential pressure without damage to the valve or operator. Table 1 contains a summary of the MOV calculations for Unit 2. Calculations for Unit 1 will be available at a later date.
III. Switch Adiustment and Demonstrate Ooerability at Maximum Differential Pressure This section of the program can be divided into three parts, including: A. MOV Diagnostic Testing, B. Signature Analysis, and C. Differential Pressure Testing.
E2-4 05/04/88 l - 31c l
l wn
. GeorgiaPower A ENCLOSURE 2 (Continued)
IE BULLE'IN 85-03 PROGRAM DESCRIPTION A. MOV Diagnostic Testing on Unit 2 In order to veri fy the operability of each valve, the Limitorque Motor A:tuator Characterizer (MAC) was utilized.
This system allows the following parameters to be monitored and recorded during the respective closing or opening stroke of the valve:
- Stem thrust.
Output torque.
- Motor current.
- Spring pack displacement. '
Open torque switch.
- Open torque switch bypass.
- Open limit switch.
- Close torque switch.
Close torque switch bypass.
- Close limit switch.
Valve signatures were taken in the ts-found condition in ,
accordance with procedures. These data were evaluated in conjunction with the calculated data in Section II. If an adjustment was required, a new set of signatures was taken to verify that the setting.; were correct and to provide a record of the as-left condition.
B. Signature Analysis The static signatures taken with the MAC equipment were
- analyzed to evaluate each valve's operability at maximum
! differential pressure. The static signatures quantified all valve loads other than those due to differential pressure in terms of thrust and/or torque. In addition, the status of the torque and limit switches was monitored with respect to time and was directly related to thrust and torque at the i
switch trip points. By comparing the thrust at the torque
, witch trip point to the calculated required thrusts to open and close the valve, a determination of valve operability at maximum differential pressure was made. If the thrust at the torque switch trip point was determined to be greater l than the required opening and closing thrust, the valve will l be capable of operating against the maximum differential pressure.
l 1995C E2-5 05/04/88 i
SL-4431c
!=m. - - . . - ... . - _ -- - -
Georgia Power A -
ENCLOSURE 2 (Continued)
IE BULLETIN 85-03 PROGRAM DESCRIPTION C. Differential Pressure Testing A differential pressure test will be conducted on the steam admission valves (2E41-F001 and 2E51-F045) of the Unit 2 HPCI and RCIC systems. Unit 1 testing will be completed at a later date. The purpose of these tests will be to help validate the methodology outlined in Section III.B for determining valve operability at maximum differential pressure. The steam admission valves were chosen due to their accessibility, high differential pressure, and the fact that these valves are stroked during system operability tests.
Each of the above valves will be outfitted with the MAC diagnostic equipment and stroked during a system operability test. The data obtained during this test will be evaluated in conjunction with the calculated thrust values to ensure that the Limitorque equations are providing sufficiently conservative results.
IV. Develooment of Procedures to Ensure that Switch Settinas are Maintained for the Life of the Plant ,
Existing plant procedures are presently being reviewed to identify changes necessary to ensure that switch settings are maintained for the life of the plant. A policy regarding periodic and/or maintenance-related retesting will have to be formulated prior to identifying all procedural requirements.
- This policy and the procedural changes necessary to implement it i will be initiated prior to making our final submittal.
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i 1995C E2-6 05/04/88 SL-4431c wn . _ _ - . - . __ . - - _ _ . . . _. ..
1 Enclosure 2 (continued)
TABLE 1 l
l LMIT 2 IfN SIAliARY OF COO.USIONS l l
RATED RE @ REW WTR TARGET tut 0)E TAICET T2tlST LIWITl2 PAAAETER VALVE ETOR WTOR TORQUE RANGE RAEE FCR )
DESCRIPTION TORO uneneneenonennennunnensee no LE TCRCLE VAL.V.EIPL
- e. nemennenuneeKRATED WINIH WAXIM WINIM WAXil&Al LPPER TARGE unnuennenennunenneennununeeeneen (FTLB)(FTLB) (LB) nunnenunneennenununununn enouenenben(FTLB) seene(FTLB) (FTLB) (LB) ennenenenenunnunnuenenenununennu '
Z414001 TWBINE STEAW SLPPLY 80 48.18 57.36 471.75 652.00 23843 32992 OPER W AVAll TOROVE Z414002 STEAW SUPPLY 120 W ISOLATION 25 20.52 25.33 472.12 639.20 23880 32305 OPER MAX AVAll TORM 2E414003 STEAW SLPPLY OllTBOARD ISCLAtl0H 40 21.80 25.96 472.12 850.00 23860 42958 OPER WM ALLOW TOR 1!
2E414304 PLAP SUCTION CST 15 4.12 4.91 91.46 361.84 4137 16727 OPER WM AVAll TORM 2E414006 PtAP 120ARD DISCmRGE 150 35.36 42.10 498.76 2200.00 19703 88908 VALV Wu A'.' 5 TCRM 2E414012 WIN FLOW BYPASS 25 7.39 8.79 123.M 342. M 8824 24000 OPER Wu A',LC# TWST Z414041 PtnP SUCTION Ff0d SLPPRESION POOL 15 8.27 9.85 173.18 348.78 7832 15775 OPER Wu AVAll TORM E41-F042 PLAP SUCTION FRCnl StPPRESION POOL 15 8.27 9.85 173.18 348.78 7832 15775 OPER Wu AVAll TOR 72 2E414059 COOLING IIATER SLPPLY 10 0.62 0.74 10.48 135.E2 1060 14000 OPIR Wu Al. LOW ThRUSi l.
2E414104 VAQAll BREAKER ISOLATION 2 0.35 0.43 8. 5*. 54.93 1015 6537 CPER Wu AVAll TORM 2E414111 VA3AAi BREAKER ISOLATION 2 0.35 0.43 8.53 54.93 1015 6537 OPER WM AVAll TORM 2514007 STEAW INBOAA0 ISOLATION 10 6.71 8.29 83.82 105.32 5618 9300 OPER MAX AVAll TOR M i
2514008 STEAM OUT20JA0 ISOLATICN 10 6.27 7.46 63.82 112.80 5618 9961 CPER W.a AVAll TORM 2E51-F010 PLAP SUCTION CST 5 1.53 1.82 26.32 fr.00 1751 5989 CPER W ALLCW TORM 2E514013 PLnP ISOLATION lh3 CARO 15 7.75 9.22 78.88 189.20 5752 12373 OPER WA AVAll TOR M 2E514019 TEST BYPASS TO CST 25 5.44 6.48 37.23 135.82 3843 14CCO OPER WM All0W THR.'ST E514029 PtnP SUCTION StPPRES$10N POOL 5 2.21 2.63 34.74 87.50 2312 5822 OPER KAX AVAll TOR 1I 2E514C01 PthP SUCTION STPRESSION POOL 5 2.21 2.83 34.74 87.50 2312 5822 OPER WAX AVA!L TORM 2E5i4045 TWBINE STEAW SLPPLY 40 15.30 18.21 192.52 295.26 15649 24000 OPER WAX ALLOW T R ST I
2E514046 (COLIN 0 RATER SLPPLY 10 1.40 1.67 23.58 135.62 2434 14000 CPER W ALLCW TR4T l
2E514104 YAD.AAI BREAXER ISOLAtlCN 2 0.36 0.44 7.31 45.00 1009 0293 CPER MAX AVAll TCRM 2E514105 YAD.I.Al BREAKER ISOLATich 2 0.36 0.44 7.31 45.00 1009 8293 CPER lLAX AVAll TORM 2E514119 STEAM ADilS$10H BYPASS 5 1.50 1.78 13.19 44.99 1625 0037 OPER WAX AVAIL TORM l.
l E2-7 5/4/88 l
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