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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] |
Text
>. .
o Georgis Power Company i t 40 inr,rness Center Parkway
'- Post Offica Box 1295
- Birmingham. Alabama 35201 '
, f ..- Telephone 205 877-7279 '
J. T. Beckham, Jr. . Georgia Pbwer . ,
24, Vice President . Nuclear g March 14, 1995 -
'l Docket Nosi50-321 HL-4794
'50-366' l
U. S. Nuclear Regulatory Commission d
~ ATTN: Document Control Desk .
Washington, D. C. 20555 Edwin I. Hatch Nuclear Plant Units 1 and 2 Response to Request for Additional Information:
Generic Letter 89-10 l
Gentlemen:
- In response to your letter dated February 13,1995, " Request for Additional Information
.Regarding Scope of Generic Letter 89-10 for Edwin I. Hatch Nuclear Plant Units 1
]q and 2," Georgia Power Company (GPC) submits the following responses to your '
questionsin Enclosure 1.
If you have questions regarding this matter, please contact this office.
Sincerely, f
UN N
. T. Beckham, Jr.
OCV/eb
Enclosure:
Request for Additional Information - Generic Letter 89-10 cc: Georgia Power Company Mr. H. L. Sumner, Jr., Nuclear Plant General Manager NORMS LI. S. Nuclear Regulatory Commission. Washington. D. C.
Mr. K. Jabbour, Licensing Project Manager - Hatch U S. Nuclear Regulatorv Comminion. Region 11 Mr. S. D. Ebneter, Regional Administrator Mr. B.' L. Holbrook, Senior Resident Inspector - Hatch 9503210526 950314 > ! -
PDR ADOCK 05000321 -
L P PDR i t
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Oh
.. Enclosure 1 I Edwin I. Hatch Nuclear Plant Request for AdditionalInformation:
Generic Letter 89-10
' 1. Plant Service Water Dilution Line Isolation Valves (P41-F312 for Unit 1 and P41-F310 for Unit 2)
This valve has an active safety function to automatically close. However, its normal position is closed.
Discuss your procedural requirements to ensure that this valve is closed at all times )
and how you ensure that it would not open.
RESPONSE
Both valves for Units 1 and 2 are listed in the valve line-up section of their respective System Operating Procedure as closed. Furthermore, to ensure the valves are not opened by inadvertent operation of the control switch, the system line-ups in these procedures also require that the breakers be turned to the "off' position.
(Ref. 34S0-P41-001-IS and 34SO-P41-001-2S.)
Under the previous guidance, before our re-evaluation was completed early last year, these valves had active safety functions to close. However, the valves were reclassified as having no active safety function since they are closed and remain closed for all design basis events. These valves were designed to be used for the ,
dilution of radwaste with Plant Service Water (PSW). However, radwaste dilution flow, for both Hatch Units 1 and 2, is provided via Circulating Water System valves, therefore, the PSW valves remain closed. These valves are opened during logic .
system functional testing, performed once per eighteen months for each unit.
However, per criterion no. 3 (Reference enclosure 3 of 2-3-94 submittal) of our safety function guidelines, valve operability for GL 89-10 purposes is not required !
during component testing.
- 2. Reactor Core Isolation Cooling (RCIC) Valves (E51-F013 and E51-524)
Tojustify removal of the above valves, confirm that they do not have to cycle during a DBA to initiate, isolate, or re-initiate RCIC operation assuming the RCS becomes .
isolated. This should consider both original initiation of the system as well as j recovery from RCIC isolation on high vessel level as the system cycles to maintain ]
RCS level.
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Generic Letter 89-10 RESPONSE .!
Re-injection with RCIC is desirable but not necessary since inventory makeup can be i
accomplished by vessel depresurization using ADS followed by vessel makeup using either LPCI or core spray.
The RCIC injection valve (E51-F013) is normally closed and receives an open signal I on an RCIC system initiation signal. Accordingly, the valve has an active function to open.
Credit is taken for the RCIC system for design basis transient events involving loss of feedwater. The RCIC is not considered an emergency core cooling system (ECCS); 4 therefore, it is not assumed to be available for the loss of coolant accident (LOCA).
The RCIC operation for design basis transient events is initiated on low water level.
Consistent with the requirements of GL 89-10, the steam admission valve (E51-F045) )
and injection valve (E51-F013) are required to open to permit initiation of RCIC. ]
The RCIC is tripped on high water level. The high water level trip signal closes the !
steam admission valve (not E51-F524) and, following closure of the steam admission . .;
valve, the injection valve is closed. It should be noted that the event has been turned 1 around (core re-flooded with pressure stabilized) prior to RCIC being secured.
Subsequently, only inventory makeup necessary to compensate for steaming is required. Numerous systems, such as HPCI, CRD, Feedwater and ADS followed by low pressure system injection, can accomplish this function.
- 3. Outboard Low Pressure Coolant Injection (LPCI) Valve (E11-F017 A & B) k The staff questions the basis for removal of the above valves from the GL 89-10 ,
program, because they may have to perform an active safety function in response to a i DBA. If one or both of these valves are closed to direct flow to containment cooling, they must be reopened to restore LPCI to the vessel for long-term accident recovery.
The February 3,1994, submittal stated that ifone valve does not open, the other loop _
could be opened to provide injection to the reactor pressure vessel (RPV). This assumption does not account for the possibility that both valves could malfunction due to a common cause defect. If these valves do not need to function in response to a design-basis event, then you should demonstrate that the restoration of RPV flow is not safety significant.
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. Request for AdditionalInformation: , i Generic Letter 89-10 ,
RESPONSE
A common cause defect which simultaneously affects both valves in the "A" and "B" loop is beyond the licensing basis of the plant. ' Initial core cooling and core 3 reflooding for the design basis LOCA is accomplished by the operation of LPCI and core spray systems, which involves the opening of the inboard injection valves.
E Adequate core cooling can be accomplished assuming the limiting single failure.
l Diversion to suppression pool cooling is not needed until approximately 30 minutes .
post DBA LOCA. At this point in time, the core has been reflooded, and only the make-up necessary to maintain water level is required. Any single low pressure ECCS system is capable of providing makeup.
Accordingly, Georgia Power Company does not intend to include the El1-F017 ,
valves in the scope of the GL 89-10 program. 4
- 4. Test Valves -
It appears that the category of safety system test valves is absent from your GL 89-10 program. These valves co defeat safety system response when in the test position.
Many of these valves receive auto reposition signals upon safety injection. Some safety systems are assumed to remain operable even when in test position due to the capability of these test valves to return to their " safety position" under DBAs. Since -
repositioning of these valves are relied upon to justify the availability of various safety system responses, they should be included in the GL 89-10 program. l
RESPONSE
i System test valves are not included in the GL 89-10 program because these valves, consistent with Technical Specifications philosophy, are not assumed to be called upon to operate during the short periods of time the system is in test.
Criterion 3 of the GPC submittal of February 3,1994, submittal states, " Valve operability requirements, as defined by GL 89-10, is not required during periods of :
system or component testing." This is consistent with the reliability assumptions that form the basis for the Unit I and Unit 2 Technical Specifications. This criteria is ,
consistent with the original GE specifications for the ECCS which states that it is not necessary for the test valves to re-align within the accident response times. This approach is considered acceptable because the period of time that standby systems are in the test mode during normal plant operation is small in comparison with the time ,
that the systems are in the standby mode of operation during normal plant operation. l Thus, these valves were not intended to recover from the test mode in a timely HL-4794 El-3 l
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. Enclosure 1
.- Request for AdditionalInformation:
+ ^ Generic Letter 89-10 l
manner to mitigate the consequences of design basis accidents. . Consequently, current design basis event analyses do not consider an additional single failure for l event mitigation when a system is in the test mode in accordance with Technical ~
Specification testing requirements. Therefore, GPC does not intend to place these l valves in the scope of the GL 89-10 program. j
. 5. Containment Sorav Valves (E11-F016 A & B. E11-F021 A & B. E11-F027 A & BF Georgia Power's analyses indicate that the drywell and suppression pool spray functions are not needed to maintain the containment below its design temperature and pressure. ' The licensee should also confum that the spray function serves no : i radiological safety function and is not needed for environmental qualification purposes. Based on these analyses and subject to these confirmations, the spray valves, for purposes of GL 89-10, may be considered not to have an active safety function to open. Since the spray valves are normally closed, it is unlikely that they would be required to reposition to the closed position for purposes of containment isolation during an accident. Thus, for purposes of GL 89-10, the spray valves may be considered not to have an active safety function to close.
RESPONSE
No credit is taken for the operation of the containment spray system in the performance of the radiological evaluations for design basis events.
The drywell and wetwell sprays are assumed to function for environmental. ;
qualification purposes.
t However, the spray valves are not required to be considered for GL 89-10 purposes for the following reasons: ;
For design basis events, small steamline breaks inside the primary containment
(<0.5 ft 2) result in the highest containment temperature. For these breaks, only the _ ,
ADS and low pressure injection systems are realistically required to function after ,
containment isolation. The emergency operating procedures require that the drywell sprays be initiated prior to the drywell reaching its design temperature. Ifit is determined that the drywell sprays cannot be initiated, emergency depressurization will be initiated. Thus, if the containment spray valves fail to open, the operators . ,
t would proceed to initiate emergency depressurization of the reactor vessel, directing steam flow to the suppression pool. The ADS components are qualified to higher
< temperatures than the qualification envelope. Further, the ADS qualification envelope includes a duration of about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for temperatures in excess of 330 F, i which is the peak calculated drywell temperature for the limiting break size. Core l
HL-4794 El-4 ,
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-. Enclosura 1 Request for Additional Information:
Generic Letter 89-10 cooling will be accomplished with the injection oflow pressure ECCS systems, _
following the emergency depressurization. Both injection valves for the LPCI and Core Spray systems are located outside the primary containment and are therefore not subjected to the harsh environment inside the drywell.
An evaluation of the small steamline break assuming no sprays with ADS initiation has not been formally done at Plant Hatch. However, a realistic, qualitative, review shows that once ADS is initiated, the driving force for the containment heatup is ,
eliminated prior to exceeding the environmental qualification for qualified components. Although it is possible, that at some point later in the event qualification
- temperatures for some components may be exceeded, the components necessary to .
mitigate the consequences of the event have already performed their functions and are no longer needed. For example, containment isolation valves have closed by the time the ADS is initiated; even in the unlikely event that these valves go open at some later .
time in the scenario, no unacceptable consequences will result since the design basis small/ intermediate steamline break does not result in fuel failures. Consequently, ,
10 CFR 100 limits are not in danger of being exceeded.
In summary, although the drywell and wetwell sprays are assumed in the Plant Hatch environmemal qualification analysis, an assessment of the limiting temperature event :
! in the drywell, shows that the containment spray valves need not be included in the GL 89-10 scope. ;
- 6. Recombiner Valves Since the recombiner valves are normally closed and the recombiners are " closed systems outside containment," it is unlikely that they would be required to reposition ,
to the closed position for purposes of containment isolation during an accident. The recombiner valves, for purposes of GL 89-10, may thus be considered not to have a closing safety function.
The hydrogen recombiner system is provided as a hydrogen control system in the '
Unit 2 licensing basis. From a licensing perspective the recombiners serve a safety
- function and on that basis, the recombiner valves should have an active safety function to open. However, if Unit 2 has a safety-grade vent / purge capability (as does Unit 1) which provides a similar hydrogen control capability using fully-tested ;
valves, the vent / purge capability provides a satisfactory basis for excluding the i recombiner valves from the more comprehensive capability evaluations. >
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.- Enclosure 1
, . Request for Additional Information!
Generic Letter 89 >
RESPONSE .
Unit 2 has a safety grade vent / purge capability.
. As we mentioned in our submittal of February 1994, the recombiner system was
- implemented as a condition for license for Unit 2. GPC does not believe these valves belong in the GL 89-10 program, because per the Nuclear Safety Operational Assessment (NSOA) report, the system is not required to function for any design basis event and tb NSOA is the governing document for defining design basis events -
for our GL 89-10 program. The design basis LOCA event for Hatch does not result-
' in sufficient buildup of hydrogen and oxygen to reach combustible concentrations, thus the recombiners are not necessary for the mitigation of design basis events, and thus, by criteria nos. 4 and no. 5 (Ref. 2-3-94 GL 89-10 submittal) the system is not required to be considered for GL 89-10 purposes.
Unit 2 does have a vent and purge capability for the primary containment and, like Unit 1, vent and purge is the primary strategy for combustible gas control. Both Unit's systems use air operated valves. Per the Hatch System Evaluation Documents, f the post accident vent and purge function for Unit 2 is not safety related. However, primarily due to the containment isolation function, the valves used for post accident vent and purge on Unit 2, as well as Unit 1, are safety related.
Accordingly, including the recombiner valves in the GL 89-10 program is not warranted.
i
- 7. HP_CI valves During a February 22,1994, meeting, GPC's representatives stated that all the high-priority valves will have their active safety functions intact. For Hatch, there are ten valves listed under this category which includes HPCI, RCIC, RHR and RWCU valves. It is not clear how some of these valves can be removed from the program, regardless of their normal positions, given their importance and that the operators may be required to manipulate them to close, open, or throttle during a design basis event such as a Small Break Loss-of-Coolant Accident. Particular' emphasis is placed on the HPCI injection valve which might be required to automatically open or close in response to reactor vessel water level.
RESPONSE
All of the 14 high priority valves for Hatch are included in the GL 89-10 program.
None of these have been removed from the program, and there are no immediate plans to remove any from the scope of GL 89-10.
HL-4794 El-6
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. Request for AdditionalInformation:
Generic Letter 89-10
- Re-injection with HPCI is desirable but not necessary since the vessel can be depressurized using ADS and inventory maintained with any low pressure system. .
As was stated in the February 22,1994, meeting, the active safety function for all 14 high priority valves is included in the GL 89-10 program. Further, these valves are tested in accordance with the Technical Specifications and applicable code requirements to provide additional assurance of their reliability for other functions. -
Thus, these valves are considered highly reliable and would be expected to have a ;
very high probability ofbeiag available for all design basis events.
With respect to the HPCI injection valve, credit is taken for the HPCI system for design basis events involving loss of feedwater and/or a small break LOCA. .The HPCI operation for design basis events is initiated on either low water level (L2) or high drywell pressure. Consistent with the requirements of GL 89-10, the steam admission valve (E41-F001) and injection valve (E41-F006) are required to open to permit initiation of the HPCI. The HPCI is tripped on high water level. Th: high water level trip signal closes the hydraulically operated turbine stop valve and, following closure of the stop valve, the injection valve is closed. If these valves are . :
closed as a result of the trip signal, the GL 89-10 requirement to open provides additional assurance that the injection valve would again open on demand. The closure requirement for the injection valve is not considered necessary for design basis events because the system can be manually tripped if the turbine stop valve does not close on high water level. Should the HPCI system be manually tripped and fail to restart on low water level, its function can be fulfilled by the RCIC system for events involving loss of feedwater and by the ADS for a small break LOCA. In the unlikely event that the RCIC system is not available for events involving a loss of +
feedwater, adequate core cooling, in accordance with the emergency operating procedures, can be provided by the combination of the automatic depressurization system (ADS) and the low pressure core cooling systems, as discussed previously.
A close active safety function is therefore not needed for the HPCI injection valve.
Toms and Shutdown Coolina Suction Valves In addition to the responses to your questions, GPC is offering this information on the ,
RHR torus and shutdown cooling valves 1/2E11-F004A-D and 1/2El1-F006A-D, which were removed from the scope of GL 89-10 subsequent to Plant Hatch's February 3,1994 submittal. Thejustification for removal of these valves is detailed below:
Formerly, the RHR torus suction valves, El1-F004, and the RHR shutdown cooling -
valves, El1-F006, were defined as having active safety functions to open and close, HL-4794 El-7 L
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Generic Letter 89-10 .
respectively. They are being changed to have no active safety function. The fonner classification'was thought to be necessary to support placing alternate shutdown -
cooling in service, on a loss of the normal shutdown cooling common suction path.'
. . (This was so because it was originally assumed that the shutdown cooling loop would . .
be placed in suppression pool cooling.) Alternate shutdown cooling involves placing Core Spray in service (with suction from the toms), and raising level to the steam lines. Coolant is then being recirculated through the SRVs to the torus and back to I the vessel.- RHR is then placed in suppression pool cooling. To place the loop. l i
previously in shutdown cooling in the suppression pool cooling mode, the F004s-would have to be opened and the F006s would have to be closed.
The loss of shutdown cooling event is described in the Nuclear Safety Operational
< Analysis (NSOA) report. However, criterion no. 2 of our guidelines states that valve :
operation requirements are limited to changing from the normal operating position required to mitigate dedgn basis events. Upon closer review, therefore, criteria no. 2 -;
may be used to remove the El1-F004 and F006 valves from the GL 89-10 active - .
' J safety function list as described below.
In States C and D (as defined in the NSOA), the reactor is pressurized below -
135 psig (Ref. figure G.6-16, Unit 1 FSAR); the Core Spray and LPCI systems are required to be operable. Therefore, the standby RHR loop is aligned for LPCI with the El1-F004 valves already open and the El1-F006 valves already closed.
Assuming a loss of the common shutdown cooling path, and assuming the F004 or F006 valves in the loop originally in shutdown cooling are unable to change position, -
the standby RHR loop could be used for suppression pool cooling'since the F006 and l F004 valves are properly aligned. This could be accomplished by opening the El1-F024 and F028 valves, or by opening the El1-F015 valve, and using LPCI l instead of Core Spray to recirculate water through the SRVs. The NSOA also !
defmes States A and B (Ref. figure G6-16, Unit 1 FSAR). This is essentially the reactor in the cold condition and vented.- In this situation, all that is required per the NSOA is to maintain core cooling. This can be accomplished either by the availability.
of a low pressure ECCS system, or having the. cavity flooded.
m In evaluating single failure, the two RHR pumps in one loop are off different power supplies. The F004, F006, and the suppression pool injection valves (El1-F024 and El1-F028) are all powered from the same motor control center, R24-S011 for the "A" loop and R24-S012 for the "B" loop. However, if this power supply should fail, or if for some other reason one of the suppression pool cooling valves fails to open, then the LPCI injection valve, El1-F015, can be opened. Since water is being recirculated between the reactor vessel and the suppression pool, it is immaterial whether the return path is to the RPV or the suppression pool. At any rate, cooling is i
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Generic Letter 89-10 being provided by the RHR heat exchangers. Thus, no single failure can prevent ,
ultimate cooling of the suppression pool /RPV. ,
The applicable abnormal operating procedures have been revised to accomm~Iate the .
change.
t The F004 and F006 valves are listed as low priority MOVs in the " Application of Probabilistic Safety Assessment to GL 89-10 Implementation" topical report developed by GE for the BWR Owner's Group.
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