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MONTHYEARML20198S9631997-11-0303 November 1997 Submits Info Pertaining to Unit 1 Implementation of Mods Associated w/GL-96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions Project stage: Other ML20216D6961998-04-14014 April 1998 Forwards Request for Addl Info Re GL 96-06,dtd 960930, Requesting Licensees to Evaluate Cooling Water Sys That Serve Containment Air Coolers to Assure That Coolers Not Vulnerable to Waterhammer & two-phase Flow Conditions Project stage: RAI HL-5645, Provides Response to 980414 RAI Re post-accident Waterhammer & two-phased Flow in Containment Coolers,Per GL 96-06.Util Plans to Complete Detailed Analysis & Any Resulting Procedure Revs by 981130.W/two Oversize Drawings1998-06-30030 June 1998 Provides Response to 980414 RAI Re post-accident Waterhammer & two-phased Flow in Containment Coolers,Per GL 96-06.Util Plans to Complete Detailed Analysis & Any Resulting Procedure Revs by 981130.W/two Oversize Drawings Project stage: Other HL-5652, Forwards,For Clarity,Original Question 2 & Complete Revised Answer to 980630 Response to 980414 NRC Ltr Re GL 96-06, Waterhammer in Containment Coolers1998-07-0808 July 1998 Forwards,For Clarity,Original Question 2 & Complete Revised Answer to 980630 Response to 980414 NRC Ltr Re GL 96-06, Waterhammer in Containment Coolers Project stage: Other 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 Project stage: Response to RAI ML20199H8941999-01-21021 January 1999 Discusses Responses to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Controls, for Plant,Units 1 & 2 Project stage: Other 1998-04-14
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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
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.' gg
!' , Vee President Operating Company,Inc.
Hatch Project Support 40 invemess Parkway Post Office Box 1295
[- Birmingham, Alabama 35201 Tel 205.992.7279 Fax 205.992.0341 so m e=N m .
COMPANY Energy ro Serve YourWorld" November 20, 1998
- Docket Nos. 321- Hi 5708 50-366 ,
i Tac Nos. 'M%819 ~
M%820 l 4
U.S. Nuclear Regulatory Commission '
ATTN: Document Control Desk
~ Washington, D.C. 20555 l ll Edwin I. Hatch Nuclear Plant Updated Response to Request for Additional Information .
Generic Letter 96 Waterhammer in Containment Coolers i
Ladies and Gentlemen:
On June 30,1998, Southem Nuclear Operating Company (SNC) responded to a Request for Additional Infonnation (RAI) concerning post-accident waterhammer and two-phased flow in .,
- containment coolers per Generic Letter (GL) 96-06. In the response for Plant Hatch, SNC
{
committed to perform detailed analyses, participate in an EPRI program for identifying more realistic waterhammer assumptions, revise procedures, and pro 5ide an update of the SNC response by November 30,1998. His letter provides that update.
Hatch Unit 1:
The Unit I containment area cooling system is an open loop system, in which the possibility of i waterhammer cannot be easily prevented by procedure changes or by simple design changes. I Detailed analysis supported by the EPRI demonstration program is necessary to close the issue.
Hus, completion of ac' ions for Unit I will be deferred pending completion and review of the EPRI !
technical basis report, which is currently scheduled for July 31,1999. Additional information will i be submitted following receipt of the EPRI report.
Hatch Unit 2:
' The Unit 2 containment area cooling system is a closed loop system, which must be manually l l' restarted following a LOCA. Procedure changes have been made which will prevent operation of l the system when conditions exist which could cause a waterhammer. The changes consider the y appropriate failure modes and effects. Bus, all actions for Unit 2 are complete.
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9811300207 981120,;
DR ADOCK O 1
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I 4 U.S. Nuclear Regulatory Commission Page 2 November 20, 1998 The enclosure to this letter provides a more detailed discussion. Should you have any questions in !
this regard, please contact this office.
l l
Respectfully submitted, y
M M H. L. Sumner, Jr.
JAW /eb
Enclosure:
Updated Response to Request for Additional Infornation.
Generic Letter 96 Waterhammer in Containment Coolers cc: S.quth em Nuclear Operatina Company Mr. P. H. Wells, Nuclear Plant General Manager SNC Document Management (R-Type A02.001)
U.S. Nuclear Reaulatory Commission. Washington. D.C.
Mr. L. N. Olshan, Project Manager - Hatch U.S. Nuclear Regulatory Commission. Region II Mr. L. A. Reyes, Regional Administrator Mr. J. T. Munday, Senior Resident Inspector - Hatch t
1 a
f
. 4 HL-5708
Enclosure Edwin I. Hatch Nuclear Plant Updated Response to Request for Additional Information:
Generic Letter % Waterhammer in Containment Coolers Summary Information Containment area cooling for Hatch Units 1 and 2 is not credited to operate after a design basis accident. However, the piping system pressure boundary must remain intact to ensure containment
- l. integrity.
, Hatch Unit 1:
l' Unit I containment area cooling water is provided by an open loop system, with water provided to fan cooling units via the safety related plant service water system. Due to the system design, waterhammer is possible in the fan cooling units / piping system whether or not the fan cooling units are operated after a LOCA with a Loss of Offsite Power (LOSP). Thus, SNC elected to perform a .
l detailed analysis to determine the potential for waterhammer and its affects on the piping system l l and support structure. SNC also elected to participate in the EPRI demonstration program, which l l should provide more realistic assumptions to use in the analysis. l
- ne detailed analysis consists of a dynamic hydraulic model and a structural model. He i preliminary detailed analysis has been performed using conservative assumptions that maximize l waterhammer loads (pressures and forces) on the piping system and piping support structure. He l assumptions include the use of maximum sonic velocities, no credit for noncondensible gases, and instantaneous void collapse. Using these conservative assumptions, the preliminary ant. lysis has shown that significant forces can be generated and transmitted to the piping support av
- 1 structural system. Bus, SNC has determined that input from the EPRI program will proside useful information necessary to complete the analysis, provide more realistic waterhammer loads, and to provide the necessary infonnation to resolve this issue. Further response to this issue is thus i deferred pending completion and review of the EPRI technical basis report.
Hatch Unit 2 Unit 2 containment area cooling water is provided by a closed loop system, with water provided to fan cooling units by the drywell chilled water system. His system (which includes the chillers, i' pumps, and fan cooling units) receives a LOCA signal, which will automatically shutdown the system. Prior to restarting the system, the operator must manually override the signal using a -
LOCA override procedure. Previously, this procedure was revised to prohibit operating specific fan cooling units in which the water may be susceptible to boiling.
As requested by the RAI, SNC considered the variaus failure modes of the system and that
- _ waterhammer could occur even after the containment has begun to cool. Dus, the procedure has
, been revised again to prohibit operating the entire system when boiling may have occurred in any of the fan cooling units / piping system in cocjunction with a LOCA.
HL-5708 E-1 l
~
Enclosure
] Updated Response to Reques? for Additional Information:
Generic Letter 96 Waterham.ner in Containment Coolers t
Detailed Discussion Hatch Unit 1:
- Water is provided to the Unit I containment area cooling system by the safety related plant service water system (PSW). The system is an open loop design. In the event of a LOCA and coincident LOSP, the normal system response would cause the PSW system pumps to trip, then restan after i
power is available. He fan cooling unit supply valves (air operated with electrical pilot solenoids) l would open, then automatically close after power is available. Containment isolation valves (motor i operated) would remain open in all cases. The cooling unit fans would trip due to the LOCA
- signal, and would not automatically restan. A failure mode exists that could cause the control l
4 valves to fail open upon loss of air or power. i
! )
To evaluate waterhammer for Unit 1, SNC decided to perform detailed analysis, including various
) failure modes and effects. Note that this analysis will not be finalized until the results of the EPRI
{ program are available.
l i The first step was to determine the extent of voiding within the system. To do this, two " drain l j down" calculations have been developed. The first assumes that the fan cooling unit inlet valves
- will close as designed after restoration of power. The second assumes that these valves fail open,
{' thus maximizing the drain down. A short time was assumed (about 3 seconds) for the temperature
- of the water in the fan cooling units to reach the boiling point. The system drain down was then I
determined in a quasi-steady state manner with the pressure in the fan cooling unit tubes assumed ;
i to be the saturation pressure corresponding to the drywell temperature given a LOCA. The total l l drain down volume is then the integrated value over the drain down period. His method assumes l perfect heat transfer from the drywell through the fan cooling unit tubes, and therefore provides a 1 conservative estimate of the volume of system voiding, independent of whether or not the fan is l j mnning, is coasting down, or has stopped during the drain down.
i Next, a hydraulic model was developed scing the Bechtel computer program HSTA (Hydraulic Systems Transient Analysis). To properly onsider failure modes, fan cooling unit inlet " valves open" and " valves closed" cases have been generated. Also, a second " valves closed" case is being
- generated to determine the effects of subsequent operator action to restart the fans as currently directed by the Emergency Operating Procedures (EOPs). The HSTA model generated preliminary
- forcing functions over each pipe mn segment along the piping system, which is the hydraulic force j on the pipe segment vs. time for the duration of the waterhammer event.
a l Third, the preliminary forcing functions were input into a piping / structural model using the Bechtel computer code ME-101. Using ME-101, several supports have been shown to need further evaluation. From this observation, several approaches are available for resolution:
1
, 1) perform more detailed modeling using a finite element code such as ANSYS to show that the y piping /suppon structure will yield without rupture of the piping system,
- 2) modify the piping /suppon structure to withstand the higher loads (mitigation),
i i
HL-5708 E-2
Enclosure Updated Response to Pequest for Additional Information:
Generic Letter % Waterhammer in Containment Coolers
[ 3) modify the system logic and valves to prevent water hammer, or l
L 4) reduce the piping / structure forcing function loads by reduction of the predicted waterhammer l
loads in the hydraulic model.
Reducing the predicted waterhammer loads (which would reduce the forcing function loads) is one of the primary expectations of the EPRI program. Thus, SNC has elected ira participate in the L EPRI program. He calculations, hydraulic model, and piping / structural model will be finalized and will be available for review upon completion of the EPRI program.
Hatch Unit 2:
Water is provided to the Unit 2 containment area cooling system by the di vwell chilled water system. He system is a closed loop design, with a surge tank maintaining some positive system pressure when the system is shut down. If a LOCA were to occur, the chilkd water system would trip and would not automatically restart If a coincident LOSP were to also occur, the cooler supply valves (air operated with electrical pilot solenoids) would open, then t utomatically close aAer power is available. Contamment isolation valves (motor operated) wouli remain open in all cases. De cooling unit fans would trip following the LOCA signal, and wout i not automatically l restart. A failure mode exists which could cause the control valves to fail open upon loss of air or power.
It was determined that the water in the fan cooling units could boil following a 1,0CA. His was 1 checked by performing a review of the elevation of the surge tank relative to the fan cooling unit !
elevations determine boiling temperatures compared to elevated containment temperatures and associated steam in the containment.' Even if the containment temperature were then reduced to below the water boiling points, voids could still exist within the piping system du: to a time delay l for the piping to cool aAer the containment cools. !
. l l The operator can restart the system aRer overriding the LOCA signal using a LOCA override ;
procedure. This procedure had previously been revised to prevent operation of the upper level fan l l ,
cooling units following a LOCA, and to prevent operation of the system if the contHnment !
temperature rose above the boiling point of all of the fan cooling units. However, this revision did i
- not consMer the effects of potential failure of the control valves, nor did it consider the possibility of continued voids aRet the containment has cooled.
l Thus, as stated in the Summary section, the LOCA override procedure has been revised again.
The procedure now prohibits the operation of the entire cooling system (including chillers, pumps, i
and fan cooling units) if a LOCA exists and the containment temperature hu at any time been above the boiling point of the water in any of the fan cooling units.
i t
HL-5708 E-3
r Enclosure l
Updated Response to Request for Additional Information- I Generic 12tter 96 Waterhammer in Containment Coolers j I
l Computer Model Descriptions '
l HSTA l De computer program HSTA (Hydraulic Systems Iransient Analysis) was used for the water hammer
! analysis ofthe Containment Area Cooling System at Hatch Unit 1. His program is a generalized fmite l difference code developed by Bechtel Corporation and used extensively over the last 20 years for water hammer design and diagnostic purposes in nuclear and non-nuclear piping systems. He metixxl of l
charactensucs is used to solve the hyperbolic parual differential equations (ofcontinuity and momentum) to obtain the liquid velocity and pressure head at a known grid location. Dese flow mriables are then utilized to generate the dynamic forcing functions on specified pipe run segments. HSTA can model a complex piping system contauung one or more of the several different types offlow devices (boundary conditions) present in the system. Examples of these boundary conditions are: time dependent pressure / flow reservoirs, valves, branches, pumps, surge and air tanks, vacuum breakers, etc.
For the validation of the HSTA code, emphasis was placed on comparison with experimental or test data. His was supplemented by comparisons against independent numerically predicted results available. When comparing time-history predictions of pressure, velocity, etc. in a piping system, both the magnitude and frequency of the variable compared could be important depending on the piping response. For this reason the comparisons are given by directly superimposing the HSTA predicted on the measured (or calculated) variable time-histories and not just by defining the percentage agreement or disagreement between them.
De formal HSTA program documentation includes the validation of the following HSTA capabilities:
- 1. Validauon of valve actuation transient and pressure wave reflection / transmission at branches / area changes in complex piping networks is against data given in the standard text books by Wylie and Streeter, and Parmakian
- 2. Validation of centrifugal and reciprocating pump actuation, surge vessels and air tank mitigation devices is agamst various test data (available in open literature and in other Bechtel proprietary
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data). j
- 3. Validation ofwater column separation and rejoining calculation schemes includes the validation of ;
both conventional and line filing schemes described above.
De conwntional scheme is validated agamst test data from several careful laboratory experiments in Europe for two different piping geometries. Further validation is against laboratory tests done for water hammer predicbons in a piping system at a nuclear power plant in U.S.
For the line filling calculation scheme, the HSTA validation was performed as follows:
a) Compansons agamst laboratory test data from Europe.
l b) Comparisons against in-plant test data from a nuclear power plant in U.S.
l HL-5708 E-4
Enclosure Updated Response to Request for Additional Information:
l Generic Letter 96 Waterhammer in Containment Coolers l
c) Co...y.risc=s against predictions from a totally differat computer program that used a simpler calculation scheme and not the Method ofCharacteristics that is used by HSTA.
d) Cu...porisees against predictions from the conventional calculation scheme in HSTA for vapor l pockets smaller than a nodal distance in lerngth Besides the vahdation in the formal code documentation discussed above, the HSTA code participatal in the EPRI water hammer computer code evaluation program during the 1987-1992 EPRI sponsored research effort into water hammer In this program, the results from HSTA were compared against !
those from several other participating codes for a set ofsix varial water hammer simulation problems. l In the problems for wiuch data (test and analytical) was available, HSTA results compared very well agamst such data.
,. i r
I The HSTA program methodology and its application has been published widely (six papers) both in i national and intenational conferences
- 1 ME-101 The ME-101 program is used to determine stresses and loads in the piping systems due to i restrained thermal expansion, deadweight, seismic inertia and anchor movements, externally }
applied loads such asjet-loads, and transient forcing functions such as created by fast relief valve opening and closing, fast check valve closure after pipe breaks in the main feedwater line, fast valve closure in main steam line, etc. ME-101 analyzes piping systems in accordance with ANSI and ASME codes.
The ME-101 program is a finite element computer program which performs linear elastic analysis of piping systems using the stiffness method of finite element analysis; the displacements of the joints of a given structure are considered basic unknowns. The dynamic analysis by the modal synthesis method utilizes known maximum accelerations produced in a single degree of freedom model of a certain frequency. The principal program assumptions are as follows:
- 1. It is a linearly clastic structure.
- 2. Simultaneous displacement of all supports is described by a single time-dependent function.
- 3. Lumped mass model satisfactorily replaces the continuous structure.
. 4. Modal synthesis is applicable;
- 5. Rotational inertia of the masses has negligible effect.
'Ihe results obtained from pipe stress program ME-101 have been compared with the following:
- 1. ME-632, computer program, seismic analysis of piping systems, VERB MOD 8, Bechtel
- International Corporation, San Francisco, California,1976.
l l-HL-5708 E-5 i
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Enclosure Updated Response to Request for Additional Information:
Generic Letter 96 Waterhammer in Containment Coolers j
- 2. ASME Benchmark problem results, Pressure Vessel and Piping 1972 computer programs verification, American Society of Mechanical Engineers.
1
- 3. Longhand calculations-ME-101 is compatible with NRC Regulatory Guide 1.92. A synthesis l of closely spaced modes is provided based on equation 4 of Regulatory Guide 1.92.
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The verification report is on file at Bechtel. i l
Other Procrams If other programs are used in the final analysis, they will be described in the fmal response after the EPRI program is complete.
SNC COMMITMENTS AND SCHEDULE:
- 1. SNC will continue to participate in the NEI/EPRI program to develop a technical basis l document to demonstrate proper input assumptions are used in the analysis. The program is currently scheduled for completion by July 1,1999. ,
l
- 2. SNC will finalize the detailed analysis of the Unit I plant service water system upon completion of the EPRI program, and will update the NRC staff by letter, within 90 days of l completion of the EPRI program. j l
i HI 5708 E-6