Letter Sequence Request |
---|
|
|
MONTHYEARHL-3014, Forwards Response for Table 3 Items from App a of IST Program Safety Evaluation.Response to Section 3.2.3.1 of SE Also Encl1992-11-17017 November 1992 Forwards Response for Table 3 Items from App a of IST Program Safety Evaluation.Response to Section 3.2.3.1 of SE Also Encl Project stage: Other ML20116M8991992-11-23023 November 1992 Forwards NRC Response to Section 3.2.3.1 of SE for Second Ten Year Insp Interval for IST Program Project stage: Other HL-4433, Forwards Revs to Relief Requests RR-V-17,RR-V-19,RR-V-20, RR-V-32,RR-V-40 & RR-V-41,providing Addl Info as Discussed in 930901 Telcon Re Second 10-yr Insp Interval IST Program1993-12-21021 December 1993 Forwards Revs to Relief Requests RR-V-17,RR-V-19,RR-V-20, RR-V-32,RR-V-40 & RR-V-41,providing Addl Info as Discussed in 930901 Telcon Re Second 10-yr Insp Interval IST Program Project stage: Request HL-4550, Informs That in Response to Listed Items from SE for 18 Responses Provided in Util ,Util Has Revised Relief Requests RR-P-6 & RR-P-7 of Second 10-yr Insp Interval IST Program.Revised Relief Requests Encl1994-04-0404 April 1994 Informs That in Response to Listed Items from SE for 18 Responses Provided in Util ,Util Has Revised Relief Requests RR-P-6 & RR-P-7 of Second 10-yr Insp Interval IST Program.Revised Relief Requests Encl Project stage: Request 1992-11-23
[Table View] |
Similar Documents at Hatch |
---|
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
~
. _ Georgia Power Company.
40 invemess Center Parkway
,.k.* Post Office Box 1295 Birmingham, Alabama 35201 Telephone 205 877-7279 c
J. T. Beckham, Jr. Georgia Power -
Vice President - Nuclear '
Hatch Project the southem electre system.'
April 4, 1994 Docket Nos. 50-321 HL-4550 50-366 TAC Nos. M59202, M83192 M59203, M83193 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant Second Ten Year Inspection Interval IST Prograra Safety Evaluation ' Response Gentlemen:
By letter dated April 5,1993, the Nuclear Regulatory Commission (NRC) staff transmitted a Safety Evaluation (SE) for the 18 responses provided in Georgia Power Company's (GPC) letter dated June 5,1992. The .SE contained three items which were granted on an interim basis for a period of one. year to allow GPC to investigate alternate testing methods or actions which could be taken to resolve the NRC's concerns. The three 4 ems for which interim relief was granted are described below.
(1) Frequency response range for the instruments used for vibration monitoring of the Standby Liquid Control Pumps (Relief Request RR-P-6, SER' paragraph 2.2.3, page 8)
(2) Axial vibration monitoring on the Standby Diesel Service . Water Pump (Relief Request RR-P-6, SER paragraph 2.2.3, pages 3 and 9)
-(3) The variance in flow measurement for the Residual Heat Removal pumps (Relief Request RR-P-7, SER paragraph 2.3.3, page 11)
In response to the items listed above, GPC has taken the appropriate actions to resolve the :
NRC staffs concerns. Relief Request RR-P-6 has been revised to include additional justification for using the ~ hang vibration monitoring equipment. The NRC questioned the frequency response range of this equipment as it is applicable to the Standby Liquid 9404120145 940404 PDR_ ADOCK 05000321 P PDR L
.I
Georgia Powerd U.S. Nuclear Regulatory Commission Page Two l
April 4, 1994 j
, Control Pumps for the detection of bearing oil whirl. A review of the subject pumps in l conjunction with conversations with the pump vendor has shown that vibration associated with pump degradation will occur at frequencies which are within the frequency response range of the equipment utilized.
Relief Request RR-P-6 has also been revised to indicate that axial vibration monitoring will be performed on all vertical line shaft pumps which includes the Standby Diesel Service Water Pump. A relief request is necessary as the location of such measurements is ditTerent from that required by the O&M Code; however, the measurements will provide data representative of axial vibration of the subject pump.
Relief Request RR-P-7 has been revised to indicate changes implemented for the RHR flow instrument calibration accuracy in order to satisfy Code requirements to the extent practical. Site calibration procedures are being revised to tighten the calibration accuracy and thus meet the Code limits on the total flow measurement variance. The Code would allow a flow measurement variance of approximately 460 GPM and the total loop accuracy of the instruments utilized is being changed to provide a maximum variance of 425 GPM.
Copies of the referenced Relief Requests (RR-P-6 and RR-P-7) are enclosed for review.
The only changes necessary for GPC to be in compliance with the proposed testing are revision to instrument calibration procedures which will be implemented by May 1,1994.
Should you have any questions in regard to this issue, please advise. )
Sincerely, eA J. T. Beckham, Jr.
DGA/cr 004550
Enclosures:
(See next page.)
1
Georgia Power A 1
i U.S. Nuclear Regulatory Commission Page Three April 4, 1994 l
Enclosures:
Relief Request RR-P-6 l Relief Request RR-R-7 i
cc: Georeia Power Company Mr. H. L. Sumner, General Manager - Nuclear Plant NORMS ES. Nuclear Regulatory Commission. Washington. D.C. ,
Mr. K. Jabbour, Licensing Project Manager - Hatch ,
U.S. Nuclear Regulatorv Commission. Region H Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector - Hatch l i l l
l l
l l
1 PUMP RELIEF REQUEST RR-P-6 ,
SYSTEM: :' Standby Liquid Control ,
Residual Heat Removal Residual Heat Removal Service Water Core Spray l High Pressure Coolant Injection l Reactor Core Isolation Cooling (optional) >
Plant Service Water l PUMP (S): 1(2)C41-C001A,B (Positive Displacement Pumps)-
1M) Ell-C001A-D (Vertical Line Shaft Pumps) 1111-C002A-D (Centrifugal Pumps) .
.2E11-C002A-D (Vertical Line Shaft Pumps) '
1E21-C001A,B (Centrifugal Pumps) l 2E21-C001A,B (Vertical Line Shaft Pumps) l 1(2)E41-C001. (Centrifugal Pumps) j 1(2)E51-C001 (optional) (Centrifugal Pumps)
- 1(2)P41-C001A-D (Vertical Line Shaft Pumps) l 2P41-C002 (Vertical Line Shaft Pump) 2 and 3 CLASS:
, TEST REQUIREMENT: Perform pump IST in accordance with ASME Section XI l Subsection IWP'
, BASIS FOR RELIEF: It has been recognized within the industry that the OM Code l requirements for pump IST are .more suitable than those of ASME XI IWP. l The testir; requirements of the OM Code 1990, Section ISTB ALTERNATE TESTING:
will be . .ilized for pump IST for those pumps required to be teste by ASME Section XI except as identified in the l continuation sheet.
l The RCIC system is included for information purposes only. The RCIC pump is not required to be tested in accordance with ASME XI.
i 5-7a Rev.'3 l
VIBRATIONAL POINTS l
In lieu of the requirements of ISTB 4.6.4, vibration measurements will be taken .
as outlined below. j
- a. On centrifugal pumps measurements will be taken in a plane approximately perpendicular to the rotating shaft in two orthogonal directions. These measurements shall be taken on each accessible pump bearing housing.
Measurements shall also be taken in the axial direction on each accessible pump thrust bearing housing. If no pump bearing housings are accessible due to pump design or physical interference, then the measurements will be taken at the accessible location that gives the best indication of lateral / axial pump vibration. This location is either on the pump casing, the motor bearing casing, or the motor casing.
- b. On vertical line shaft pumps measurements will be taken in three orthogonal directions, one of which is in the axial direction in the area of the upper pump bearing housing (pump to motor mounting flange). This is the closest accessible location to a pump bearing housing and should provide readings which are at least as representative of pump mechanical condition as those l
required by the OM Code which are to be taken on the top of the pump motor.
The OM Code required vibration measurements on the upper motor-bearing housing are impractical because of the following reasons.
- 1. Plant design did not include permanent scaffolding or ladders which provide access to the top of the motors for the subject pumps. (All but Standby Plant Service Water Pump)
- 2. Physical layout of the pumps and interference with adjacent components does not allow for the installation of temporary scaffolding or ladders which are adequately safe for routine usc. (All but Standby Plant Service Water Pump)
- 3. There is a relatively thin cover plate bolted to the top-center of each motor which prevents measurements in line with the motor bearing. Measurement on the edge of the motor housing would be influenced by eccentricity and may not be representative of actual axial vibration.
- 4. Special tools (extension rod) for placing the vibration transducers are not practical because placement would not be sufficiently accurate for l trending data.
l
- 5. The Standby Plant Service Water Pump is accessible, but the motor has a cooling fan mounted at the top which is attached to the rotating shaft. The fan is protected by a relatively thin cover plate which prevents access to the motor housing for vibration measurements. Removing the cover does not provide for transducer placement since the rotating fan would still be in the way.
Research within the industry revealed that vibration monitoring of vertical line shaft pumps has been of limited benefit for detecting mechanical degradation due to problems inherent with pump design. The OM Code imposes more stringent 5-7b Rev. 3 l
l
_ _ . __ ,__ ..- ~ __
hydraulic acceptance criteria on these pumps than for centrifugal or positive displacement pumps. This more stringent hydraulic acceptance criteria would place more emphasis on detection of degradation through hydraulic test data than through mechanical test data.
Therefore, application of the OM hydraulic testing criteria alo1g with radial and axial vibration monitoring in the area of the top pump bearing housing should provide adequate data for assessing the condition of the subject pumps and for monitoring degradation.
- c. On reciprocating pumps, a measurement will be taken on the bearing housing of the crankshaft, approximately perpendicular to both the crankshaft and the line of the plunger travel. (As required by the OM Code.)
VIBRATION ACCEPTANCE CRITERIA In lieu of the requirements of TABLE ISTB 5.2-2a, ranges for vibration acceptance criteria for smooth running pumps will be as outlined below.
Small absolute changes in vibration for smooth running pumps (e.g. s.075 in./sec.) would potentially result in Alert and Required Action Ranges being declared for exceeding the 2.5Vr or 6Vr limits even though the pump is operating satisfactorily.
The Alert Range for smooth running pumps will be > 0.19 to 0.45 in./sec. and the Required Action Range starts at any value above 0.45 in./sec.
FREQUENCY RESPONSE RANGE OF VIBRATION INSTRUMENT The OM Code (ISTB 4.6.l(f)) requires a frequency response range of one-third pump operating speed to at least 1000 Hertz. The Standby Liquid Control (SBLC) Pumps operate at 370 RPM (6.2 HZ), therefore the instrument frequency response range of the Plant Hatch IST Program instrumentation does not satisfy the code requirement.
In lieu of the requirements of ISTB 4.6.l(f), the vibration measuring instrument frequency response range utilized for the Standby Liquid Control Pumps will be as described below.
- 1. An I.R.D. Model 810 with accuracy of d5% full scale over a frequency response range of 5.8 - 2,000 HZ for displacement measurements and 5.8 -
10,000 HZ for velocity measurements is utilized for IST.
- 2. The I.R.D. Model 810 lower frequency response limits result from high-pass ,
filters which eliminate low-frequency elements associated with the input signal from the integration process. These filters prevent low frequency electronic noise from distorting vibration readings thus any actual vibration occurring at low frequencies is filtered out.
- 3. The SBLC pumps are Union Pump Company reciprocating pumps. The subject pumps utilize roller bearings instead of sleeve bearings. Sleeve bearings can exhibit vibration at subsynchronous frequencies when a condition of oil whirl is present. However, oil whirl does not occur in roller or ball bearings.
1 5-7c Rev. 3 )
Roller and ball bearing degradation symptoms typically occur at IX shaft rotational frequency and greater. Therefore, vibration measurements at frequencies less than shaft speed would not provide meaningful data relative to degradation of the pump bearings.
- 4. The SBLC pumps are standby pumps only. They are only operated during Technical Specification Surveillance and Inservice Testing which results in very little run time. In the unlikely event that the system is required to perform its safety function, the pump run time would only be from 19 to 74 minutes to exhaust the volume of the sodium pentaborate storage tank.
- 5. In addition to the IST vibration monitoring program, these pumps are included in the site maintenance department vibration program. This program has the capability to perform spectral analysis with equipment which would satisfy the frequency response range requirement of the OM Code. The maintenance vibration monitoring is not performed at a frequency equivalent to that required for IST, but based on the infrequent operation of these pumps, the likelihood that a vibration problem would go undetected by both programs is minimal. The maintenance vibration program will also be utilized to analyze any IST vibration data which placed the pumps in the ALERT or ACTION Ranges. The need for any corrective actions would be based on evaluation of IST and maintenance testing program data.
- 6. Based on the pump bearing design, the combination of vibration monitoring implemented and the limited operation time, it seems unlikely that a vibration problem not detectable by the equipment being utilized would prevent these pumps from fulfilling their design safety function.
Use of the existing vibration monitoring equipment which is calibrated to at least 5% full scale over a frequency response range of 5.8 -2000 Hz (SBLC pump nominal shaft speed - 6.2 Hz) should provide sufficient data for monitoring the mechanical condition of the SBLC pumps. This equipment will provide accurate vibration measurements over the frequency range in which typical roller bearing vibration problems occur. This monitoring program should meet the intent of the code and will relieve the utility from the burden and expense involved with procurement, calibration, training and administrative control of new testing equipment which seems unjustified for assessing the mechanical condition of the subject pumps.
5-7d Rev. 3
4 3
PUMP RELIEF REQUEST l A RR-P-7 j i
SYSTEM: Residual Heat Removal, Core Spray, High Pressure Coolant Injection, l l Reactor Core Isolation Cocling and Plant Service Water Systems s
- PUMP (S): 1(2) Ell-C002A-D 1(2)E21-C001A,8 1(2)E41-C001 i
! 1(2)E51-C001 2P41-C001A-D l
- CLASS: 2-1 TEST REQUIREMENT: ASME OM Code, 1990, Section ISTB 4.6.1 and Table ISTB
, 4.6.1-1 define the required accuracy and full-scale range i for each instrument used to measure the test parameters.
- BASIS FOR RELIEF
- The original installed instrumentation associated with
! these pumps was not designed with the instrument accuracy l and ranges of OM Code ISTB Table 4.6.1-1 taken into ,
i consideration. The actual instrument ranges and loop accuracies are itemized on the attached sheets. These
, attached sheets provide information relative to the range, individual accuracy and total loop accuracy of those instruments that do not satisfy the OM requirements.
ALTERNATE TESTING: Test gages calibrated to 0.5 % accuracy will be utilized j for RHR and Core Spray pump inlet pressure measurement.
For. all other pump parameters the installed
- instrumentation will be utilized. The installed
- instrumentation should provide data that is sufficiently -
j accurate to allow assessment of pump condition and to l detect pump degradation.
See continuation sheets for individual evaluations and
- data relevant to accuracy of each instrument loop. '
The RCIC System is included for information purposes only. The RCIC pump is not j required to be tested in accordance with ASME XI.
1 I
i 4
- l 5-8a Rev. 2 ,
l l
l 1
)
INSTRUMENT RANGE TEST RANGE ALLOWABLE RANGE ACCURACY lEll-PI-R003A-D 0-600 psig = 182 psig 0-546 psig i 2 % (2) lEll-FI-R603A(B) 0-25000 gpm = 7700 gpm 0-23100 gpm i 1.66 % (2) l lE21-PI-R600A(B) 0-500 psig = 290 psig 0-870 psig i 2.06 % l lE41-SI-R610 0-6000 rpm = 3810 rpm 0-11430 rpm i 2 % (1) lE41-PI-R004 15"HG-100 psig = 27 psig 0-81 psig i 1 % (2) lE41-FI-R612 0-5000 gpm = 4250 gpm 0-12750 gpm 2.12 %
IE51-SI-R610 0-6000 rpm = 4250 rpm 0-12750 rpm 2 % (1) lE51-PI-R002 15"HG-100 psig = 24 psig 0-72 psig 1 % (2) lE51-FI-R612 0-600 gpm = 400 gpm 0-1200 gpm 2.12 %
2 Ell-PI-R003A-D 0-600 psig = 186 psig 0-558 psig 2 % (2) 2 Ell-FI-R603A(B) 0-25000 gpm = 7850 gpm 0-23550 gpm 1.22 % (2) l 2E21-PI-R600A(B) 0-500 psig = 308 psig 0-924 psig i 2.06 % l 2E41-SI-R610 0-6000 rpm = 3800 rpm 0-11400 rpm i 2 % (1) 2E41-PI-R004 15"HG-100 psig = 30 psig 0-90 psig 1 % (2) 2E41-FI-R612 0-5000 gpm = 4250 gpm 0-12750 gpm = 2.12 %
2E51-SI-R610 0-6000 rpm = 4250 rpm 0-12750 rpm 2 % (1) 2E51-PI-R002 15"HG-100 psig = 30 psig 0-90 psig 1 % (2) 2E51-FI-R612 0-600 gpm = 400 gpm 0-1200 gpm i 2.12 %
i NOTES:
- 1. An electronic speed element provides a signal to the speed indicator which is calibrated to
- 2 % of full scale. Therefore speed indication should satisfy the requirements of TABLE ISTB 4.6.1-1.
- 2. Exceeds code allowable range limit of three times reference value.
5-8b Rev. 2
l COMPONENT / COMPONENT / COMPONENT / TOTAL LOOP ACCURACY ACCURACY ACCURACY ACCURACY PER OM ISTB 1.3 lEll-FT-N015A,B IE11-K600A,8 IE11-FI-R603A,8 1.66 % I i 0.5 % 0.5 % 1.5 % !
IE21-PT-N001A,B IE21-PI-R600A,8 2.06 % ! )
, N/A 2% 2% N/A :
1E41-FT-N008 lE41-K691 lE41-FI-R612 2.12 %.
0.5 % 0.5 % 2.0 %
1E41-SI-R610 N/A N/A 2 % (1)
, IE51-FT-N003 1E51-K601 lE51-FI-R612 2.12 % '
O.5 % 0.5 % 2.0 %
1E51-SI-R610 N/A N/A 2 % (1) i 2E11-FT-N015A,8 2E11-K600A,B 2E11-FI-R603A,8 1.22 % I 0.5 % 0.5 % 1%
2E21-PT-N001A,B 2E21-PI-R600A,B N/A 2.06 % !
0.5 % 2% N/A: :
2E41-FT-N008 2E41-K601 2E41-FI-R612 2.12 %
0.5 % 0.5 % 2.0 % !
2E41-SI-R610 -N/A N/A. 2 % (1)
I 2E51-FT-N003 2E51-K601 2E51-FI-R612 2.12 % i 0.5 % 0.5 % 2.0 %-
2E51-SI-R610 N/A N/A 2 % (1) ;
Ii See page 5-8d for notes. ,
i 5-8c Rev. 2 I
1 1(3)E11-PI-R003A-D exceed the range limit of three times the reference value, however the additional gage range results in approximately 1 psig maximum allowable variance in the measured parameter. (i.e. 02 x 546 = 11 versus .02 x l 600 = 12) 1(2)E11-FI-R603A(B) exceed the Code allowable full scale range limit of three times the reference value. The indicator range includes consideration for LPCI flow rate (17,000 gpm for two pumps), whereas the IST pump flow rate is 7,700 gpm for Unit 1 and 7,850 for Unit 2. The code maximum allowable variance in measured flow rate would be 462 gpm (i.e. 02 x 23,100) for Unit 1 and 471 gpm (i.e. 02 x 23,550) for Unit 2. Whereas the actual maximum variance in measured flow is 425 gpm (i.e. 017 x 25,000) for Unit 1 and 325 gpm (i.e. 013 x 25,000) for Unit 2.
Therefore, the actual accuracy o, the installed flow indicators is greater than allowed by the Code, thus the range of the indicator exceeding the Code limit of three times the reference value is of no consequence.
1(2)E21-PI-R600A(B) exceed the maximum code allowable total loop accuracy however I the indicator used has a full scale range less than that allowed. The maximum code allowable variance in measurement is 17 psig (.02 x 870) for unit I and 18 psig for unit 2 (.02 x 924). However, by using a gage with a range less than allowed, the actual maximum allowable variance is 11 psig (.021 x 500). Therefore the actual accuracy is within the code allowable for the maximum allowable range.
1(2)E41-PI-R004 exceed the range limit of three times the reference value.
However, the gages are calibrated to 1 % full scale accuracy which results in the final variance being within the maximum allowable by the code (i.e.1.6 psig versus 1 psig for unit 1 and 1.8 psig versus 1 psig for unit 2).
1(2)E41-FI-R612 exceed the maximum code allowable total loop accuracy however the indicator used has a full scale range less than that allowed. The maximum variance allowable by the code is 255 gpm (.02 x 12750) whereas the actual maximum variance is 106 gpm (.0212 x 5000). Therefore the actual accuracy of the l instrument loop is better than that allowable by the code.
1(2)E51-PI-R002 exceed the range limit of three times the reference value. l However, the gages are calibrated to i 1 % full scale accuracy which results in the final variance being within the maximum allowable by the code (i.e. 1.4 psig l versus 1 psig for unit I and 1.8 psig versus 1 psig for unit 2).
1(2)E51-FI-R612 exceed the maximum code allowable total loop accuracy however the I indicator used has a full scale range less than that allowed. The maximum variance allowable by the code is 24 gpm (.02 x 1200) whereas the actual maximum variance is 13 gpm (.0212 x 600). Therefore the actual accuracy of the instrument loop is better than that allowable by the code. l 5-8d Rev. 2
- - -