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Category:CORRESPONDENCE-LETTERS
MONTHYEARNPL-99-0564, Forwards Response to NRC Request During 990720 Meeting,To Provide Addl Details to Several Questions Re Amend Currently Under Review by Staff Pertaining to CR Habitability1999-10-19019 October 1999 Forwards Response to NRC Request During 990720 Meeting,To Provide Addl Details to Several Questions Re Amend Currently Under Review by Staff Pertaining to CR Habitability ML20217A5911999-09-30030 September 1999 Advises of NRC Plans for Future Insp Activities at Facility for Licensee to Have Opportunity to Prepare for Insps & to Provide NRC with Feedback on Any Planned Insps Which May Conflict with Plant Activities.Plant Issues Matrix Encl 05000266/LER-1999-007, Forwards LER 99-007-00 for Point Beach Nuclear Plant,Unit 1. Condition Would Be Outside App R Design Basis for Plant.New Commitments within Rept Indicated in Italics1999-09-30030 September 1999 Forwards LER 99-007-00 for Point Beach Nuclear Plant,Unit 1. Condition Would Be Outside App R Design Basis for Plant.New Commitments within Rept Indicated in Italics ML20212J7431999-09-30030 September 1999 Forwards Insp Repts 50-266/99-15 & 50-301/99-15 on 990830- 0903.No Violations Noted.Inspectors Concluded That Util Licensed Operator Requalification Training Program Satisfactorily Implemented NPL-99-0555, Discusses Rev 1,suppl 1 to GL 92-01, Reactor Vessel Structural Integrity. Calculation That Provides Evaluation of New Surveillance Data for Assessing Integrity of Unit 1 Reactor Vessel1999-09-29029 September 1999 Discusses Rev 1,suppl 1 to GL 92-01, Reactor Vessel Structural Integrity. Calculation That Provides Evaluation of New Surveillance Data for Assessing Integrity of Unit 1 Reactor Vessel ML20212K7651999-09-29029 September 1999 Forwards Insp Repts 50-266/99-13 & 50-301/99-13 on 990714-0830.No Violations Noted.Operators Responded Well to Problems with Unit 1 Instrument Air Leak & Unit 2 Turbine Governor Valve Position Fluctuation ML20212D5771999-09-15015 September 1999 Discusses Review of Response to GL 88-20,suppl 4,requesting All Licensees to Perform Ipeee.Ser,Ter & Supplemental TER Encl ML20211Q6451999-09-0808 September 1999 Forwards Operator Licensing Exam Repts 50-266/99-301OL & 50-301/99-301OL for Exams Conducted on 990726-0802 at Point Beach Npp.All Nine Applicants Passed All Sections of Exam ML20211Q4171999-09-0606 September 1999 Responds to VA Kaminskas by Informing That NRC Tentatively Scheduled Initial Licensing Exam for Operator License Applicants During Weeks of 001016 & 23.Validation of Exam Will Occur at Station During Wk of 000925 05000266/LER-1999-004, Forwards LER 99-004-01,re Fuel Oil Transfer Pump Cable in AFW Pump Room Being Outside App R Design Basis.Suppl to LER Provides Corrective Actions to Address Concerns Re Fire Disrupting Electrical Power to Fuel Oil Transfer Pump1999-09-0202 September 1999 Forwards LER 99-004-01,re Fuel Oil Transfer Pump Cable in AFW Pump Room Being Outside App R Design Basis.Suppl to LER Provides Corrective Actions to Address Concerns Re Fire Disrupting Electrical Power to Fuel Oil Transfer Pump ML20211K5261999-08-31031 August 1999 Forwards Insp Repts 50-266/99-14 & 50-301/99-14 on 990726- 30.Areas Examined within Secutity Program Identified in Rept.No Violations Noted ML20211F6941999-08-27027 August 1999 Provides Individual Exam Results for Applicants That Took Initial License Exam in July & August of 1999.Completed ES-501-2,copy of Each Individual License,Ol Exam Rept, ES-303-1,ES-303-2 & ES-401-8 Encl.Without Encl NPL-99-0473, Informs of Change Being Made to Plan Third 10-year Interval ISI Long Term Plan.Change Extends Interval from Current End Date of 001130 to 020831,due to Operating Cycle Being Increased from 12 to 18 Months1999-08-27027 August 1999 Informs of Change Being Made to Plan Third 10-year Interval ISI Long Term Plan.Change Extends Interval from Current End Date of 001130 to 020831,due to Operating Cycle Being Increased from 12 to 18 Months ML20211E8791999-08-24024 August 1999 Discusses Completion of Licensing Action for GL 96-01, Testing of Safety-Related Logic Circuits, for Point Beach Nuclear Power Plant,Units 1 & 2.Licensees Provided Requested Info & Responses Required by GL 96-01 ML20211F1501999-08-24024 August 1999 Submits Summary of Meeting Held on 990729,in Region III Office with Util Re Proposed Revs to Plant Emergency Action Level Criteria Used in Classifying Emergencies & Results of Recent Improvement Initiatives in Emergency Preparedness 05000266/LER-1999-006, Forwards LER 99-006-00 Which Describes Discovery That Postulated Fire in Central Zone of Primary Auxiliary Bldg Could Result in Spurious Operation of Pressurizer Porv. New Commitments within Rept Are Indicated in Italics1999-08-19019 August 1999 Forwards LER 99-006-00 Which Describes Discovery That Postulated Fire in Central Zone of Primary Auxiliary Bldg Could Result in Spurious Operation of Pressurizer Porv. New Commitments within Rept Are Indicated in Italics NPL-99-0477, Forwards Revised Procedures to Point Beach Nuclear Plant Epips.Revised Procedures Dtd 990723,should Be Filed in NRC Copies of Manual IAW Attached Instructions1999-08-18018 August 1999 Forwards Revised Procedures to Point Beach Nuclear Plant Epips.Revised Procedures Dtd 990723,should Be Filed in NRC Copies of Manual IAW Attached Instructions NPL-99-0426, Requests Relief from Section II of ASME B&PV Code, Nuclear Vessels, 1965 Edition,No Addenda.Detailed Info Attached1999-08-16016 August 1999 Requests Relief from Section II of ASME B&PV Code, Nuclear Vessels, 1965 Edition,No Addenda.Detailed Info Attached ML20210L9141999-08-0404 August 1999 Informs That Versions of Info Re WCAP-14787,submitted in 990622 Application for Amend,Marked Proprietary,Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954,as Amended ML20210K5221999-08-0404 August 1999 Discusses Point Beach Nuclear Plant,Units 1 & 2 Response to Request for Info in GL 92-01,Rev 1,Suppl 1, Rv Structural Integrity NPL-99-0436, Forwards fitness-for-duty Performance Data for six-month Period Ending 9906301999-08-0202 August 1999 Forwards fitness-for-duty Performance Data for six-month Period Ending 990630 ML20210G6011999-07-30030 July 1999 Discusses 990415 Complaint OSHA Received from Employee of Wisconsin Electric Power Co Alleging That Employee Received Lower Performance Appraisal for 1998 Because Employee Raised Safety Concerns While Performing Duties at Point Beach NPL-99-0406, Provides Response to NRC GL 99-02, Lab Testing of Nuclear- Grade Activated Charcoal1999-07-29029 July 1999 Provides Response to NRC GL 99-02, Lab Testing of Nuclear- Grade Activated Charcoal ML20210H0211999-07-28028 July 1999 Forwards Insp Repts 50-266/99-09 & 50-301/99-09 on 990528-0713.Two Violations of NRC Requirements Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy ML20210G2441999-07-26026 July 1999 Discusses 990714 Meeting with PRA Staff to Discuss Initiatives in Risk Area & to Establish Dialog Between SRAs & PRA Staff NPL-99-0408, Forwards Pbnps,Units 1 & 2 Plant Simulation Four-Yr Rept, IAW 10CFR55.45(b)(5)(ii).Rept Describes Certification Program Tests Conducted from 1996-1999,identifies Test Discrepancies Still Outstanding & Schedules for 2000-20031999-07-15015 July 1999 Forwards Pbnps,Units 1 & 2 Plant Simulation Four-Yr Rept, IAW 10CFR55.45(b)(5)(ii).Rept Describes Certification Program Tests Conducted from 1996-1999,identifies Test Discrepancies Still Outstanding & Schedules for 2000-2003 ML20209H5471999-07-14014 July 1999 Forwards Insp Repts 50-266/99-12 & 50-301/99-12 on 990614-18.One Violation Noted,But Being Treated as non-cited violation.Long-term MOV Program Not Sufficiently Established to close-out NRC Review of Program,Per GL 89-10 NPL-99-0395, Forwards Partial Response to NRC 990512 RAI Re TS Change 204 Re Control Room Habitability.Meeting Is Planned with NRC to Discuss Issues Related to Control Room & Primary Auxiliary Bldg Ventilation Sys Modifications1999-07-12012 July 1999 Forwards Partial Response to NRC 990512 RAI Re TS Change 204 Re Control Room Habitability.Meeting Is Planned with NRC to Discuss Issues Related to Control Room & Primary Auxiliary Bldg Ventilation Sys Modifications NPL-99-0390, Projects Listed Major near-term License Amend Requests That Could Be Expected to Impact Staff Resources Into Fiscal Years 2000 & 2001,in Response to Administrative Ltr 99-021999-07-0808 July 1999 Projects Listed Major near-term License Amend Requests That Could Be Expected to Impact Staff Resources Into Fiscal Years 2000 & 2001,in Response to Administrative Ltr 99-02 NPL-99-0388, Forwards MORs for June 1999 & Revised MORs for May 1999 for Pbnps,Units 1 & 21999-07-0707 July 1999 Forwards MORs for June 1999 & Revised MORs for May 1999 for Pbnps,Units 1 & 2 NPL-99-0381, Submits Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. GL 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl1999-06-30030 June 1999 Submits Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. GL 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl ML20196J4161999-06-30030 June 1999 Discusses Relief Requests Submitted by Wisconsin Electric on 980930 for Pump & Valve Inservice Testing Program,Rev 5. Safety Evaluation Authorizing Relief Requests VRR-01,VRR-02, PRR-01 & ROJ-16 Encl NPL-99-0379, Documents Telcon with Hg Ashar of NRC Re Licensee Intentions & Basis for Reselection of Control Tendons in Pbnps Containment Structures.Plants Are Currently Completing 28th Year Tendon Surveillance During Summer of 19991999-06-29029 June 1999 Documents Telcon with Hg Ashar of NRC Re Licensee Intentions & Basis for Reselection of Control Tendons in Pbnps Containment Structures.Plants Are Currently Completing 28th Year Tendon Surveillance During Summer of 1999 NPL-99-0376, Forwards Errata to Pbnp 1998 Annual Monitoring Rept, Originally Submitted by Ltr Dtd 990427.List of Corrections, Provided1999-06-28028 June 1999 Forwards Errata to Pbnp 1998 Annual Monitoring Rept, Originally Submitted by Ltr Dtd 990427.List of Corrections, Provided NPL-99-0353, Forwards June 1999 Rev to FSAR for Point Beach Nuclear Plant,Units 1 & 2, IAW Requirements of 10CFR50.71(e).Each Package Contains Revised FSAR Pages That Are to Be Inserted IAW Instructions1999-06-23023 June 1999 Forwards June 1999 Rev to FSAR for Point Beach Nuclear Plant,Units 1 & 2, IAW Requirements of 10CFR50.71(e).Each Package Contains Revised FSAR Pages That Are to Be Inserted IAW Instructions ML20196D4931999-06-18018 June 1999 Forwards Insp Repts 50-266/99-08 & 50-301/99-08 on 990411- 0527.No Violations Noted.Operator Crew Response to Equipment Induced Challenges Generally Good.Handling of Steam Plume in Unit 1 Turbine Bldg Particularly Good ML20195J9471999-06-16016 June 1999 Discusses Ltr from NRC ,re Arrangements Made to Finalized Initial Licensed Operator Exam to Be Administered at Point Beach Nuclear Plant During Week of 990726 ML20196A2931999-06-16016 June 1999 Ack Receipt of Transmitting Changes to Listed Sections of Point Beach Nuclear Plant Security Plan & ISFSI Security Plan,Submitted IAW 10CFR50.54(p).No NRC Approval Is Required Since Changes Do Not Decrease Effectiveness ML20195J9251999-06-14014 June 1999 Discusses 990610 Telcon Between Wp Walker & D Mcneil Re Arrangements for NRC to Inspect Licensed Operator Requalification Program at Point Beach Nuclear Power Plant for Week of 990816 05000266/LER-1999-005, Forwards LER 99-005-00,re Failure of Shell of 4B FW Heater Which Resulted in Significant Steam Leak & Manual Trip. New Commitments within Rept Are Indicated in Italics1999-06-11011 June 1999 Forwards LER 99-005-00,re Failure of Shell of 4B FW Heater Which Resulted in Significant Steam Leak & Manual Trip. New Commitments within Rept Are Indicated in Italics NPL-99-0336, Forwards Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1, IAW ASME Section Xi,Subsection IWA-62301999-06-10010 June 1999 Forwards Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1, IAW ASME Section Xi,Subsection IWA-6230 NPL-99-0330, Forwards Revs to Pbnp Security Plan Sections 2.1,2.4,3.1, Figures A,D & T & Pbnp ISFSI Security Plan Section 2.0, Dtd 990604.Plans Withheld1999-06-0404 June 1999 Forwards Revs to Pbnp Security Plan Sections 2.1,2.4,3.1, Figures A,D & T & Pbnp ISFSI Security Plan Section 2.0, Dtd 990604.Plans Withheld 05000301/LER-1999-003, Forwards LER 99-003-00 for Point Beach Nuclear Plant,Unit 2. Rept Is Provided in Accordance with 10CFR50.73(a)(2)(i)(B), as Any Operation or Condition Prohibited by Plant Tech Specs1999-05-28028 May 1999 Forwards LER 99-003-00 for Point Beach Nuclear Plant,Unit 2. Rept Is Provided in Accordance with 10CFR50.73(a)(2)(i)(B), as Any Operation or Condition Prohibited by Plant Tech Specs NPL-99-0319, Provides Main Control Board Wiring Separation Project Status Update Rept for Pbnps,Units 1 & 21999-05-28028 May 1999 Provides Main Control Board Wiring Separation Project Status Update Rept for Pbnps,Units 1 & 2 ML20206T3691999-05-17017 May 1999 Ltr Contract,Task Order 242 Entitled, Review Point Beach 1 & 2 Conversion of Current TS for Electrical Power Systems to Improved TS Based on Standard TS, Under Contract NRC-03-95-026 ML20206N5561999-05-13013 May 1999 Informs That NRC Office of Nuclear Reactor Regulation Reorganized Effective 990328.As Part of Reorganization,Div of Licensing Project Mgt Created.Cm Craig Will Be Section Chief for Point Beach Npp.Organization Chart Encl ML20206P2551999-05-12012 May 1999 Forwards Handout Provided to NRC by Wisconsin Electric at 990504 Meeting Which Discussed Several Recent Operational Issues & Results of Recent Improvement Initiatives in Engineering ML20206N5331999-05-12012 May 1999 Forwards RAI Re & Suppl by Oral Presentation During 980604 Meeting,Requesting Amend for Plant,Units 1 & 2 to Revise TSs 15.3.12 & 15.4.11 ML20196F3211999-05-11011 May 1999 Requests Proprietary WCAP-14787, W Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wepc Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt-NSSS Power), Be Withheld from Public Disclosure ML20206K0391999-05-0707 May 1999 Forwards Insp Repts 50-266/99-06 & 50-301/99-06 on 990223- 0410.Ten Violations of NRC Requirements Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARNPL-99-0564, Forwards Response to NRC Request During 990720 Meeting,To Provide Addl Details to Several Questions Re Amend Currently Under Review by Staff Pertaining to CR Habitability1999-10-19019 October 1999 Forwards Response to NRC Request During 990720 Meeting,To Provide Addl Details to Several Questions Re Amend Currently Under Review by Staff Pertaining to CR Habitability 05000266/LER-1999-007, Forwards LER 99-007-00 for Point Beach Nuclear Plant,Unit 1. Condition Would Be Outside App R Design Basis for Plant.New Commitments within Rept Indicated in Italics1999-09-30030 September 1999 Forwards LER 99-007-00 for Point Beach Nuclear Plant,Unit 1. Condition Would Be Outside App R Design Basis for Plant.New Commitments within Rept Indicated in Italics NPL-99-0555, Discusses Rev 1,suppl 1 to GL 92-01, Reactor Vessel Structural Integrity. Calculation That Provides Evaluation of New Surveillance Data for Assessing Integrity of Unit 1 Reactor Vessel1999-09-29029 September 1999 Discusses Rev 1,suppl 1 to GL 92-01, Reactor Vessel Structural Integrity. Calculation That Provides Evaluation of New Surveillance Data for Assessing Integrity of Unit 1 Reactor Vessel 05000266/LER-1999-004, Forwards LER 99-004-01,re Fuel Oil Transfer Pump Cable in AFW Pump Room Being Outside App R Design Basis.Suppl to LER Provides Corrective Actions to Address Concerns Re Fire Disrupting Electrical Power to Fuel Oil Transfer Pump1999-09-0202 September 1999 Forwards LER 99-004-01,re Fuel Oil Transfer Pump Cable in AFW Pump Room Being Outside App R Design Basis.Suppl to LER Provides Corrective Actions to Address Concerns Re Fire Disrupting Electrical Power to Fuel Oil Transfer Pump NPL-99-0473, Informs of Change Being Made to Plan Third 10-year Interval ISI Long Term Plan.Change Extends Interval from Current End Date of 001130 to 020831,due to Operating Cycle Being Increased from 12 to 18 Months1999-08-27027 August 1999 Informs of Change Being Made to Plan Third 10-year Interval ISI Long Term Plan.Change Extends Interval from Current End Date of 001130 to 020831,due to Operating Cycle Being Increased from 12 to 18 Months 05000266/LER-1999-006, Forwards LER 99-006-00 Which Describes Discovery That Postulated Fire in Central Zone of Primary Auxiliary Bldg Could Result in Spurious Operation of Pressurizer Porv. New Commitments within Rept Are Indicated in Italics1999-08-19019 August 1999 Forwards LER 99-006-00 Which Describes Discovery That Postulated Fire in Central Zone of Primary Auxiliary Bldg Could Result in Spurious Operation of Pressurizer Porv. New Commitments within Rept Are Indicated in Italics NPL-99-0477, Forwards Revised Procedures to Point Beach Nuclear Plant Epips.Revised Procedures Dtd 990723,should Be Filed in NRC Copies of Manual IAW Attached Instructions1999-08-18018 August 1999 Forwards Revised Procedures to Point Beach Nuclear Plant Epips.Revised Procedures Dtd 990723,should Be Filed in NRC Copies of Manual IAW Attached Instructions NPL-99-0426, Requests Relief from Section II of ASME B&PV Code, Nuclear Vessels, 1965 Edition,No Addenda.Detailed Info Attached1999-08-16016 August 1999 Requests Relief from Section II of ASME B&PV Code, Nuclear Vessels, 1965 Edition,No Addenda.Detailed Info Attached NPL-99-0436, Forwards fitness-for-duty Performance Data for six-month Period Ending 9906301999-08-0202 August 1999 Forwards fitness-for-duty Performance Data for six-month Period Ending 990630 NPL-99-0406, Provides Response to NRC GL 99-02, Lab Testing of Nuclear- Grade Activated Charcoal1999-07-29029 July 1999 Provides Response to NRC GL 99-02, Lab Testing of Nuclear- Grade Activated Charcoal NPL-99-0408, Forwards Pbnps,Units 1 & 2 Plant Simulation Four-Yr Rept, IAW 10CFR55.45(b)(5)(ii).Rept Describes Certification Program Tests Conducted from 1996-1999,identifies Test Discrepancies Still Outstanding & Schedules for 2000-20031999-07-15015 July 1999 Forwards Pbnps,Units 1 & 2 Plant Simulation Four-Yr Rept, IAW 10CFR55.45(b)(5)(ii).Rept Describes Certification Program Tests Conducted from 1996-1999,identifies Test Discrepancies Still Outstanding & Schedules for 2000-2003 NPL-99-0395, Forwards Partial Response to NRC 990512 RAI Re TS Change 204 Re Control Room Habitability.Meeting Is Planned with NRC to Discuss Issues Related to Control Room & Primary Auxiliary Bldg Ventilation Sys Modifications1999-07-12012 July 1999 Forwards Partial Response to NRC 990512 RAI Re TS Change 204 Re Control Room Habitability.Meeting Is Planned with NRC to Discuss Issues Related to Control Room & Primary Auxiliary Bldg Ventilation Sys Modifications NPL-99-0390, Projects Listed Major near-term License Amend Requests That Could Be Expected to Impact Staff Resources Into Fiscal Years 2000 & 2001,in Response to Administrative Ltr 99-021999-07-0808 July 1999 Projects Listed Major near-term License Amend Requests That Could Be Expected to Impact Staff Resources Into Fiscal Years 2000 & 2001,in Response to Administrative Ltr 99-02 NPL-99-0388, Forwards MORs for June 1999 & Revised MORs for May 1999 for Pbnps,Units 1 & 21999-07-0707 July 1999 Forwards MORs for June 1999 & Revised MORs for May 1999 for Pbnps,Units 1 & 2 NPL-99-0381, Submits Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. GL 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl1999-06-30030 June 1999 Submits Response to NRC GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants. GL 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Disclosure Encl NPL-99-0379, Documents Telcon with Hg Ashar of NRC Re Licensee Intentions & Basis for Reselection of Control Tendons in Pbnps Containment Structures.Plants Are Currently Completing 28th Year Tendon Surveillance During Summer of 19991999-06-29029 June 1999 Documents Telcon with Hg Ashar of NRC Re Licensee Intentions & Basis for Reselection of Control Tendons in Pbnps Containment Structures.Plants Are Currently Completing 28th Year Tendon Surveillance During Summer of 1999 NPL-99-0376, Forwards Errata to Pbnp 1998 Annual Monitoring Rept, Originally Submitted by Ltr Dtd 990427.List of Corrections, Provided1999-06-28028 June 1999 Forwards Errata to Pbnp 1998 Annual Monitoring Rept, Originally Submitted by Ltr Dtd 990427.List of Corrections, Provided NPL-99-0353, Forwards June 1999 Rev to FSAR for Point Beach Nuclear Plant,Units 1 & 2, IAW Requirements of 10CFR50.71(e).Each Package Contains Revised FSAR Pages That Are to Be Inserted IAW Instructions1999-06-23023 June 1999 Forwards June 1999 Rev to FSAR for Point Beach Nuclear Plant,Units 1 & 2, IAW Requirements of 10CFR50.71(e).Each Package Contains Revised FSAR Pages That Are to Be Inserted IAW Instructions 05000266/LER-1999-005, Forwards LER 99-005-00,re Failure of Shell of 4B FW Heater Which Resulted in Significant Steam Leak & Manual Trip. New Commitments within Rept Are Indicated in Italics1999-06-11011 June 1999 Forwards LER 99-005-00,re Failure of Shell of 4B FW Heater Which Resulted in Significant Steam Leak & Manual Trip. New Commitments within Rept Are Indicated in Italics NPL-99-0336, Forwards Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1, IAW ASME Section Xi,Subsection IWA-62301999-06-10010 June 1999 Forwards Unit 2 Refueling 23 Inservice Insp Summary Rept for Form NIS-1, IAW ASME Section Xi,Subsection IWA-6230 NPL-99-0330, Forwards Revs to Pbnp Security Plan Sections 2.1,2.4,3.1, Figures A,D & T & Pbnp ISFSI Security Plan Section 2.0, Dtd 990604.Plans Withheld1999-06-0404 June 1999 Forwards Revs to Pbnp Security Plan Sections 2.1,2.4,3.1, Figures A,D & T & Pbnp ISFSI Security Plan Section 2.0, Dtd 990604.Plans Withheld NPL-99-0319, Provides Main Control Board Wiring Separation Project Status Update Rept for Pbnps,Units 1 & 21999-05-28028 May 1999 Provides Main Control Board Wiring Separation Project Status Update Rept for Pbnps,Units 1 & 2 05000301/LER-1999-003, Forwards LER 99-003-00 for Point Beach Nuclear Plant,Unit 2. Rept Is Provided in Accordance with 10CFR50.73(a)(2)(i)(B), as Any Operation or Condition Prohibited by Plant Tech Specs1999-05-28028 May 1999 Forwards LER 99-003-00 for Point Beach Nuclear Plant,Unit 2. Rept Is Provided in Accordance with 10CFR50.73(a)(2)(i)(B), as Any Operation or Condition Prohibited by Plant Tech Specs ML20196F3211999-05-11011 May 1999 Requests Proprietary WCAP-14787, W Revised Thermal Design Procedure Instrument Uncertainty Methodology for Wepc Point Beach Units 1 & 2 (Fuel Upgrade & Uprate to 1656 Mwt-NSSS Power), Be Withheld from Public Disclosure NPL-99-0242, Submits Commitment Schedule Update,Per GL 95-07 Re Pressure Locking & Thermal Binding of safety-related power-operated Gate Valves.Unit 1 Block Valve Replacement Will Be Performed During Upcoming 1999 U1R25 Outage1999-04-27027 April 1999 Submits Commitment Schedule Update,Per GL 95-07 Re Pressure Locking & Thermal Binding of safety-related power-operated Gate Valves.Unit 1 Block Valve Replacement Will Be Performed During Upcoming 1999 U1R25 Outage NPL-99-0246, Forwards 1998 Annual Monitoring Rept, for Pbnps Units 1 & 2.Revised ODCM & Environ Manual Are Encl1999-04-27027 April 1999 Forwards 1998 Annual Monitoring Rept, for Pbnps Units 1 & 2.Revised ODCM & Environ Manual Are Encl ML20206C2361999-04-22022 April 1999 Forwards 1998 Annual Rept to Stockholders of Wepc Which Includes Certified Financial Statements,Per 10CFR50.71 NPL-99-0230, Submits Clarification of Which Portions of OMa-1988 Parts 6 & 10 Are Being Utilized at Pbnp for IST Program Implementation & Cold SD & RO Justifications,Per 990218 Telcon with NRC1999-04-19019 April 1999 Submits Clarification of Which Portions of OMa-1988 Parts 6 & 10 Are Being Utilized at Pbnp for IST Program Implementation & Cold SD & RO Justifications,Per 990218 Telcon with NRC 05000301/LER-1999-002, Forwards LER 99-002-00 Re Discovery That Cable Necessary to Provide Plant Parameter Required to Be Monitored for App R Safe SD Location Was Not Routed Independent of Appropriate Fire Zone.Commitments in Rept Indicated in Italic1999-04-16016 April 1999 Forwards LER 99-002-00 Re Discovery That Cable Necessary to Provide Plant Parameter Required to Be Monitored for App R Safe SD Location Was Not Routed Independent of Appropriate Fire Zone.Commitments in Rept Indicated in Italics NPL-99-0219, Provides Final Notification of Change to Commitments Documented in LER 266/97-022-00 Re Electrical Short Circuits During CR Fire1999-04-15015 April 1999 Provides Final Notification of Change to Commitments Documented in LER 266/97-022-00 Re Electrical Short Circuits During CR Fire 05000266/LER-1999-001, Forwards LER 99-001-01,describing Discovery That Common Min Recirculation Flow Line Return to RWST for Safety Injection & Containment Spray Pumps Was Partially Frozen & Would Not Pass Flow.New Commitments Indicated in Italics i1999-04-0808 April 1999 Forwards LER 99-001-01,describing Discovery That Common Min Recirculation Flow Line Return to RWST for Safety Injection & Containment Spray Pumps Was Partially Frozen & Would Not Pass Flow.New Commitments Indicated in Italics in Rept NPL-99-0174, Confirms Completion of Requested Actions in Accordance with Required Response of GL 96-01 for Unit 2.Confirmation of Completion for Unit 1 Was Provided in Ltr Npl 98-0591,dtd 9807141999-03-30030 March 1999 Confirms Completion of Requested Actions in Accordance with Required Response of GL 96-01 for Unit 2.Confirmation of Completion for Unit 1 Was Provided in Ltr Npl 98-0591,dtd 980714 ML20206B8231999-03-30030 March 1999 Forwards Final Exercise Rept for Biennial Radiological Emergency Preparedness Exercise Conducted on 981103 for Point Beach Power Plant.One Deficiency Identified for Manitowoc County.County Corrected Deficiency Immediately NPL-99-0177, Forwards Decommissioning Funding Status Info for Pbnp,Units 1 & 2,per 10CFR50.751999-03-30030 March 1999 Forwards Decommissioning Funding Status Info for Pbnp,Units 1 & 2,per 10CFR50.75 05000301/LER-1999-001, Forwards LER 99-001-00,re Loss of Safeguards Electrical Bus During Refueling Surveillance Testing Which Resulted in Temporary Unavailability of One Train of Decay Heat Removal. Commitments Made by Util Are Identified in Italics1999-03-10010 March 1999 Forwards LER 99-001-00,re Loss of Safeguards Electrical Bus During Refueling Surveillance Testing Which Resulted in Temporary Unavailability of One Train of Decay Heat Removal. Commitments Made by Util Are Identified in Italics NPL-99-0122, Forwards Relief Requests RR-1-19 & RR-2-25,requesting Relief from Section XI of ASME B&PV Code, Rules for Inservice Exam of NPP Components, 1986 Edition,No Addenda.Requirements for Relief Apply to Third ten-yr ISI Interval for Units 1 &1999-03-0303 March 1999 Forwards Relief Requests RR-1-19 & RR-2-25,requesting Relief from Section XI of ASME B&PV Code, Rules for Inservice Exam of NPP Components, 1986 Edition,No Addenda.Requirements for Relief Apply to Third ten-yr ISI Interval for Units 1 & 2 NPL-99-0111, Informs NRC That IAW Provisions of ASME Boiler & Pressure Code,Section Xi,Paragraphs IWA-2430(d) & IWA-2430(e),WEPC Has Extended Third 10 Yr Interval for Pressure Testing Program at Pbnp,Unit 1 by 21 Months1999-03-0303 March 1999 Informs NRC That IAW Provisions of ASME Boiler & Pressure Code,Section Xi,Paragraphs IWA-2430(d) & IWA-2430(e),WEPC Has Extended Third 10 Yr Interval for Pressure Testing Program at Pbnp,Unit 1 by 21 Months NPL-99-0116, Forwards Proprietary & non-proprietary Revised Point Beach Nuclear Plant Emergency Plan IAW 10CFR50.54(q).Proprietary Plan Withheld1999-03-0101 March 1999 Forwards Proprietary & non-proprietary Revised Point Beach Nuclear Plant Emergency Plan IAW 10CFR50.54(q).Proprietary Plan Withheld NPL-99-0115, Forwards Proprietary & non-proprietary Revised EPIPs to Point Beach Nuclear Plant,Units 1 & 21999-03-0101 March 1999 Forwards Proprietary & non-proprietary Revised EPIPs to Point Beach Nuclear Plant,Units 1 & 2 NPL-99-0114, Provides Results of Wepcs Insp,Replacement & Mechanical Testing of Reactor Internals Baffle Former Bolts During Recent Point Beach Refueling Outage1999-02-25025 February 1999 Provides Results of Wepcs Insp,Replacement & Mechanical Testing of Reactor Internals Baffle Former Bolts During Recent Point Beach Refueling Outage NPL-99-0086, Documents Commitment Change Which Is to Discontinue Actions Contained in Util Ltr Dtd 970613,after NRC Approval of LAR & Lower Containment Leak Rate Limit Is Implemented. Change Is Acceptable IAW Applicable Plant Procedure1999-02-24024 February 1999 Documents Commitment Change Which Is to Discontinue Actions Contained in Util Ltr Dtd 970613,after NRC Approval of LAR & Lower Containment Leak Rate Limit Is Implemented. Change Is Acceptable IAW Applicable Plant Procedure NPL-99-0101, Forwards Proprietary & non-proprietary Version of Rev 20 to EPIP 3.2, Emergency Response Organization Notification & Revised Index.Proprietary Info Withheld1999-02-19019 February 1999 Forwards Proprietary & non-proprietary Version of Rev 20 to EPIP 3.2, Emergency Response Organization Notification & Revised Index.Proprietary Info Withheld ML20203F7301999-02-10010 February 1999 Forwards Revs to Security Plan Sections 1.2,1.3,1.4,2.1,2.5, 2,6,2.8,6.1,6.4,6.5,B-3.0,B-4.0,B-5.0 & Figure R Dtd 990210. Evaluation & Description of Plan Revs Also Encl to Assist in NRC Review.Encls Withheld NPL-99-0067, Submits 30 Day Rept of Changes & Errors Discovered in ECCS Evaluation Models for Pbnp,Unit 21999-02-0202 February 1999 Submits 30 Day Rept of Changes & Errors Discovered in ECCS Evaluation Models for Pbnp,Unit 2 NPL-99-0064, Forwards Revised TS Bases Page 15.4.4,correcting References to Pbnp FSAR Re Reactor Containment Design.Changes Are Administrative Only & Do Not Alter Facility or Operation,As Described in FSAR or Any TS Requirement1999-02-0202 February 1999 Forwards Revised TS Bases Page 15.4.4,correcting References to Pbnp FSAR Re Reactor Containment Design.Changes Are Administrative Only & Do Not Alter Facility or Operation,As Described in FSAR or Any TS Requirement NPL-98-1032, Forwards Revs to Pbnp Security Plan Sections 1.1,1.2,2.1, 2.6,2.8,6.1 & 6.4 & Revs to Pbnp ISFSI Security Plan Sections 1.0 & 7.0,per 10CFR50.54(p).Encl Withheld1999-01-27027 January 1999 Forwards Revs to Pbnp Security Plan Sections 1.1,1.2,2.1, 2.6,2.8,6.1 & 6.4 & Revs to Pbnp ISFSI Security Plan Sections 1.0 & 7.0,per 10CFR50.54(p).Encl Withheld 05000266/LER-1998-029, Forwards LER 98-029-00,describing Discovery of Isolation of Autostart Feature for Svc Water Pumps from Unit 2,safeguards Buses During Modifications1999-01-26026 January 1999 Forwards LER 98-029-00,describing Discovery of Isolation of Autostart Feature for Svc Water Pumps from Unit 2,safeguards Buses During Modifications NPL-99-0031, Informs That Wepc Reviewed Contents of NEI to NRC & Have Verified Info Provided in Ltr Pertaining to WOG Member Plants Is Applicable to Pbnp.Attachment Responds to NRC Questions by Ref to Info in 981211 NEI Ltr1999-01-15015 January 1999 Informs That Wepc Reviewed Contents of NEI to NRC & Have Verified Info Provided in Ltr Pertaining to WOG Member Plants Is Applicable to Pbnp.Attachment Responds to NRC Questions by Ref to Info in 981211 NEI Ltr NPL-99-0004, Provides Status Update on Program Activities & Schedule for Final Resolution of Items Re Verification of Seismic Piping Class Interfaces for Point Beach Nuclear Plant,Units 1 & 21999-01-11011 January 1999 Provides Status Update on Program Activities & Schedule for Final Resolution of Items Re Verification of Seismic Piping Class Interfaces for Point Beach Nuclear Plant,Units 1 & 2 NPL-99-0012, Forwards Proprietary & Nonproprietary Revs to Epips. Proprietary Version of EPIPs Withheld1999-01-0808 January 1999 Forwards Proprietary & Nonproprietary Revs to Epips. Proprietary Version of EPIPs Withheld 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059H9391990-09-13013 September 1990 Forwards Amended Response to Notice of Violations Noted in Insp Repts 50-266/89-27 & 50-301/89-26.Corrective Action: Revised Procedures Will Not Be Issued Until After Unit 2 Refueling Outage.Other Changes Anticipated by 901231 ML20059G7211990-09-0505 September 1990 Responds to Generic Ltr 90-03, Vender Interface for Safety- Related Components. Implementing Formal Vendor Interface Program for Every safety-related Component Impractical ML20028G9071990-08-31031 August 1990 Advises That long-term erosion/corrosion-induced Program for Pipe Wall Thinning in Place,Per Generic Ltr 89-08.Program Assures Erosion/Corrosion Will Not Lead to Degradation of Single & two-phase High Energy Carbon Steel Sys ML20059G8741990-08-31031 August 1990 Forwards Revised Security Plan,Per NRC .Summary of Revs Listed.Rev Withheld (Ref 10CFR3,50,70 & 73) ML20064A4711990-08-29029 August 1990 Forwards Semiannual Monitoring Rept,Jan-June 1990, Rev 1 to Process Control Program, Rev 7 to Environ Manual & Rev 5 to Odcm ML20058N6771990-08-0303 August 1990 Forwards Public Version of Revised Procedures to Emergency Plan manual.W/900813 Release Memo ML20058L1471990-08-0303 August 1990 Responds to NRC Re Weaknesses Noted in Insp Repts 50-266/90-201 & 50-301/90-201 Re Electrical Distribution. Corrective Actions:Design Basis Documentation Will Be Developed to Alleviate Weaknessess in Diesel Generators ML20058L5041990-07-30030 July 1990 Discusses & Forwards Results of fitness-for-duty Program Performance Data for 6-month Period Ending 900630 ML20055J2031990-07-25025 July 1990 Responds to NRC Bulletin 89-002 Re Insp of safety-related Anchor/Darling Model S350W Check Valves Supplied w/A193 Grade B6 Type 410 SS Retaining Block Studs.Studs Visually Inspected & No Cracks Found ML20055H7781990-07-24024 July 1990 Forwards Corrected Monthly Operating Rept for June 1990 for Point Beach Unit 2.Correction on Line 18 Regards Net Electrical Energy Generated ML20055H6621990-07-23023 July 1990 Forwards Central Files & Public Versions of Revised Epips, Including Rev 2 to EPIP 1.1.1,Rev 16 to EPIP 4.1,Rev 6 to EPIP 6.5,Rev 20 to EPIP 1.2,Rev 8 to EPIP 6.3,Rev 0 to EPIP 7.3.2,Rev 10 EPIP 10.2 & Rev 11 to EPIP 11.3 ML20058K8941990-07-23023 July 1990 Forwards June 1990 Updated FSAR for Point Beach Nuclear Plant Units 1 & 2.Steam Generator Upper Ph Guideline in Table 10.2-1 Changed from 9.3 to 9.4 ML20044A9091990-07-0606 July 1990 Responds to NRC Bulletin 90-001 Re Loss of Fill Oil in Transmitters Mfg by Rosemount.None of Listed Transmitters Installed at Plant in Aug 1988 Identified as Having High Failure Fraction Due to Loss of Fill Oil ML20055D4421990-07-0303 July 1990 Forwards Reactor Containment Bldg Integrated Leak Rate Test Point Beach Nuclear Plant Unit 1,1990, Summary Rept ML20055D3471990-06-29029 June 1990 Provides Addl Response to Bulletin 88-008, Thermal Stresses in Piping Connected to Rcss. Engineering Evaluations Performed to Assure Code Compliance Due to Unanalyzed Condition of Thermal Stratification Addressed ML20055D6221990-06-29029 June 1990 Provides Suppl to Re Loss of All Ac Power.Test Demonstrated That Ventilation Mod & Recalibration of High Temp Trip for Auxiliary Power Diesel Improved Performance of Gas Turbine Generator as Alternate Ac Source ML20055D2291990-06-22022 June 1990 Informs NRC That Gj Maxfield Promoted to Plant Manager effective,900701 ML20055D6231990-06-22022 June 1990 Advises of Decision to Proceed W/Leak Testing of Sys During Plant Refueling Outage Due to Delay in Delivery of Gamma-Metrics Hardware Fix Kits.Test Revealed That Both in-containment Cable & Detector Assembly Cable Had Leaks ML20044A0011990-06-18018 June 1990 Provides Current Implementation Status of Generic Safety Issues at Plant,In Response to Generic Ltr 90-04 ML20043D6511990-05-25025 May 1990 Discusses Cycle 18 Reload on 900519,following 7-wk Refueling & Maint Outage.Reload SER for Cycle 18 Demonstrates That No Unreviewed Safety Questions,As Defined in 10CFR50.59, Involved in Operation of Unit During Cycle ML20043B1481990-05-18018 May 1990 Advises That Necessary Info Received from Westinghouse Re Revised Administrative Controls for NRC Bulletin 88-002, Rapidly Propagating Fatique Cracks in Steam Generator Tubes. ML20043B1101990-05-17017 May 1990 Documents Status of Evaluations Committed to Be Performed Re IE Bulletin 79-14 Program.Support CH-151-4-H50 Modified During Unit 1 Refueling Outage & Now in Code Compliance. Meeting Proposed During Wks of 900618 or 900716 ML20043A9921990-05-16016 May 1990 Advises of Typo in Item 2.C Re Emergency Diesel Generator Meter Accuracy in Submittal Re Corrective Actions in Response to Concerns Identified During Electrical Insp.Meter Calibr Reading Should Be 3,050 Kw Not 350 Kw ML20043B0481990-05-16016 May 1990 Updates 890330 Response to NRC Bulletin 88-010, Nonconforming Molded Case Circuit Breakers. Util Will Replace Unit 1 Inverter & Battery Charger Circuit Breakers within 30 Days After Receipt & QA Verification ML20043A7631990-05-15015 May 1990 Responds to Notice of Violation & Forwards Civil Penalty in Amount of $87,000 for Violations Noted in Insp Repts 50-266/89-32,50-266/89-33,50-301/89-32 & 50-301/89-33. Addl Employees Added in QA & Corporate Nuclear Engineering ML20042H0201990-05-10010 May 1990 Forwards List of Concerns Identified at 900417 Electrical Insp Exit Meeting to Discuss Preliminary Findings of Special Electrical Insp Conducted on 900319-0412 Re Adequacy of Electrical Distribution Sys ML20043A2181990-05-10010 May 1990 Forwards Nonproprietary & Proprietary Version of Point Beach Nuclear Plant,Emergency Plan Exercise,900314. ML20042G7441990-05-0909 May 1990 Forwards LER 90-003-00 ML20042G7361990-05-0808 May 1990 Forwards LER 90-004-00 ML20042E4571990-04-10010 April 1990 Documents Basis for Request for Temporary Waiver of Compliance of Tech Spec 15.3.7.A.1.e Re Diesel Generator Fuel Oil Supply ML20012F2961990-03-29029 March 1990 Withdraws Tech Spec Change Request 120 Re Staff Organization Changes & Deletion of Organizational Charts,Based on Further Corporate Restructuring within Util ML20012D8301990-03-20020 March 1990 Responds to Generic Ltr 89-19, Safety Implication of Control Sys in LWR Nuclear Power Plants. No Limiting Condition for Operation Required for Overfill Protection Sys at Plant ML20012D4241990-03-0808 March 1990 Forwards Public Version of Revised Epips,Including Rev 17 to EPIP 1.3,Rev 8 to EPIP 3.1,Rev 15 to EPIP 4.1,Rev 1 to EPIP 6.7,Rev 1 to EPIP 7.1,Rev 11 to EPIP 7.2.1 & Rev 11 to EPIP 7.2.2 ML20011F7531990-02-26026 February 1990 Informs NRC of Apparent Inconsistency Between Min Level of Boric Acid Solution to Be Maintained in Boric Acid Storage Tanks Per Tech Specs & Amount of Deliverable Boric Acid Assumed in Safety Analyses ML20006B7091990-01-25025 January 1990 Responds to NRC Bulletin 89-002 Re Check Valve Bolting Insp. All Anchor-Darling Model S35OW Check Valves Inspected for Cracked Internal Bolting During Refueling Outage of Unit.No Indications of Cracks Found ML20006A3381990-01-18018 January 1990 Forwards PDR & Central Files Versions of Rev 16 to EPIP 9.2 & Forms, Radiological Dose Evaluation. ML20006A3411990-01-16016 January 1990 Forwards Rev 16 to EPIP 9.2, Radiological Dose Evaluation to Be Inserted in EPIP Manual ML20005G0901990-01-12012 January 1990 Responds to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Outside of Intake Structure Will Be Inspected for Excessive Corrosion on Semiannual Basis & Forebay & Pumphouse Inspected ML20005G1751990-01-12012 January 1990 Responds to NRC 891213 Ltr Re Violations Noted in Insp Repts 50-266/89-30 & 50-301/89-30.Corrective Action:Procedure RP-6A, Steam Generator Crevice Flush (Vacuum Mode), Initiated ML20005H0551990-01-11011 January 1990 Responds to NRC Bulletin 89-003, Potential Loss of Required Shutdown Margin During Refueling Operations. Util Will Provide Specific Training to All Members Responsible for Refueling Operation to Emphasize Importance of Procedures ML20005G9031990-01-0909 January 1990 Forwards Monthly Operating Repts for Dec 1989 for Point Beach Nuclear Plant Units 1 & 2 & Revised Monthly Operating Rept for Nov for Point Beach Unit 2 ML20005G5641990-01-0808 January 1990 Updates Progress Made on Issues Discussed in Insp Repts 50-266/89-12 & 50-301/89-11 Re Emergency Diesel Generator Vertical Slice SSFI Conducted by Util.By Jul 1990,revised Calculation Re as-built Configuration Will Be Performed ML20005E5441989-12-29029 December 1989 Describes Actions & Insps Completed During Recent U2R15 Refueling Cycle & Proposed Schedule for Completion of NRC Bulletin 88-008 Requirements,Per Util 881221 & 890616 Ltrs. Extension Requested Until 900631 to Submit Data Evaluation ML20005E5451989-12-28028 December 1989 Advises That Addl Info Required from Westinghouse to Meet Util 890621 Commitment to Adopt Administrative Control Re Rapidly Propagating Fatigue Cracks in Steam Generator Tubes Per NRC Bulletin 88-002.Info Anticipated by End of Mar 1990 ML20005E5381989-12-27027 December 1989 Provides Update of Status of Implementation of Resolution of Human Engineering Discrepancies Documented During Dcrdr. Lighting Intended to Document Deficiencies Per NUREG-0700, Eleven Human Engineering Discrepancy Computers Resolved ML19354D5781989-12-21021 December 1989 Certifies Implementation of Fitness for Duty Program Which Meets Requirements of 10CFR26 for All Personnel Having Unescorted Access to Plant Protected Areas.Periodic Mandatory Random Chemical Testing Will Commence on 900103 ML20005D8071989-12-21021 December 1989 Forwards Response to Violations Noted in Insp Repts 50-266/89-29 & 50-301/89-29.Response Withheld (Ref 10CFR73.21) ML20005E2301989-12-21021 December 1989 Forwards Reactor Containment Bldg Integrated Leak Rate Test Point Beach Nuclear Plant Unit 2, Summary Rept,Per 10CFR50,App J.Type A,B & C Leak Test Results Provided ML20042D2391989-12-21021 December 1989 Responds to Violations Noted in Insp Repts 50-266/89-27 & 50-301/89-26.Corrective Actions:Superintendent of Health Physics Discussed Log Book Entry Requirements W/Health Physics Contractor Site Coordinator ML19354D6231989-12-15015 December 1989 Responds to Generic Ltr 89-10 Re safety-related motor- Operated Valve Testing & Surveillance.Util Intends to Meet All Recommendations Discussed in Ltr Except for Item C Re Changing motor-operated Valve Switch Settings 1990-09-05
[Table view] |
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t ntsconsm Electnc mencourany 231 W. MICHIGAN, P.o. BOX 2046, MILWAUKEE. WI 53201 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U.S. NUCLEAR REGULATORY COMMISSION Washington, D.C. 20555 Attention: Mr. D. G. Eisenhut, Director Division of Licensing Gentlemen:
DOCKETS 50-266 and 50-301 9ESPONSE TO GENERIC LETTER 83-28 REQUIRED'AClIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 On July 11, 1983, we received generic letter 83-28 entitled " Required Actions Based on Generic Implications of Salem ATWS Events" which was dated July 8, 1983. Generic letter 83-28, and the enclosure to the letter, presented the Commission's requirements for intermediate-term actions to be taken by licensees as a result of the NRC review of the Salem anticipated transient without scram (ATWS) events. These actions were characterized as addressing i issues related to reactor trip system reliability and general management
! capability. However, many of the actions discussed in the enclosure were related to all safety-related systems within the plant.
l l Licensees were required pursuant to 10CFR 50.54(F) to furnish a l
report on the status of current conformance with the positions contained in generic letter 83-28 and to present plans and schedules for any improvements for conformance with the positions. The attachment to this letter provides this information for the Point Beach Nuclear Plant Units 1 and 2 in an item by item format. You will note that as suggested in your letter, our actions i regarding some of these items are dependent upon the results of our participa-tion in several Owners' Group activities.
l 8311170222 831107
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PDR ADOCK 05000266 P PDR
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Mr. H. R. Denton -
2- November 7, 1983 We trust the attached discussions are responsive to your action requirements. Should you have any questions regarding our proposed activities or the schedules for accomplishment of those activities, please let me know.
Very truly yours, fg$
C.W. Fay Vice President-Nuclear Power Enclosures Copies to NRC Resident Inspector Subscribpd and sworn to before me this 8Hs day of November 1983.
Notary Public, State of Wisconsin.
My Commission exp h [5 pers awe /I.
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O RESPONSE TO NRC GENERIC LETTER 83-28 POINT BEACH NUCLEAR PLANT l
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1.1 Post-Trip Review
' Activities currently performed at PBNP which constitute post-trip review of unplanned reactor shutdowns are being reevaluated. .It is anticipated that modifications to our current practices will be instituted that will result in a post-trip review procedure reflecting the guidelines of the INPO draft good practice OP-211 entitled, " Post-Trip Review." The modified procedure will be in place by February 29, 1984.
A description of current practices associated with post-trip review follows:
1.1.1 Criteria for Restart NOTE: RESTART MEANS ATTAINMENT OF CRITICALITY PER PLANT OPERATING PROCEDURES.
1.1.1-1 The cause of the trip must be known and have been corrected. If these conditions are not met, a Manager's Supervisory Staff review of the event is necessary -
before any restart activities are resumed.
1.1.1-2 Equipment which did not function properly during and following the transient must be repaired or a determi-nation made by the Duty Shift Superintendent and Duty &
Call _ Superintendent that the equipment is not required by Technical Specifications er r,ecessary for safe and reliable operation subsequent to restart.
1.1.1 ~ Equipment'which is necessary for critical operation under Technical Specifications is verified operable by a precriticality checklist.
1.1.1-4 Permission to restart must be obtained from the Duty &
Call. Superintendent.
1.1.2 Responsibilities and authorities of personnel who perform the review:
1.1.2-1 The Duty Shift Superintendent is responsible for the safe operation of PBNP Units 1-and 2. He has the authority to maintain either unit in a shutdown condition until he is satisfied that it is safe to change the condition of the unit. He.is assisted in making necessary decisions of this nature by the Duty &
Call Superintendent and Duty Technical Advisor.
1.1.2-2 The Duty _& Call Superintendent is responsible to act as an advisor to the Duty Shift Superintendent on matters relating to safe operation of PBNP Units 1 and 2. He L
Q may make a determination that further evaluation by the Manager's Supervisory Staff is necessary before the condition of a shutdown unit is changed.
1.1.2-3 The Duty Technical Advisor is responsible to provide to the Duty Shift Superintendent and Duty & Call Superintendent a diagn'osis of plant response during off-normal events.
L 1.1.3 Necessary Training and Qualifications l 1.1.3-1 The Duty Shift Superintendent must have, as a minimum, 5 years power plan't experience of which one year is in a nuclear power plant and hold a Senior Reactor Operator's j License. It should be noted that there are 8 Duty Shift Superintendents employed and each has more than 13 i years' experience working at PBNP in an operating crew.
l 1.1.3-2 The Duty & Call Superintendent is an indiv.idual l appointed by the Manager --PBNP, taking into
- l consideration the following:
- 1. Past SRO license and experience.
- 2. Current SRO license at PBNP.
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- 3. Professional degree and extensive nuclear power l
plant experience.
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( 4. Member of Manager's Supervisory Staff as defined in Technical Specifications.
It should be noted that currently there are 5 Duty &
Call Superintendents, 4 currently hold and one has held a.SRO license at PBNP. All have greater than 4.5 years working experience at PBNP.
1.1.3-3 The Duty Technical Advisor must have a bachelor's degree in an engineering or physical science discipline and 18 months of nuclear power plant experience of which 12 months experience must be at PBNP as well as success-
-fully completing the Duty Technical Advisor training program. It should be noted that there are currently 15 Duty Technical Advisors who meet these requirements.
1.1.4 Sources of Plant Information Necessary for Review Following any unscheduled reactor shutdown, a post-trip data package is put together for review by the operating crew. It consists of the following:
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I l'. Strip chart recordings for primary system parameters, t
- 2. Strip chart recordings for secondary system parameters, 7
- 3. Sequence of event listing from plant process control,
- 4. Analog trend post-trip review from plant process computer,
- 5. Operator observation of alarms (including first outs),
instrumentation displays including recorders, equipment status lights as well as first hand knowledge of equipment which has not functioned as required by emergency operating procedures or other plant operating requirements.
1.1.5 Methods and Criteria for Comparing Event Information With Expected Plant Response From the sources of information listad in 1.1.4 above, an evaluation of the transient is made by the Duty Shift Superintendent and the Duty Technical Advisor. This evaluation
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' must satisfy the restart criteria outlined in Section 1.1.1 above before any decision to restart can be made. The Duty Shift Superintendent. reviews his decision with the Duty & Call Superintendent who will concur if he believes the' restart i criteria are' met. The Duty & Call Superintendent will contact the Manager-PBNP, if he is available, to inform him of the
, circumstances and any decision regarding restart.
- 1.1.6 Criteria for Determining Need for Independent Assessment The determination that an independent assessment of the event by the Manager's Supervisory Staff can occur at three levels in the event review process:
- 1. The Duty Shift Superintendent is not satisfied that the restart criteria of 1.1.1 above are met.
- 2. The Duty & Call Superintendent is not satisfied that the restart criteria of 1.1.1 above are met.
- 3. The Manager - PBNP, if available, is not satisfied that the restart criteria are met.
', The physical evidence necessary to support an independent analysis i
of an event in which the cause is not immediately obvious is retained in accordance with plant practice and procedure.
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1.1.7 Systematic Method to Assess Unscheduled Reactor Shutdowns Although 'it is clear that current plant. practices do provide a systematic method for the assessment of unscheduled reactor shutdown, the present practice will be reviewed as stated above.
This-review, taking into consideration the draft INPO good practice
" Post-Trip Review", will be completed by February 29, 1984.
1.2 ' Post-Trip Review Data and Information Capability 1.2.1 Sequence of Events Capability 1.2.1-1 Description of Equipment This capability is provided by the plant computer system (a Westinghouse P-250 computer). There are currently separate computers for e;.ch unit.
1.2.1-2 Parameters Monitored A list of the sequence of events parameters monitored by the current computer system is provided in Attachment A.
The same parameters are monitored for both units.
1.2.1-3 Time Discrimination The time discrimination between events on the current computer system is 1/60 of a second, (16-2/3 msec).
1.2.1-4 Format Sequence of events data is output to the alarm type-writer. The output starts out with a line listing the time in hours, minutes and seconds of the first event. Each event is then output with its address, point description, status (open or close) and the cycle count (in 60ths of a second) since the first event. The data is output after either 25 events have occurred or after one minute has elapsed.
1.2.1-5 Data Retention The computer can collect data for a second " sequence of events" while the first is being output. When printout of the second sequence begins, data collection resumes for the next sequence. Once data is output to the alarm typewriter, it is no longer retained by the computer.
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1.2.1-6 Power Sources The Unit 1 computer power supply is 2831 (a nonvital motor control center supplied from Unit 2 803 safeguards 480v bus) and the Unit 2 computer power supply is 1831 (a nonvital motor control center supplied from Unit 1 B03 safeguards 480v bus).
1.2.2 Time History of Analog Variables (Post-Trip Review) 1.2.2-1 Equipment Description This capability is provided by the plant computer system. There are currently separate computers for each unit.
1.2.2-2 Parameters Monitored A list of the post-trip review parameters monitored by the current computer system is provided in Attachment B.
There is a small group sampled every 2 seconds and another larger group sampled every 8 seconds. Parameters which indicate power mismatched between the primary and secondary system were selected for the 2-second group.
In addition to the critical monitored parameters, several other parameters are monitored in the plant which provide more detailed information on these safety systems and other plant systems necessary to take the plant to a safe shutdown condition.
1.2.2-3 Duration of Time History l
l l The parameters in the 2-second group are collected for i approximately 8 seconds before the trip and 8 seconds after the trip. The parameters in the 8-second group are collected for approximately 2 minutes before the trip and 3 minutes after the trip.
1.2.2-4 Format The post-trip review data is output to a typewriter with l a column for each parameter. The pretrip data starts l off the column and is separated from the post-trip data by a line listing the trip time. The actual time of data collection for sub groups of parameters is included with each line of data.
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1.2.2-5 Data Retention Data collection is suspended while the printout takes place. Once data is output to the typewriter, it is no longer retained by the computer.
1.2.2-6 Power Sources See 1.2.1-6 above.
1.2.3 The following is a brief summary of the information that is available in the control room and is utilized in the evaluation of plant behavior during unplanned reactor shutdowns:
- 1. Rod position indicators,
- 2. Reactor trip breaker status lights,
- 3. Nuclear Power indicators and recorders including overpower and delta flux recorders,
- 4. AT and AT setpoints indicators and recorders,
- 5. . Pressurizer pressure indicators and recorder,
- 6. RCS wide range pressure recorder,
'7. RCS T hot and T cold temperature recorder,
- 8. Pressurizer level indicators and recorder,
- 9. Pressurizer level program recorder,
- 10. RCS T,yg and T ref recorder and indicators,
- 11. Various pressurizer system temperature indicators,
- 12. Various CVCS system parameter indicators and recorders including boration and dilution flow recorders,
- 13. First-out alarms for reactor trips, safeguards actuation and turbine trips,
- 14. Reactor protection and safeguards system bistable status lights,
- 15. Steam generator. narrow range level indicators and wide range level recorders,
- 16. Steam. flow / feed flow indicators and recorders,
- 17. -Steam generator pressure indicators,
- 18. Turbine parameter recorders,
- 19. Condenser pressure,
- 20. Status indication of various valves and pumps associated with important systems including safeguards,
- 21. Four computer trend recorders, one dedicated to subcooling and one to VCT level, the other two are operator selectable.
- 22. Various other system alarms including safety system channel alerts.
1.2.4 Planned Changes to Existing Data and Information Capability The present plant computers will be replaced by an updated computer system monitoring both units. This sytem should be in operation in 1985.
This computer system will. consist of redundant computers with automatic pickup if the operating computer fails, as well as have battery backed instrument bus power supplies.
The sequence of events monitoring capability will be changed as follows:
- 1. The list of parameters monitored will be expanded. A list of the additional points is Attachment C.
- 2. The t,ime discrimination will be improved to 5 milliseconds.
- 3. The maximum number of events per sequence will be enlarged to 100 and the capacity for up to 10 events will be provided.
- 4. The sequence of events reports will be outputted to a printer and stored on a disc and/or magtape for later reprinting, if desired.
The post-trip review capability will be changed as follows:
- 1. The current 2-second group will be sampled at one-second intervals for 48 seconds before and after the trip.
- 2. The current 8-second group will be sampled at five-second intervals for 4 minutes before and after the trip.
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- 3. The post-trip review report will be outputted to a printer and stored on disc and/or magtape for later reprinting, if desired.
2.1 Equipment Classification and Vendor Interface An ',nitial review has confirmed that those safety-related components necessary to assure a reactor trip are included within the quality assurance scope at Point Beach Nuclear Plant. To further ensure the proper quality assurance coverage of safety-related systems and components, a detailed intensive review of the drawings used to categorize the quality assurance scope has been initiated. This review, scheduled to be completed by February 29, 1984, will include all aspects of those safety-related systems and components including those necessary to provide a reactor trip function.
A program designed to control vendor technical information has existed at Point Beach since initial construction. This program includes the distribution of vendor technical information to the appropriate personnel so that this information can be utilized and incorporated in plant procedures, as necessary. Also included in this program is a control process for vendor technical manuals. Although this program undergoes continuing review in the normal course of business, a specific review will be conducted to ensure all aspects of this program function efficiently. This specific review will be conducted in concert with other INP0 sponsored efforts in the control and dissemination of vendor information.
Wisconsin Electric is a participating member of the INPO Nuclear Utility Task Action Committee (NUTAC) which has been chartered to study and make recommendations concerning vendor interface in response to Section 2.2.2 of this letter. (See Section 2.2.2 for further detail.) We would expect that the recommendations of the NUTAC will envelop vendors that provide components whose function is to trip the reactor. Also, since these components fall under the NSSS scope of l supply, we believe vendor feedback and communication mechanisms l already exist through the NSSS vendor.
2.2 Equipment Classification and Vendor Interface 2.2.1 Equipment classification and the administration of the program is accomplished through existing quality assurance procedures.
The Quality Assurance and Reliability Manual (Point Beach Nuclear Plant QA Volume II) contains a rigorous listing of those systems and major components which are defined as l QA-scope (includes those systems and components considered to be safety-related). The QA-scope boundaries of the l
systems and components are then further delineated in drawings, lists and tables to clearly establish the scope ;
and extent of the equipment classification. Specific, more detailed classifications such as environmental equipment j qualification or containment isolation functions are l included in the lists and tables.
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-The information in QA Volume II pertaining to equipment classification is distributed and is available to all Nuclear Power Department (NPD) personnel. This information is used to establish applicability of the Quality Assurance Program to equipment and services pertaining to activities such as procurement, maintenance, modifications and design. Documents, such as purchase requisitions, maintenance and modification requests, maintenance procedures, etc. , are clearly marked to indicate their QA applicability. These documents receive appropriate QA reviews to assure inclusien of all pertinent technical and quality assurance requirements. Items or activities determined to be QA-scope are subjected to rigorous controls as defined in each NPD organization's quality assur-ance and administrative procedures.
Compliance with procedures and the described information handling system is verified through technical audits by the on-site audit group and through overall QA Program audits by the Quality Assurance Division (off-site organization).
2.2.2 As indicated in the response to Section 2.1, Wisconsin Electric is a participating member of the INPO NUTAC regarding Section 2.2.2. The NUTAC is formulating recommendations and guidance to be used by utilities in responding to the specific issues in this section. Wisconsin Electric will evaluate these recom-mendations when they become available and at that time will establish a specific response to this section. The NUTAC has established a goal of February 1,1984 for completion of this task. In concert with this schedule, we would expect to have a more complete response to this section by April 1, 1984.
3.1 Post-Maintenance Testing (Reactor Trip System Components) c 3.1.1 Post-Maintenance Operability Testing Review It has been a practice at P3NP to verify that instrumentation and control devices (including breakers) associated with the reactor trip system are verified to be operable in accordance with Technical Specification requirements prior to placing the equipment in service after maintenance. The verification is accomplished utilizing test requirements contained in periodic test procedures and calibration procedures employed to satisfy
' the operability verification of this equipment required in Technical Specifications.
It has also been determined that no administrative requirement exists which requires this verification to be done. Therefore, maintenance request procedures will be modified prior to February 29, 1984, to include such a requirement for post-maintenance testing of equipment.
3.1.2 Vendor & Engineering Recommendations Although it is believed that the testing currently performed on reactor trip system equipment includes appropriate guidance concerning vendor and engineering recommendations, a review of test procedures, technical manuals and other equipment information will be completed and any findings incorporated into appropriate test procedures by February 29, 1984.
3.1. 3 Post-Maintenance Test Requirements Which Degrade Safety No post-maintenance test requirements have been identified in Technical Specifications that degrade safety.
3.2 Post-Maintenance Testing (All Other Safety-Related Components) 3.2.1 Post-Maintenance Operability Testing Review The safety-related equipment discussed in this section is divided into the following general categories; pumps and valves, piping, instrumentation and controls, and breakers.
In addition to periodic surveillance testing, it has been our practice to perform post-maintenance testing on safety-related pumps and valves. This testing is. performed in accordance with the guidelines of ASME Code Section XI.
This post-maintenance testing is performed by completing applicable portions of periodic tests employed to monitor the inservice performance of these pumps and valves. The same acceptance criteria used in evaluating performance during periodic testing is used in evaluating performance after maintenance.
Following repairs to piping in safety-related systems' post-maintenance testing as delineated by the applicable codes or standards is performed. This testing may include various methods such as dye penetrant, visual, radiographic, hydrostatic, or other nondestructive test methods.
- In addition to the instrumentation and control system testing mentioned in a previous section, it has been our practice that safety-related instrumentation and control devices be verified operable in accordance with Technical Specifica-tion prior to placing the equipment in service after mainten-ance. The verification is accomplished utilizing test
, requirements contained in periodic test procedures and cali-l bration procedures employed to satisfy the operability verification of this equipment required by Technical Specifica-tions.
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The' post-maintenance testing described above is adequate
'to demonstrate that safety-related pumps,; valves,. piping systems and' instrumentation;and controls are capable of 1 performing their safety functions following repairs. It
'has also been determined that, although.this testing is being performed, adequate. administrative controls do not
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exist.to require that.this. verification be done. .Therefore, in place administrative controls, such as the Maintenance' .
Request Procedure, will'be modified prior to February 29, 1984, to include such a requirement for post-maintenance testing
, of safety-related equipment.
The adequacy of post-maintenance testing of safety-related i breakers is still under investigation. In addit. ion.to checking proper protective tripping setpoints,. breakers used e to supply power to safety-related pumps or fans are presently
, tested by-verifying that the applicable pump or fan can be~
- . started and stopped. . Safety-related bus breakers are also post-maintenance cycled to verify their operation. .The i' investigation to determine if this testing is adequate will.be completed prior to April 30,_1984.
- 3.2.2 Vendor and Engineering Recommendations Although it is believed that the testing currently performed on the majority of safety-related equipment includes appropriate guidance concerning vendor and engineering recommendations, a review of test procedures, technical manuals and other equip-ment information will be completed and any findings incorporated into appropriate test procedures by November 1, 1985.
j 3.2.3 Post-Maintenance Test Requirements Which Degrade Safety i
Although, to date, no post-maintenance test requirements have been identified in Technical Specifications that degrade safety, some may be identified as a result of the detailed review discussed above. You will be informed if any are identified.
4.1 Reactor Trip System Reliability (Vendor-Related Modifications)
The, Licensee has reviewed and verified that the reactor trip system vendor-related modifications applicable to the Point Beach Nuclear Plant, specifically modifications in accordance with Westinghouse document NCD-ELEC-18 for the DB-50 switchgear, have been implemented.
4.2 Reactor Trip Preventive Maintenance-(Existing) 4.2.1 Reactor trip and bypass breakers are inspected annually during ea::h refueling. The inspection is made per the DIS-50 breaker technical manual. The UV trip attachment is cleaned (if
- . . . . . . - - ~ . .
necessary) and. lubricated with silicone spray. A discussion of-Licensee's maintenance program for reactor trip breakers was also.provided in our response.to IE Bulletin No. 83-01 i
dated March 3, 1983.
4.2.2 The dropout voltage of the UV trip attachment is checke'd
'at each refueling inspection and is recorded. This infor-mation is compared with prev A measurements in order to detect signs of. degradation whict may warrant more involved checks or replacement.'
.4.~2.3 'We.have no. records of formal life testing; however, Westing-t house letter NS-EPR-2737 to Mr. H. Denton of the NRC, dated March 22, 1983,' makes reference to testing done on UV trip attachments in 1972. The letter states that a UV trip device 4
"was successfully tested more than 8000 operations without malfunctions." It is doubtful .that our reactor trip breaker
- .would experience that many operations in the life of the plant. . Life cycle testing of the shunt trip attachment and
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3' the.undervoltage trip attachment of the reactor trip switch-gear is presently being conducted by Westinghouse for the Westinghouse Owners Group. This program is aimed toward establishing the service' life of these devices and sub-stantiating periodic test requirements with proper mainten-ance. The results of this program may be. factored into maintenance, replacement and qualification programs if necessary. The test program is scheduled for completion in the.second quarter of 1984.
- 4.2.4 We have no knowledge of any periodic replacement of breakers or components required or recommended by Westinghouse.
P The conclusion of the program discussed in 4.2.3 may result in periodic replacement criteria.
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4.3 Reactor Trip System Reliability (Automatic Actuation of Shunt Trip Attachment for Westinghouse and B&W Plants)
Wisconsin Electric is working closely with the Westinghouse
- Owners Group and Westinghouse in developing a design modifica-l tion relating to reactor trip system reliability. The design modification will incorporate'both the requirements for automatic actuation of the shunt trip attachment and the capability t for the on-line surveillance requirement. A detailed generic design package for incorporation of an automatic shunt trip feature was developed under the sponsorship on the Westinghouse Owners Group and submitted to the NRC in Mr. J. J. Sheppard's letter dated
- .. June 14, 1983.
This generic design package for an automatic shunt trip modifica-
- tion included the design basis, functional requirements, conceptual design and discussion of the conformance to safety criteria. The
. design of the system includes hard-wired component installation provisions for on-line surveillance testing that independently
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verifies by manual means the operability of the UVTA and the automatic shunt trip on the main reactor trip breakers.
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The NRC has issued a safety evaluation report (SER) on this generic design as an attachment to Mr. Darrell G. Eisenhut's letter to Mr. Sheppard dated August 10, 1983. The SER concluded that the conceptual design was generally acceptable and noted several exceptions. The SER included a listing of plant-specific
-information required by the NRC to perform the plant specific reviews.
Wisconsin Electric is presently reviewing this generic design modification and coordinating with Westinghouse in the develop-ment of a plant specific design. We are committed to incorpora-tion of an automatic shunt trip actuation modification and intend to incorporate the provisions for on-line testability. It is anticipated that the details for a Point Beach specific modifica-tion will be finalized over the next six months. At that time we will prepare an ammendment to this submittal to provide a report describing the modification and provide the additional information required by the NRC's SER of the generic design package. We expect to submit this report by May 1984. Assuming NRC approval of this plant-specific modification by August 1, 1984, we would expect to install the modification on each Point Beach unit during the next unit refueling shutdown. These are presently scheduled for fall 1984 for Unit 2 and spring 1985 for Unit 1.
4.4 Not Applicable To Point Beach Nuclear Plant 4.5 Reactor Trip System Reliability 4.5.1 Diverse Trip Features The current design of the reactor trip system does not include a shunt coil trip of the reactor trip breakers when an auto-matic reactor trip signal is generated. Periodic testing of all automatic reactor trip functions can be accomplished on line. Modifications will be made to add a shunt coil trip as well as the capability for on-line independent testing of each trip device. The schedule for accomplishing this modification is outlined in Section 4.3.
4.5.2 On-Line Testing Design The reactor protection system at PBNP is designed for on-line testing of its components except for the manual trip func-tion from the main control board. This portion of the reactor trip system is tested prior to startup each refueling shut-down. The manual reactor trip actuates both the shunt and undervoltage coil trip devices. To accomplish independent l testing of these devices without use of jumpers, lifted leads l
or pulled fuses, modifications to this equipment will be made in conjunction with modifications described in Section 4.5.1 above.
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4.5.3 Reactor Trip System Performance fhe reactor trip system testing interval based on operating
' experience.is considered adequatc. Throughout the approximate ten year history of PBNP, the reactor trip system has required only minor maintenance due to nonconservative component failures and has resulted in very few incidents due to operator error.
WCAP-10271 submitted by the Westinghouse Owners Group to the
.NRC in January 1983 documents a more formal evaluation of the impact on reactor protection system (RPS) unavailability of current and extended surveillance intervals. Supplement 1 to WCAP-10271 which was submitted in September 1983 is an extension of the evaluation and provides a discussion of.
component wearout caused by testing. The conclusion of WCAP-10271 and Supplement 1 is that less frequent testing of RPS components is warranted and will result in an improvement in overall plant . safety and equipment reliability.
We shall follow the NRC's review and disposition of these documents and will consider adjustments to our testing intervals based on the resolution of the issues discussed in these documents.
ATTACHMENT A SEQUENCE OF EVENTS PARAMETERS P-250 COMPUTER Computer Address Description F0403D Loop "A" Low Flow 2/3 Trip F0423D Loop "B" Low Flow 2/3 Trip LO406D Steam Generator "A" Low-Low Level 2/3 Trip LO426D Steam Generator "B" Low-Low Level 2/3 Trip LO483D Pressurizer High Level 2/3 Trip N0005D PR High Power 2/4 Trip N0010D Low Power Range 2/4 Trip N0020D IR 35 High Flux Trip Train "A" N0021D IR 36 High Flux Trip Train "B" N0030D SR 31 High Flux Trip Train "A" N0031D SR 32 High Flux Trip Train "B" P0399D Turbine AST,0il Low 2/3 Trip P0403D "A" Steam Pressure Low SI 2/3 Trip PO423D "B" Steam Pressure Low SI 2/3 Trip PO483D Pressurizer High Pressure 2/3 Trip PO488D Pressurizer Low Pressure 2/4 Trip P1003D Containment High Pressure SI 2/3 Trip T0498D- Overtemperature ATsp1 2/4 Trip T0499D Overpower ATsp2 2/4 Trip
- V0324D RCP Bus Undervolt 2/4 Trip Y0004D Manual Reactor Trip ,
Y0005D Manual Reactor Trip Y0006D Reactor Trip Breaker "A" Y0007D Reactor Trip Breaker "B" Y0026D Reactor Bypass Breaker "A" Y0027D Reactor Bypass Breaker "B" Y0335D Unit On-Line output Breaker YO394D Turbine Stop Valve Closed 2/2 Trip YO400D Reactor Coolant Pump "A" Breaker - Open - Trip YO401D Steam Generator "A" Low-Level SF/FF Mismatch Trip YO420D Reactor Coolant Pump "B" Breaker - Open - Trip Y0421D Steam Generator "B" Low-Level SF/FF Mismatch Trip
, .YO480D Pressurizer Low Pressure SI Trip YO920D Manual SI Trip Train "A" YO921D Manual SI Trip Train "B"
POST TRIP REVIEW P-250 COMPUTER Eight Second Group (continued)
Computer Address Description Units F0405C Steam Generator "A" Steam Flow F464 And F0406C Steam Generator "A" Steam Flow F465 KBH F0423C 5 team Generator "B" Feedwater Flow F476 KBH F0424C Steam Generator "B" Feedwater Flow F477 KBH
'F0425C Steam Generator "B" Steam Flow F474 KBH ;.
F0426C Steam Generator "B" Steam Flow F475 KBH~
PO400A Steam Generator "A" Steam Pressure PT-468 psig P0401A Steam Generator "A" Steam Pressure PT-469 psig P0402A Steam Generator "A" Steam Pressure PT-482 psig P0420A Steam Generator "B" Steam Pressure PT-478 psig P0421A Steam Generator "B" Steam Pressure PT-479 psig P0422A Steam Generator "B" Steam Pressure PT-483 psig l T0400A RCLA 1 Tavg T40lW Deg F T0401A' RCLA 2 Tavg T402W Deg F T0403A RCLA 1 AT T405P Deg F T0404A RCLA 2 AT T406P Deg F T0407A RCLA Overpower ATspi T40lS Deg F T0408A RCLA Overpower ATsp2 T402S Deg F T0410A RCLA Overtemp ATsp1 T405D Deg F T0411A RCLA Overtemp ATsp2 T406D Deg F T0420A RCLB 1 Tavg T403W Deg F T0421A RCLB 2 Tavg T404W Deg F T0423A RCLB l'AT T407P Deg F T0424A RCLB 2 AT T408P Deg F T0427A RCLB Overpower ATsp1 T403S Deg F T0428A RCLB Overpower ATsp2 T404S Deg F
, T0430A RCLB Overtemp ATspl T407D Deg F i
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POST TRIP REVIEW P-250 COMPUTER Eight Second Group (Continued)
Computer Address Description Units T0431A RCLB Overtemp ATsp2 T408D Deq F LO480A Pressurizer Level LT426 PC LO481A Pressurizer Level LT427 PC LO482A Pressurizer Level LT428 PC LO483A Pressurizer Level Cont Setpoint LC428F PC PO480A Pressurizer Pressure PT-429 psig PO481A Pressurizer Pressure PT-430 . psig PO482A Pressurizer Pressure PT-431 psig P0483A Pressurizer Pressure PT-449 psig T0406A RCLA Cold Leg Temp Deg F T0426A RCLB Cold Leg Temp Deg F T0497A RC average AT T405S Deg F T0499A RC Auctioneered Tavg T401F Deg F LO400A Steam Generator "A" Narrow Range Level LT461 PC LO401A Steam Generator "A" Narrow Range Level LT462 PC LO402A Steam Generator "A" Narrow Range Level LT463 PC LO403A Steam Generator "A" Wide Range Level LT460 PC LO420A Steam Generator "B" Narrow Range Level
! LT471 PC LO421A Steam Generator "B" Narrow Range Level LT472 PC LO422A Steam Generator _"B" Narrow Range Level LT473 PC LO423A Steam Generator "B" Wide Range Level LT470 PC P1000A Containment Pressure FT-945 psig P1001A Containment Pressure PT-947 psig P1002A Containment Pressure PT-949 psig T0481A Pressurizer Steam Temp T425 Deg F
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ATTACHMENT B POST TRIP REVIEW P-250 COMPUTER Two'Second Group Computer Address Description Units N4100P Power Range Channel N41 Power' Level PC N4200P- - Power Range Channel N42 Power Level PC-N4300P Power Range Channel N43 Power Level PC N4400P Power Range Channel N44 Power Level PC P0398A Turbine First Stage Pressure Channel 1 PT-485 . psig P0399A' . Turbine First Stage Pressure Channel 2 PT-486 psig Q0340A Unit Generator Gross MW MW T0496A RC Tref T401Y Deg F Eight Second Group Computer
. Address Description Units-N0031C' Source Range Channel 1 Count Rate DKCS
.N0032C Source Range Channel 2 Count Rate DKCS N0035C ' Intermediate Range Channel 1 Power Level MCAMP N0036C Intermediate Range Channel 2 Power Level MCAMP N0041A Power Range 1 Top Detector Flux N41A Volts N0041B Power Range 1 Bot Detector Flux N41B Volts N0042A Power Range 2 Top Detector Flux N42A Volts N0042B Power Range'2 Bot Detector Flux N42B Volts N0043A' Power Range 3 Top Detector Flux N43A Volts N00438 Power Range 3 Bot Detector-Flux N438 Volts N0044A - Power Range 4 Top Detector Flux N44A Volts N0044B Power Range 4 Bot Detector Flux N448 Volts ,
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N4100P Power Range Channel N41 Power Level PC N4200P Power Range Channel N42 Power Level PC
.N4300P . Power Range Channel N43 Power Level. PC t N4400P. Power Range Channel N44 Power Level PC P0398A Turbine First Stage Pressure Channel 1 PT-485 psig P0399A' Turbine First Stage Pressure Channel 2 i PT-486 psig i
QO340A Unit Generator Gross MW MW
! T0496A RC Tref 'T401Y Deg F l F0403C Steam Generator "A" Feedwater Flow-F466 KBH
.F0404C Steam Generator '"A" Feedwater Flow
- i. F467 KBH ,
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ATTACHMENT C SEQUENCE OF EVENTS PARAFETERS ADDITIONAL ON NEW COMPUTER Computer Address Description ICDROP RPI Rod Drop Indication INCRDROP NIS Rod Drop 1PC484A Low Vacuum Steam Dump Interlock 1PCV430 Pressurizer PORV 1PCV431C Pressurizer PORV 1PCV434 Pressurizer Safety Valve IPCV434R Pressurizer Safety Valve IPCV435 Pressurizer Safety Valve IPCV435R Pressurizer Safety Valve IRECFANA Containment Recirc Fan W1A1 Breaker 1RECFANB Containment Recirc Fan W1B1 Breaker 1RECFANC Containment Recirc Fan WlCl Breaker 1RECFAND Containment Recirc Fan W1D1 Breaker 1RHRP10A RHR Pump P10A Breaker 1RHRP10B RHR Pump P10B Breaker ITBSVACL Turbine Stop Valve "A" Close l ITBSVBCL Turbine Stop Valve "B" Close i
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