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MONTHYEARML20041F9931982-03-12012 March 1982 Opposes Licensee 820127 & 0305 Request to Extend Interval Between Steam Generator Insps & to Rely on full-cycle Insps.Ginna Incident Impels Necessity to Require Immediate Shutdown & Insp Project stage: Request ML20063B8531982-08-20020 August 1982 Requests NRC Approval of Encl Inservice Insp Relief Requests for Second 10-yr Insp.Approval of Tech Spec Change Request 42 Necessary for Adoption of ASME Section XI as Governing Document for Inservice Insp Project stage: Request ML20070R2861983-01-19019 January 1983 Forwards Inservice Insp Relief Requests for Second 10-yr Insp.Nrc Acceptance of ASME Section XI Relief Requests in Attachments 2 & 3 to Enable Effective Implementation of Inservice Insp Plan Requested Project stage: Request ML20080R8541984-02-22022 February 1984 Provides Tabulated Listing of Plant First 10-yr Inservice Insp Interval Exams & Applicable Relief Requests Project stage: Request 1982-08-20
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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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l O " POWER COMPANY 231 W. MICHIGAN, P.O. E ;X 2046, MILWAUL.l, WI 53201 January 19, 1983 Mr.-H. R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Mr. R. A. Clark, Chief Operating Reactors Branch 3 Gentlemen:
DOCKET NOS. 50-266 AND 50-301 INSERVICE INSPECTION PLAN POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 As required by 10 CFR 50.55a(g) (5) (iii) , Wisconsin Electric Power Company (Licensee) is providing herewith the inservice inspection relief requests for the second ten-year inspection for Point Beach Nuclear Plant, Unit 2. In a letter dated October 6, 1981, Licensee notified the NRC of its plans to develop the Unit 2 long-term inservice inspection plan including the applicable code editions and addenda to be used for this tipdate. In Mr. Colburn's letter dated October 21, 1981, the NRC staff indicated that these proposed plans were acceptable.
- The Unit 2 inservice inspection plan has subsequently been developed i
and approved by the Licensee. As is-the case with the Unit 1 plan, l the detailed Unit 2 long-term plan for the second ten-year inservice
, inspection interval is'available for review at the Point Beach facility.
! As you know, due to the vintage of Point Beach Nuclear l . Plant, Unit 2 (the unit went into commercial service on October 1, 1972), the piping systems were designed prior to the concepts of l Code Class 1, 2, and 3 and Safety Class 1, 2, and 3 systems now l
.being used industry wide for newer plants. The pressure vessels at Point Beach Nuclear Plant, which are within the scope of the Unit 2 inservice inspection plan, were~ designed and constructed to Class A, B, or C standards of ASME Section III. The piping systems we"e designed and constructed in accordance with USAS B31.1.
The designY criteria utilized for Point Beach Nuclear Plant, Unit 2 wer$' derived from the then proposed general design criteria of 10 CFR 'iO as clarified in the Point Beach Nuclear Plant Final Safety Ana' lysis Report.
Is 8301270395 830119 f
PDR ADOCK 05000266 f,#
i, Q PDR l3
3 , .
+.
Mr. H. R. Denton - January 19, 1983
.s 1
4 As stated in. Paragraphs 5.2.1 and 5.2.2 of Attachment 1, 4 , - Regulatcry. Guide . l.26 and ANSI Standard - N18.2_ wer: _ utilized as guidance'to determine Code Class 1,:2, and 3 boundaries;for
-inservice.inspectidn purposes. .Since Point Beach _ Nuclear Plant, Unit 2 :was not originally designed to tlut specific requirements of-Regulatory. Guide 1.26.and ANSI Standard N18.2, the. Code Class boundaries established for inservice inspectica purposes meet the
- . intent.of theJcriteria established in these standards. The Point
- . Beach Nuclear Plant Final Safety Analysis Report was the basis upon which' determination of the-Code Class boundaries
- was performed since this documentiis the foundation for the design and safety analysis of tho' unit. Attachment.1 provides a description of the, Unit 2 inservica inspection second ten-year' interval plan.
t Unit 2 relie'f re'questsifor specific ASME Section XI ,
I code requirements are:provided"in Attachment 2. As the second..
-inspection interval progresses and other problem areas are identi-fled, additional' relief req'uests-may need to be submitted.
~
Attachment 3 contains relief requests 6, 7, 8, and 9 for Point Beach Unit 1. .These relief requests are an addition to i' _those submitted.in our. August 20,-1982 letter. These additional requests were determined to be necessary_either as-the. result of fully-implementing the-Unit-1 plan during the 1982 refueling.
outage,or were identified during the finalization of the Unit 2
<- long-term plan.
e As discussed via telephone with Mr. Tim Colburn and
' -other members of your staff on September- 14, 1982, the resolution of' inconsistencies ~concerning two of the.10 CFR Section 50.55a reliefs granted by,Mr. Clark's letter of' August 31,~1982~ has not ,
yet been received. This. resolution is a-necessary part of
. Wisconsin Electric's attempt to satisfactorily cloce out the
' Point. Beach Unit'2 first. ten-year inservice inspection interval 1
requirements.- Your prompt resolution of these inconsistencies
- ' is requested.
t i We hereby request NRC's. acceptance of the ASME Section i XI relief requests presented in_ Attachments 2 and 3 so that the inservice inspection plan can~be effectively implemented. It is the' intention of Wisconsin Electric to perform examinations in
- accordance with this-approved plan and it will be the basis for.
- ' compliance with. regulatory requirements.
Very truly yours, gsu'f :
Vice Presid t- clear Power
[ C. W. Fay-4 Attachments
! ' Copy to NRC . Resident Inspector s
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ATTACHMENT 1 i' POINT BEACH NUCLFJR PLANT
- l UNIT 2 ISI LONG-TERM PLAN (LTP)
DEVELOPMENT BASIS 4
1.0' PURPOSE -
'Ib present a sununary of the Unit .2 ISI LTP.
2.0 DESCRIPTION
OF LTP 2.1 The LTP provides a listing of all areas required to be examined under
. Articles IWB, IWC, IWD, and IWF of ASMS Section XI.
2.2 .The LTP provides a planning schedule of when a given examination will 3
-be performed during the second 10-year interval.
2.3 The LTP will pr. ovide an updated status of all required ISI.
> 3.0 APPLICABLE CODE EDITION / ADDENDA 3.l' All ISI work will-be conducted in accordance with the 1977/Sununer 1979 Edition / Addenda of ASME Section XI with the exception of Class 1
- and 2 piping welds.
'3.2 . The extent of examination (i.e. , frequency, number of welds) of Class 1 and 2 pipirig welds will be determined in accordance with .the 1974/ Summer 1975 Edition / Addenda of ASME Section XI.
- 3. 3' -The examination methods, acceptance standards, and other aspects of
'the e - ination of Class 1 and 2 piping welds will be in accordance with the 1977/ Summer 1979 Edition / Addenda of ASME Section XI.
I- 24.0- BASIS FOR CODE SELECTION,
-4.1 10 CFR 50-4.1.1 In accordance with 10 CFR 50.55a(g) (4) (ii) , inservice examin-L- ation of components and system pressure tests conducted
(' during the second and successive 10-year intervals shall comply with the requirements of the latest edition of the Code incorporated by reference in 10 CFR 50.55a(b) (2) 12 months prior to the start of the 120-month inspection interval.
4.1.2~ In accordance with 10 CFR 50.55a(b) (2) (ii) and (b) (2) (iv) ,
WE chose to determine the extent of examination of Class 1
'and 2 piping welds in accordance with the 1974/ Summer 1975 Edition / Addenda of ASME Section XI. -
l
'1 i
, z
=^' Attac % nt 1
-Page 2 5
i 4.1.3 The if74/ Summer 1975 Ed. tion / Addenda of ASME Section XI must be utilized to. determine the extent of examination of Class 2 piping welds in the residual heat removal and the tsfety injection systems.
4.2 In the' Fay to Denton letter dated 10/06/81, WE informed the NRC that
'the Units 1 and 2 LTP's would be written to the Codes referenced in 4 3.a, 3.b, and 3.c.
4.3 In the Colburn to Burstein letter, dated 10/21/81, the NRC accepted the WE plans stated in the 10/06/81 letter.
5.0 CODE CLASS BOUNDARIES 5.1 For the purposes of 'ISI, determination of Code class boundaries on
-systems and equipment designated as nuclear safety related and impor-
~
tant to safety were established using the following selection criteria.
5.2 Selection Criteria References 4
5.2.1 Regulatory Guide 1.26 (Revision 2) , " Quality Group Classifi-cations and Standarde for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants."
5.2.2 ANSI Standard N18.2 (1973) and N18.2a (1975), " Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants."
5.2.3 PBNP Final Safety Analysis Report
- a. Section 4 - Reactor Coolant System
- b. Section 6 - Engineered Safety Features
- c. Section 9 - Auxiliary and Emergency Systems
- d. Section 11 - Waste Disposal and Radiation Protection
- e. Section 14 - Safety Analysis
- f. Appendix A - Seismic Design for Structures and Equipment 5.3 Exemptions 5.3.1 Class 1
- a. Atticle IWB-1220 of the 1977/ Summer 1979 Edition / Addenda of ASME Section XI for all components except piping welds.
- b. Article IWB-1220 of the 1974/Sununer 1975 Edition / Addenda of ASME Section XI for piping welds.
-- _ m , _ , _ - . %
6 Q Attachment 1 Page,3 5.3.2 class 2
- a. Article IWC-1220 of the 1977/Sununer 1979 Edition / Addenda of ASME Section'XI for all components except piping,
- b. Article IWC-1220 of the'1974/ Summer 1975 Edition / Addenda of ASME Section XI for piping.
- c. Due to the nonacceptance by the IGC of exemption criteria IWC-1220(c) of the 1974/Stanmer 1975 Edition / Addenda of ASME Section XI at other plants, this exemption was not
! taken. This paragraph permits exemption of components which perform an emergency core cocling function, provided the control of the chemistry of the contained fluid is verified by periodic sampling and testing.
- d. The suction piping for the safety injection, containment spray, and residual heat removal pumps from the refueling water storage tank was exempted in accordance with the clarification provided by IWC-1220(a) of the 1977/ Summer 1979 Edition / Addenda of ASME Section XI. Piping from the boric acid storage tank to the safety injection pump suction was also exempted in accordance with IWC-1220(a) .
6.0 INSPECTION REQUIREMENTS 6.1 For all components, except for the extGnt of examination of Class 1 and 2 piping welds, the following tables from the 1977/ Summer 1979 Edition / Addenda of ASME Section XI was utilized.
6.1.1 Class 1 - Table IWB-2500-1 6 .1. 2 Class 2 - Table IWC-2500-1 6.1.3 Class 3 - Table IWD-2500-1 6.1.4 Class 1, 2, and 3 supports - Table IWF-2500-1 6.2 For the extent of examination of Class 1 and 2 piping welds, the following tables from the 1974/ Summer 1975 Edition / Addenda of ASME Section XI will be utilized.
6.2.1 Class 1 - Table IWB-2500, Category BJ, and Table IWB-2600.
6.2.2 Class 2 - Table IWC-2520, Categories CF and CG, and Table IWC-2600.
6.3 The reactor coolant pump flywheel will be examined as specified in our 02/23/E2 letter from C. W. Fay to H. R. Denton. The inspection method and frequ.-ncy meet the intent of Regulatory Guide 1.14.
- 'W'- -
-% m- -
.s Attachment 1
, Pago 4 7.0 RELIEF REQUESTS .
Those items for which Code .elief is requested can be found in Attachment 2.
8.0 CODE INTERPRETATIONS 8.1 Class 3 piping supports on piping greater than 4" diameter require examinations under Table IWF-2500-1. No examination will be performed on Class 3 piping supports on piping less than or equal to 4" diameter.
8.2 The regenerative heit exchanger. consists of three identical shells interconnected by piping as shown in Figure A-7. Each of the three identical shells are being considered similar in accordance with Note 3 of Table IWC-?500-1, Category CA. This applies to the Class 2 side only.
8.3 According to IWC-5210(a) (1), a system pressure test conducted during a system functional test of those Class 2 systems not required to operate during normal reactor operation is required every 40 months.
All Class 2 systems, according to IWC-5210(a) (2) , are required to have a 10-year hydrostatic pressure test.
9.0 ITEMS NOT COVERED BY THE LONG-TERM ISI PLAN 9.1 Snubber testing.is covered under Technical Specification 15.4.13.
l 9.2 Pump and valve testing, as specified under subsections IWP and IWV of t ASME Section XI, are handled under a separate pump and valve testing
- program. This program was sul
- snitted to the NRC on 02/10/81. The
! pump and valve program was written in accordance with the 1977/ Summer 1978 Edition / Addenda of ASME Section XI.
9.3 Steam generator tubing inspection is governed by Technical Specification 15.4.2. A versus ASME Section XI, as allowed by 10 CFR 50.53a(b) (2) (iii) .
)
. ff*- - s'
, n-I ATTACHMENT 2 12 FIT 2~ RELIEF REQUESTS
, , ' RELIEF REQUEST NO. DESCRIPTION.
Reactor vessel-interior exam.
.RR-2-1 .
'8R-2 -
Safety injection nozzle to safe end weld.'.
RR-2-3 Regenerative heat exchanger integrally welded supports.
~
RR-2-4 Reactor coolant pump integrally welded supports.
- d. RR-2-5
~
Piping to penetration welds on the auxiliary coolant and safety injection systems.
RR-2-6 Reactor vessel integrally welded support attachment.
RR-2 -7 Class 3 integrally welded attachments.
RR-2-8 Reactor. coolant system' safety injection piping.
RR-2-9 Safety injection-reducer to safe-end weld.
i
~ DRAWINGS.
i FIGURE NO. DESCRIPTION ,
! 1 Regenerative heat exchanger' integrally welded attachment.
2 Containment wall penetration detaf1.
3 Reactor vessel internals.
A-1 Reactor pressure vessel.
A-7 Regenerative heat' exchanger.
A-8 Reactor coolant pump supports.
11 B-10 AC-10-RHR-2001 B-14 AC-10-RHR-2006~
B SIS 6-SI-2008 A-28 .
RC-6-SI-2001 and RC-4-SI-2001 A-32 RC-6-SI-2002 and RC-4-SI-2002 f
-m,- ---m , - - , , . , . s .f,., , ..E,. vr' _ + r - ,m... . - .. ,__. ,. _ _ _
m AttOchment 2 Page 2
,_ RR-2 '
COMPONENT
-Reactor pressure vessel.
EXAM AREA' Reactor vessel interior surfaces.
ISOMETRIC OR COMPONENT DRAWING None.
ASME SECTION XI CATEGORY
- BN1 ASME SECTION XI ITEM NUMBER B13.10 ,
ASME SECTION XI EXAMINATION REQUIREMENT A visual examination (VT-3) is required every 3 years of the accessible areas of the vessel interior surfaces'during a normal refueling outage.
ALTERNATIVE EXAMINATION A visual examination (VT-3) of the reactor vessel interior will be performed when the core barrel is removed but not at a frequency greater than that specified in the Code. The core barrel will not be removed
, specifically for this examination.
< REASON FOR LIMITATION As can be seen from Figure .3, only a small portion of the reactor vessel interior surfaces are accessible with the core barrel in place. There is approximately 10" of the vessel interior surface accessible from the top of the core barrel flange to the reactor vessel flange during a normal refueling outage. A meaningful examination cannot be performed unless the
- - core barrel is removed. Removal of the core barrel requires a complete defueling of the reactor and significant ALARA impacts including exposure and contamination problems.
9-6 o.'
RR-2-2 COMPONENT IDENTIFICATION Reactor pressure vessel safety injection nozzles (2 nozzles).
EXAMINATION AREA Safe end-to-nozzle weld
- l. RC-4-SI-2001-26
- 2. RC-4-SI-2002-29
~ ISOMETRIC OR COMPONENT DRAWING A-1 ASME SECTION XI CATEGORY BF ASME SECTION XI ITEM NUMBER B5.10 4
ASME SECTION XI EXAMINATION REQUIREMENT A surface and volumetric exam of each nozzle-to-safe end weld every 10 years.
ALTERNATIVE EXAMINATION A volumetric examination of each nozzle-to-safe end weld every 10 years.
This examination is performed from the ID of the nozzle using mechanized equipment.
AE.ASON FOR LIMITATION A surface examination is not possible due to the inaccessibility of this area. These welds are located between the vessel and biological shield wall.
', katochment 2 Page 4 RR-2-3 COMPONENT IDENTIFICATION Regenerative heat exchanger.
-EXAMINA. TON AREA Integrally welded supports (3) . .
ISOMETRIC OR COMPONENT DRAWING A-7 ASME SECTION XI CATEGORY BH ASME RECTION XI ITEM NUMBER B8.40
'ASME SECTION XI EXAMINATION REQUIREMENT A volumetric examination of each integrally welded attachment every 10 years,.
ALTERNATIVE EXAMINATION A visual examination (VT-3) of each integrally welded attachment every 10 years.
REASON F0R LIMITATION As can be seen in Figure 1, the integrally welded attachment is attached to the regenerative heat exchanger shell by an intermittent fillet weld.
These welds do not lend themselves to a meaningful ultrasonic, double wall radiographic, or alternative surface examination.
e ,' 9-
'^
Attrchment 2 Page 5 RR-2-4
, COMPONENT
' Reactor coolant pumps A and B.
EXAMINATION AREA Integrally welded support lugs (3).
- ISOMETRIC OR CCMPONENT DRAWING A-8 ASME SECTION XI CATECORY BKl ASME SECTION XI~ ITEM NUMBER B10.20
- ASME SECTION XI EXAMINATION REQUIREMENT A volumetric examination of each integrally welded lug.is required every 10 years.
ALTERNATIVE EXAMINATION A visual examination (VT-3) of each integrally welded attachment will be performed every 10 years.
REASON FOR LIMITATION
' Ultrasonic ar.3 liquid penetrant examination of this area is not feasible due to the attenuative properties and surface roughness of the cast stainless steel.
e e -
a 4 4 4 f-
.,- Attichment 2 Page 6 RR-2-5 .
COMPONENT IDENTIFICATION Auxiliary coolant and safety injection systems.
_ EXAMINATION AREA
- Piping-to-penetration cap welds.
- 1. AC-10-RHR-2001-24
- 2. AC-10-RHR-2006-18
- 3. SIS-6-SI-2008-19 ISOMETRIC OR COMPONENT DRAWING B-10,.B-14, B-29 ASME SECTION XI CATEGORY =
CF ASME SECTION XI ITEM NUMBER C5.ll ASKE SECTION XI EXAMINATION REQUIREMENT A surface examination of these welds is required.
ALTERNATIVE EXAMINATION
.No exam practical.
REASON FOR LIMITATION As shown in Figure 2, the welds of concern are totally enclosed within
'the containment penetration and are inaccessible.
4
. Attachment a Page 7 RR-2-6 COMPONENT Reactor pressure vessel.
EXAM AREA Reactor vessel integrally welded support attachment (2 attachments).
ISOMETRIC OR COMPONENT DRAWING None.
ASME SECTION XI CATEGORY
- BH ASME SECTION XI ITEM NO.
B8.10.
ASME SECTION XI EXAMINATION REQUIREMENT A volumetric or surface exam of each integrally welded attachment is required every ten years. According to Table IWB-2412-1, inspection Program B, the examination should be allocated as follows.
Inspection Period, Minimum Exams Maximum Exams Years Into Completed % Completed %
Interval 3 16 34 7 50 67 10 100 100 For two integrally welded attachments, one should be examined in the
' first or second period and one should be examined in the third period.
ALTERNATIVE EXAMINATION 4
A volumtric examination of both integrally welded attaclunents will be done once every ten years when the core barrel is removed from the reactor vessel.
s o
Attachment 2 Page 8 REASON FOR LIMITATION The vessel to integral attachment weld is examined from the inside surface of the reactor. vessel using a remote operated, mechanized, ultrasonic test
. device. To perform this exam,-the core barrel must be removed to permit access to the vessel wall'for ultrasonic examination. There is not access to the supports from the outside since they are located between the vessel and biological shield wall.
J t
Attachment 2 Page 9
, RR-2-7 COMPONENT Clans 3 integrally welded attachments.
EXAM AREA
' Component supports.
ISOMETRIC OR COMPONENT DRAWING None.
ASME SECTION XI CATECORY ,
DA, DB, DC ASME SECTION XI ITEM NUMER Dl.20 through Dl.60 D2.20 through D2.60 D3.20 through. D3.60 ASME SECTION XI EXAMINATION REQUIREMENT A visual examination (VT-3) is requir9d each inspection period of the integral. attachment including those attachments for c:mponents exceeding 4" nominal pipe size, whose strnettital integrity is relied upon to with-stand design loads when the system function is required.
ALTERNATE EXAMINATION l-A visual examination (VT-3) will be performed each inspection interval of the integral attachment including those attachments for components exceeding 4" nominal pipe size, whose structural integrity is relied upon to withstand design loads when the system function is required.
I REASON FOR LIMITATION
!- In the ASME Code Section XI 1977/S79 edition / addenda, the frequency of Class 3 support (section IWF) exams and their integrally welded attachments (section IWD) exams are as follows:
Exam Area Table _ Category _ Frequency Supports IWF-2500-1 FA, FB, FC Once per interva)
Integrally welded attachments IWD-2500-1 DA, DB, DC Once per period l
Attachment 2 Page 10
- The ext.mination requirements of Class 3 supports.and their integrally welded attachments are not consistent in Sections IWD and IWF of the 1977/S79 edition / addenda of ASME Section XI. The 1980/W80 edition / addenda revised the examination frequency requirements of the integrally welded attachments to once every interval. Based on the revision in the 1980/W80 edition /
addenda of ASME Section XI, and the inconsistency between sections IWD and IWF in the 1977/S79 edition / addenda of ASME Section XI, PBNP requests to examine Class 3 supports and their integrally wolded attachments at a frequency of once per interval.
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Attachment 2 Page 11 RR-2-8 COMPONENT IIANTICICATION Reactor coolant systems safety injection piping.
EXAMINATION AREA
- 1. 'RC-6-SI-2001-23A
- 2. RC-6-SI-2001-24
- 3. RC-6-SI-2002-26A
- 4. 'RC-6-SI-2002-27 I_SOMETRIC OR COMPONENT DRAWING
- 1. A-28
! 2. A-32 ASME SECTION XI CATEGORY BJ ASME SECTION XI ITEM NUMBER B9.11 ASME SECTION XI EXAMINATION REQUIREMENT A surface and volumetric exam of 25% of the circumferential joints each inspection interval (10 years) to accomplish 100s of the circumferential l joints over the service life of the plant.
ALTERNA2IVE EXAMINATION 1
l A surface and volunetric exam of 25% of the accessible circumferential joints each inspection interval (10 years) to accomplish 100% of the accessible circumferential joints over the service life of the plant will be performed.
The above examination areas are not accessible, leaving the number of accessible ASME Section XI, Category B-J circunferential welds at 92% of the
- - total number of RC/AC-6-SI-2001 piping circumferential welds and 93% of the total number of RC/AC-6-SI-2002 piping circumferential welds.
REASON FOR LIMITATION
!~ A surface and volumetric exam from the exter.'ar of the piping is not possible due to the inaccessibility of this area. These welds are located between the reactor pressure vessel and biological shield wall. A volumetric examination l
from the . interior of the piping by mechanized equipment is also not possible due'to the pipe size and configuration.
Attachment 2 Page 12 RR-2-9 COMPONENT IDENTIFICATION Safety injection reducer to safe-end weld.
EXAMINATION ~ AREA
- 1. RC-4-SI-2001-25
- 2. RC-4-SI-2002-28 ISOMETRIC OR COMPONENT' DRAWING
- 1. A-28
- 2. A-32 ASME SECfION XI' CATEGORY BJ A3ME SECTION XI ITEM NUMBER 39.11 ASME SECTION XI EXAMINATION REQUIREMENT A surface and volumetric examination of 25% of the circumferentill joints each inspection interval (10 years).
ALTERNATIVE EXAMINATION A volumetric examination of each reducer to safe-end weld every 10 years when the associated nozzle to safe-end weld is performed. This examination is performed from the inside diameter of the safe-end using mechanized equipment.
REASON FOR LIMITATION A surface examination is not possible due to the inaccessibility of this area. These welris are located between the vessel and biological shield wall.
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ATTACHMENT 3 UNIT 1 RELIEF REQUESTS RELIEF REQUEST NO. DESCRIPTION RR-1-6 Reactor vessel integrally welded attachments.
RR-1-7 Class 3 integrally welded attaclanents.
RR-1-8 Reactor coolant syst m safety injection piping.
RR-1-9 Safety injection reducer to safe-end weld.
DRAWINGS FIGURE NO. DESCRIPTION A-26 RC-4-SI-1001, RC-6-SI-1001, AC-6-SI-1001 and SIS-2-SI-1001 A-29 RC-6-SI-1002, AC-6-SI-1002, RC-4-SI-1002, and SIS-2-SI-1003 e
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Attachment 3 Page 2 RR-1-6 COMPONENT Reactor pressure vessel.
t EXAM AREA Reactor vessel integrally welded support attachment (2 attachments).
ISOMETRIC OR COMPONENT' DRAWING .
None.-
ASME SECTION~XI CATEGORY BH
- ASME SECTION XI ITEM NO.
Bs.10 ASME SECTION XI' EXAMINATION' REQUIREMENT A' volumetric or surface exam of each integrally welded attachment is required every ten years. According to Table IWB-2412-1, Inspection Program B, the examination should be allocated as follows.
Inspection Period, Minimum Exams Maximum Exams Years into Interval Completed % Completed %
3 16 34 7 50 67 i 10 100 100 i
For two integrally welded attachments, one should be examined in the first or second period md one should be examined in the third period.
ALTERNATIVE EXAMINATION A vi.lumetric examination.of both-integrally welded attachments will be done once every ten years when the core barrel is removed from the reactor vessel.
REASON FOR L*.MITATION The vessel to integral attachment weld is examined from the inside surface of the reacter vessel using a remoto operated, mechaniszed, ultrasonic test device.
To perform this exam the core barrel must be removed to permit' access to the vessel wall for ultrasonic examination. There is no access to the supports from the outside since they are located between the vessel and biological shield wall.
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Attachment 3 Page 3 RR-1-7' COMPONENT Class 3 integrally welded attachments.
EXAM AREA Component sigports.
ISOMETRIC OR COMPONENT DRAWING None.
ASME Sm'"fION XI CATEGORY DA, DB, DC ASME SECTION'XI ITEM ~ NUMBER Dl.20 through D1.60 D2.20 through-D2.60 D3.20 through D3.60
~ASME SECTION XI~ EXAMINATION REQUIREMENT A visual exarination (VT-3) is required each inspection period of the integral attachmentrincluding those attachments for components exceeding 4" nominal-pipe size, whose structural integrity is relied upon to with-stand design loads when the system function is required.
ALTERNTl"E EXAMINATION A visual examination (VT-3) will be performed each inspection interval of the integral attachment, including those attachments for components exceeding 4" nominal pipe size whose structural integrity is relied upon to withstand
. design loads when the. system function is required.
- REASON FOR LIMITATION In the ASME Code Section XI 1977/S79 edition / addenda, the frequency of Class 3 support (Sectic,n IWF) exams and their integrally welded attachments (Section IWD) exams are as follows:
g Area- Table Category Frequency Supports IWF-2500-1 FA, FB, FC Once per interval Integrally welded IWD-2500-1 DA, DB, DC Once per period
'Attachsents
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Attachment 3 Page 4 The examination requirements of Class 3 sugyorts and their integrally welded attachments are not consistent in Sections.IWD and IWF of the 1977/S79 edition /
addenda of ASME Section XI. The 1980/W80 edition / addenda revised the examination frequency requirements of the integrally welded attachments to once every interval. Based en the revision in the 1980/W80 edition / addenda of ASME Section XI, and the inconsistency between Sections IWD and IWF in the 1977/S79 edition / addenda of ASME-Section XI, PBNP requests to examine Class 3 supports and their integrally welded attachmente at a frequency of once per interval.
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4 Attachment 3 Page-5 RR-1 -8 COMPONENT IDENTIFICATION-Reactor coolant system safety injaction piping.-
EXAMINATION AREA
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l'. RC-6-SI-1001-28
^ 2. RC-6-SI-1001-29
- 3. RC-6-SI-1001-30
- 4. RC-6-SI-1001 31 t
- 5. RC-6-CI-1002-15
- 6.~ RC-6-SI-1002-16
- 7. RC-6-SI-1002-17 ISOMETRIC OR COMPONENT DRAWING
'l. A-26'
- 2. A-29 ASME SECTION XI CATEGORY BJ-ASME SECTION XI' ITEM' NUMBER i
B9.ll ASME SECTION XI EXAMINATION REQUIREMEfff A surface and volumetric . exam of 25% of the circumferential joints each inspection interval (10 years) = to accomplish 100% of the circumferential
. joints over the service life of the plant.
ALTERNATIVE EXAMINATION A' surface and volumetric exam of.25% of the accessible circumferential joints
- .each inspection interval (10 years) to accomplish 100% of the accessible circumferential joints over the service life of the plant will be performed.
~ The above examination areas are not accessible, leaving the number of
. accessible ASME Section XIi Category.BJ circumferential welds at 88% of the
- total number-of BC-6-SI-1001 piping circumferential' welds and 84% of the j e /g(
total number of BC-6-SI-1002 piping circumferential welds. p
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Attachment 3 ,
Page 6 REASON FOR LIMITATION A' surface and voltsnetric exam from the exterior of the piping i3 not -
possible due to the inaccessibility of this area. These welds are located
.between-the. reactor pressure vessel and biological shield wall. A volumetric _ examination from ths interior of the piping by mechanized equipment is also not possible due to the pipe size and configuration.
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.Page 7 RR-1-9 l l
COMPONENT IDENTIFICATION' i-Safety injection reducer-to safe-end weld.
EXAMINATION AREA 1.- RC-4-SI-1001-32'-
2 . -- RC-4-SI-1002-18 ISOMETRIC OR COMPONENT DRAWINGS
- 1. 7-26 3
- 2. A-29 ASME SECTIOti XI CATEGORY BJ ASME SECTION XI ITEM NUMBER
, ASME SECTION XI' EXAMINATION REQUIREMENT A surface and. volumetric examination of 25% of the circumferential joints each inspection interval ' (10 years) .
ALTERNATIVE EXAMINATION A volumetric examination of each reducer.to safe-end weld every 10 years r
when the associated nozzle to safe-end weld is performed. This examination is performed from the inside diameter of the safe-end using mechanized i equipment.
REASON FOR LIMITATION 4
A surface examination is not possible,due to '.he inaccessibility of this r -area. These welds are located between the vessel and biological shield wall.
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, NOMINAL THICKNESS: .562" .438" .344" RC/AC-6-SI-1001 CALIBRATION BLOCK: 4-PTB, 6-PTB SIS-2-SI-1001
) REFERENCE ISO: BECHTEL P-136
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