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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:VENDOR/MANUFACTURER TO NRC
MONTHYEARML20042H0101990-04-0404 April 1990 Forwards Special Instructions Forwarded to Utils Re X-SAM Sys for Nuclear safety-related Cranes.Instructions Describe Actions for Making Critical Lifts Before Retrofit of Slide Device ML20196E1621988-11-29029 November 1988 Requests That Proprietary Addendum 2 to Rev 1 to WCAP-10924-P, Westinghouse Large Break LOCA Best Estimate Methodology... Be Withheld from Public Disclosure (Ref 10CFR2.790) ML20206F8811988-11-14014 November 1988 Requests That WCAP-12042 Be Withheld from Public Disclosure Per 10CFR2.790 ML20147C6611987-12-16016 December 1987 Requests That Point Beach Unit 2 Evaluation for Vibration Induced Fatigue Be Withheld from Public Disclosure (Ref 10CFR2.790) ML20236T0011987-11-20020 November 1987 Requests That Proprietary Rept Entitled Point Beach Unit 2 Steam Generator Tube Fatigue Presentation Be Withheld from Public Disclosure (Ref 10CFR2.790(b)(4)).Affidavit Encl ML20237E3721987-10-16016 October 1987 Requests That Proprietary WCAP 11573, Point Beach Unit 2 Steam Generator Sleeving Rept (Mechanical Sleeves) Be Withheld from Public Disclosure,Per 10CFR2.790(b)(4). Affidavit Encl NRC-86-3181, Forwards Listed Nonproprietary Documents Re Rod Cluster Control Assembly Performance,Per Request.Topical Rept WCAP-8183, Operational Experience W/Westinghouse Cores Issued for Input to NRC Annual Fuel Performance Rept1986-11-25025 November 1986 Forwards Listed Nonproprietary Documents Re Rod Cluster Control Assembly Performance,Per Request.Topical Rept WCAP-8183, Operational Experience W/Westinghouse Cores Issued for Input to NRC Annual Fuel Performance Rept ML20116A1481985-04-16016 April 1985 Requests Proprietary Upper Plenum Injection LOCA Program Using SECY-83-472 Methodology Be Withheld (Ref 10CFR2.790) ML20095F3181984-08-0707 August 1984 Requests Proprietary Responses to NRC Questions Re Optimized Fuel Assembly Design Be Withheld (Ref 10CFR2.790).Affidavit Encl ML20073Q2921983-04-27027 April 1983 Authorizes Utilization of Encl Affidavit Supporting Request to Withhold Info Re Steam Generator Repair from Public Disclosure,Per 10CFR2.790 ML20204F8121983-04-13013 April 1983 Forwards Affidavit AW-80-27 in Support of Request to Withhold Proprietary Info Re Effect of Radiation on Insulating Matls in Westinghouse Medium Motor Per 10CFR2.790 Re Encl Util Draft ML20070U7941983-02-0404 February 1983 Memorializes 830131 Telcon Resolving Open Procedural Question Re Inadvertent Use of Proprietary Data in Responding to ASLB & Wi Environ Decade Exam ML20062A7721982-07-15015 July 1982 Submits Application for Withholding Supplemental Info to WCAP-9960, Point Beach Nuclear Plant Steam Generator Sleeving Rept. Affidavit (AW-80-53) Encl ML20054M7641982-07-0707 July 1982 Memorializes 820707 Telcon.Westinghouse Will Require Until 820722 to Provide Info to Be Released Per ASLB 820526 Memorandum & Order ML20054L5151982-06-29029 June 1982 Advises That Westinghouse Will Make Info Available Per ASLB 811221 & 0526 Memoranda & Orders.Discussions on How to Accomplish Task Underway ML20054L2451982-06-28028 June 1982 Requests Withholding Per 10CFR2.790 of Supplementary Info to Steam Generator Sleeve Rept.Pages Withheld (Ref 10CFR2.790) ML20054K4961982-06-23023 June 1982 Informs That Vendor Will Not Appeal from or File Exceptions to ASLB 820517 Memorandum & Order & to Interlocutory Orders Re Vendor Proprietary Info at This Time ML20052D8061982-05-0303 May 1982 Advises That Westinghouse Will Not File Reply Brief in Response to Wi Environ Decade Brief.Svc List Encl ML20054C5791982-04-0606 April 1982 Authorizes Utilization of Affidavit AW-80-53 to Justify Withholding of Reissued Revision 1 to Point Beach Steam Generator Sleeving Rept. ML20050B8911982-03-30030 March 1982 Forwards RA Wiesemann & Ta Christopher 820329 Affidavits Re 820225 & 0323 Testimony.Svc List Encl.Related Correspondence ML20049K1001982-03-23023 March 1982 Forwards RA Wiesemann Supplemental Prefiled Testimony. Related Correspondence ML20042A8411982-03-19019 March 1982 Informs That Westinghouse Rept, Point Beach Steam Generator Sleeving Rept,WCAP-9960,Revision 1 (Proprietary),WCAP-9960, Revision 1, Was Incorrectly Marked.Entire Rept Should Be Considered Proprietary.Corrected Version to Be Sent 820322 ML20042A9481982-03-0404 March 1982 Authorizes Utilization of Affidavits AW-80-32 & AW-80-53 to Justify Withholding of Revision 1 to Point Beach Steam Generator Sleeving Rept. ML20039D7361981-12-28028 December 1981 Confirms 811221,23 & 28 Telcons Re Due Date for Westinghouse Actions Per ASLB 811221 Memorandum & Order.Due to Unavailability of Key People,Date for Westinghouse to Respond to Order Revised Until 811231 ML20039B5381981-12-17017 December 1981 Forwards Westinghouse Reply Brief in Issue of ASLB Jurisdiction to Decline to Afford Proprietary Info Protection from Public Disclosure 1990-04-04
[Table view] |
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Westinghouse PowerSystems pisdnpennsyson,a 15230 03s5 Electric Corporation November 25, 1986 NS-NRC-86-3181 Mr. Earl J. Brown Office for AEOD U. S. Nuclear Regulatory Commission N~ W Washington D.C. 20555
Subject:
Request for Information on Westinghouse Rod Cluster Control Assembly Performance
Dear Mr. Brown:
p p In response to your request to Wiscons-In Electric Power Company for information related to the Point Beach LER on RCCA cladding wear, enclosed is documentation summarizing Westinghouse RCCA performance.
The following non-proprietary documents attached for your information are representative of technical information on RCCA wear submirted to utilities:
- 1) Summary of visual inspections of the RCCAs at the Point Beach Nuclear Plant (Unit 1, Cycle 11/12, Unit 2, Cycle 9/10).
- 2) Westinghouse customer information letter on RCCA wear.
- 3) Extracts from a brochure containing a summary of Westinghouse current design experience regarding RCCA wear.
In the past, the vehicle used to transmit information on RCCA performance has been the topical WCAP-8183 " Operational Experience with Westinghouse Cores". This is updated annually and issued for input to the NRC Annual Fuel Performance Report.
Very truly yours, ww.,
. Rahe , Jr. , Manager Nuclear Safety Department
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Attachment (1)
Sum dry of Visual Inspections of ECCAs at the Point Beach Nuclear Plant During the Cycle 9-10 refueline shutdown at Point Beach Unit 2 in June, 1983, site p.Iw 1 observed what appeared to be t lam 4 tug wear in excess of the current Westi, pause critaria on sane Rod Cluster Centrol Assablies Doms). 'Jhe K%2s in cuestion were of the spider urx:nted design with 16 redlets per IC2 and are ocupatible with the 14x14 fuel design used at Point Beach. '!he abscIber material is Silver - Indlun -
Cadmium. All of the Unit 2 KIms were examined with a pe.%, and ho w on the observed wear, Westinghouse reo:mnended rep 1=m of one KIA and axial repositioning of the,norml parked position of the run61ni:vy XI2s by 2-3 steps in order to m4M=49e additional wear. A further rer===rx5ation was to inspect the Unit 1 KI2s for wear indi-caticms during the Cycle 11/12 r=*=14=' shutdown (Unit I had appmci-nately 2 more cycles of % i.icn than Unit 2). 'Jhis inspecticri was recently ccupleted and resulted in a Westir-?+1- reconnendaticrt to
( replace 21 of the 33 KI2s with an inferred e =1m1=ted wear depth in excess of the current wear depth criteria.
WEAR CESERVATICES
'Ihe redlets were pv4 4-:4 using a 35m camera coupled to a peri-scope. An analytical nodel was developed which was used to infer the depth of wear indications fmn the observed wear scar widths as measured fmn the p /4 Mis. Quantitative measurunents of wear depth have not been pere ctned to date.
1 .
'No types of war patterns were c$setvie and are characterized as fol-laws: (1) Axially ocntinuous wear scar on the redlet of varvbg depth l t
probably caused by RIA axial stesping notien and scrams, (2) Approxi-nately ene-inch long smre marks at elevaticms coitwiding to the guide plate surfaces.
The wear indicaticms are r--- mmtly en the portion of the cladding surface facing the center or hub of the ICCA. 'Ihe attached sketch shows ,
l wear patterns which are typical of the worst d=2ve wear. Note that, I by design, scue wear is expex+ad due to contact between the RIA redlets and the guide surfaces.
l MIX 2T WEAR CRITERIA
'Ihe um.a.A limiting wear depth crittricn is related to =>- . : -r redlet Clad collapse. 'Ble allmbl* Wear depth was determined based Cn auto-clave tests (at taiperature and pressure) wherein maald, hollow tubi.w samples with similated wear scars were r . sized until collapse hw. .d. '!his test of a hollow tube did not take into account the sup-port provided by the =hr material, and therefore is conservative. -
Westinghouse roommenr%d that the Point Beach RIAs be replaced based on this critericn and on plant specific operating history.
A preliminary evaluation by Westinohouse indicated that an unreviewed safety questicm does not exist.
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, . NFD CUSTOMER INFORMATION LETTER ON RCCA WEAR
Subject:
NFD Customer Information Letter RCCA. Wear
Reference:
March, 1984 letter TO: Customer -
Dear Sir:
The referenced letter issted to you earlier this year provided a summary of visual inspections of the Rod Cluster Control Assemblies (RCCAs) at the Point Beach Nuclear Plant and recommendations for future actions. Those inspections identified the existence of wear marks on the cladding of the RCCA absorber rodlets. Some of the wear scars were estimated to exceed the existing Westinghouse maximum wear depth limit.
As a result the affected RCCAs were replaced based on our recommendation.
The Point Beach RCCAs had operated for 13 years from Cycle 1 initiation with approximately 80% availability.
New developments related to RCCA rodlet cladding wear have been completIed.
These include a revision to the wear depth limit and a new method of obtain-
- ing wear depths from photographs. I have attached an update to inform you of the status of on-going efforts by Westinghouse in the area of RCCA wear. I would be happy to answer any questions you may have on this subject.
Signed Fuel Projects Engineer o
J .
RCCA WEAR UPDATE Introduction ,,
In our last information letter on Silver-Indium-Cadmium Rod C1'uster Control Assembly (RCCA) wear, we brought to your attention in-reactor RCCA wear experience. The following sumarizes that letter.
The RCCAs at Point Beach Unit 2 (End of Cycle 9-June, 1983) and Unit 1 (end of Cycle 11-December,1983) showed significant levels of wear. Photographs of the wear scars were taken and a method was developed to estimate wear depth. Some of the RCCAs had estimated wear depths exceeding the then applicable wear depth limit and were therefore replaced. Two types of wear scars were seen : (1) Stepping Wear - an axially continuous scar on the rodlet caused by RCCA
- axial stepping motion and scrams, and (2) Fretting Wear - an approximately one-inch long score mark at elevations corresponding to guide cards (see Figure 1). Our letter on this subject also described the basis of the Westinghouse wear depth limit and presented preliminary recommendations for future plant specific action. ,
Note that it was concluded that no unreviewed safety question existed at the Point Beach units.
During the inspections it was observed that both types of wear occurred at the interface between the RCCA and the upper internals guide tube guide plates. Based on this study,as well as' previous studies, no unusual or significant wear has. occurred at the interface between the RCCA rodlet tip and the fuel assembly guide thimble tube. Therefore, the observed RCCA wear has no adverse impact on the operation or safe performance of Westinghouse fuel assemblies. .
, i:- .,
The following sections provide updated information on.further inspections, t /
new developments, and plans for hot cell work on the Point Beach RCCAs. -
l RCCA INSPECTIONS PERFORMED -
In addition to the two Point Beach (2 loop, .14x14 fuel) inspections r
described in our previous letter, Westinghouse ha> visually . inspected, , /
l RCCAs at three other reactors, two 2 loop'14x14 and one 4-loop 15x15 plant.I. ~ '
l As a result of these inspections we havehoncluded that the types of r I wear observed at other reactors are similar to those observed at Point '
i Beach. The wear did not appear as severe as that at Poirt Beach;- ' '
l however, the RCCAs were several years you'nger.
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NEW DEVELOPMENTS -
. Since the last information letter Westinghouse has completed a more exhaustive analysis in three areas described below: (1) the wear depth
_ limit; (2) the interpretation of photographs to estimate wear depths; and (3) axial repositioning of RCCAs. The following briefly describes these new approaches.
First, a new approach has been taken to establish a Westinghouse RCCA cladding wear depth limit which is used to meet the existing FSAR critiria. )
The criteria related to control rod clad wear are that the absorber material shall be isolated from the coolant and that the RCCA must be capable of moving freely into the fuel assembly. The impact of absorber clad wear on these criteria for the Ag-In-Cd RCCA was previously considered by Westinghouse by basing the allowable wear on prevention of clad collapse.
For conservatism no credit was taken for the support provided by the absorber material. Two dimensional elastic-plastic finite element analyses subsequent to the original Point Beach evaluation have demonstrated that the absorber material can prevent the clad from collapsing in the observed wear geometry.
Where the absorber material backs the cladding, the allowable wear depth has been increas'ed by several mils (4-6 mils depending on the RCCA design) while still meeting the existing FSAR criteria.
The second area of improvement is the wear scar photograph interpreta-tion method. Based on test data and a thorough review of the wear phe-nomenon a new method of estimating wear depth from wear scar width has been developed. The data are derived from out-of-pile tests performed in reactor grade water at reactor operating temperature and pressure and include measurements of wear scar width and maximum wear depth. Results of the application of the new method suggest that our original method of interpreting photographs may yield over estimates of wear depths by
. several mils at most elevations above the continuous guide area.
- Because the geometry of the aforementioned tests is not identical to the
. reactor geometry, an effort is also being made under EPRI sponsorship to obtain hot cell measurements of worn RCCAs to improve our understanding of the RCCA wear phenomenon. ,
The third area which has been addressed is axial repositioning of RCCAs.
If a significant level of wear is observed at your reactor, a redefi-l nition of the axial parking elevation of RCCAs by 2-3 steps can extend the service life of RCCAs for at least one additional cycle. The bene-fit is derived from spreading the fretting weer over more clad surface, thus reducing the maximum wear depth achieved. An evaluation shows that the RCCAs can be inserted several steps deeper than the normal parking position without adversely impacting operation or safety of the plant.
It should be noted that a Technical Specification change may be i required. If a change to parking elevations becomes necessary, Techni-cal and licensing support can be obtained through your NFD Fuel Projects Engineer.
With the increase in the wear depth limit and the application of the new i photograph interpretation method, the rejected RCCAs at Point Beach Unit I appear to be acceptable for continued use. Tnose RCCAs had oper-ated for 13 years with about 80% availability before the examination.
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ATTACHMENT 3 Westinghouse Rod .
R uluster Control .
Assembly Performance Review May 1986 S
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Introduction
. . As nuclear reactor systems log service rodlet cladding has occurred. In order beyond a decade, components not sub- to obtain a more comprehensive lect to periodic maintenance or evaluation of these observations, replacemen; may be expected to show detailed examinations of RCCA rodlets indications of aging. An ever-increasing having seen extended service were number of Westinghouse-supp!!sd PWRs performed by Westinghouse at its have now been in operation for such Research Laboratory in Churchill, Pa.,
extended periods, and some plant under a contract with the Electric Power operators have observed indications of Research Institute. This report provides reduced component integrity. Recently, a summary of the laboratory Rod Cluster Control Assemblies (RCCAs) examinations and an evaluation of the have shown signs that some wear of information obtained.
Cumulative Westinghouse RCCAs in Domestic Service 2800 2600 2400 2200 2000 -
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Background -
As active components in reactors, con- RCCA/ Upper Internals Layout trol rods are expected to have a lifetime limit in all light water reactors. It is -
Important to monitor and evaluate the operational experience of control rods in order to assess component performance and provide an updated basis for predic- Spider assembly-ting service lifetime. Wisconsin Electric l}
Power Company's Point Beach Nuclear i i hg Units 1 and 2 are among the leaders in RCCA rodlet j l accumulated service life for Westinghouse- \ ,
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supplied PWRs. After years of successful operation, Wisconsin Electric Card 8 Power replaced the RCCAs in Point Card 7 Beach and examinations were performed on the RCCAs taken out of service. Card 6 Westinghouse, along with Wisconsin Electric Power and EPRI, evaluated Card 5 O Guide tube selected RCCAs from Point Beach Unit 1 Card 4 I- l )] surfaces in order to determine a basis for Card 3 predicting operationallifetime reflecting experience in representative plant 2R operation. A campaign initiated in Card 1 October 1983 to investigate wear pheno- -
mena provided visual and photographic Continuous guide region ' ' "
information on 32 RCCAs from Point Beach Unit 1. - -
p T of p op One RCCA from Point Beach Unit I was fuel assembly selected for detailed examination and testing. Three rodlets were selected from this RCCA for sectioning and detailed testing in the Westinghouse R&D Hot Cell Facility in Churchill, Pa.
Ten segments of rodlets were taken for testing. Selection of the specific RCCA and rodiets was based on predeter-mined criteria including RCCA service history, visual observations of wear, and cracks in rodlet cladding.
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Description of Indications Examined ,
RCCA Performance are the result of control rod withdrawal i
Three potentist mechanical performance and insertion movements.
effects on RCCA cladding were studied. 3. Cracking l 1. Fretting Wear Short hairline cracks at the lower extrem-
- Wear spots at specific locations along Ity of the cladding were observed on the length of RCCA rodlets were noted. some rodlets. The cracks extended These locations corresponded to loca- axially for approximately four inches, tions of the control rod guide cards in and penetrated through the cladding.
i the upper reactor internals when the RCCA would be fully withdrawn from :
the core (l.a. the parked position). 'T There are eight guide cards in each '
control rod guide tube, which function to provide lateral support for individual RCCA rodlets when the RCCA is .:
withdrawn. There is also a longer ' "#
continuous guide directly above the upper core plate.
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I Fretting '
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i Service History The indications were observed after I approximately eleven years of "in plant .
- operation". Measurable wear is consis-tent with expected effects following 1
, service of such extended duration.
Wear was originally expected to be life limiting for Westinghouse RCCAs, and a
- 2. Sliding Wear life of approximately fifteen years had Wear indications appearing as scratches been predicted based on sliding wear in oriented axially on the surface of the conjunction with anticipated plant load rodlets were noted. These scratches change operations. [
correspond to interaction between RCCA rodlets and the continuous guide, and j
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Laboratory and Engineering Analysis A. Fretting Weer stepping of the RCCA during load Fretting wear is a result of vibrational c ange maneuvers.
contact between RCCA rodlets and the
. guide cards which provide lateral sup- Wear due to RCCA tripping occurs along port for the rodlets when RCCAs are the length of the cladding and, for the rodlets examined, averaged less than withdrawn from the core. Vibration is hydraulically induced by flow of the ten percent of the cladding thickness.
reactor coolant, and is therefore a Wear due to stepping is most pronounced continuous process when the reactor at the lowest portion of the control coolant pumps are in operation. In rodlet from sliding in the continuous general, wear scars resulting from guide region. The combined tripping and stepping wear at the worst location fretting were found to be approximately one-half Inch in length and included measured less than 20% of the cladding approximately one-third of the thickness. RCCA contact with the circumference of the rodlet. Depths of continuous guide causes most of the clad penetration observed varied in sliding wear, and sliding wear between approximate correspondence with the the RCCA and guide cards is minimal, amount of time the RCCAs were in the parked position and location. The C. Cracking degree of wear varied at different guide card elevations. The worst measured Several RCCA rodlets experienced hair-i depth of clad penetration from fretting line cracks in the cladding at the lower I
wear was approximately 65% of nominal extremity of the rodlet. The cracks were cladding thickness. Since there is a typically about four inches long and ,
large margin in load-bearing capability axially oriented. No circumferential for RCCA cladding, the immediate component of cracking was found. The concern if breach were to occur would lower extremity of the rodlet typicaliy l be deposition of radioactive isotopes experiences the highest fluence.
Into the reactor coolant.
Destructive examination of cracked Based on the history of this particular rodlets was undertaken to identify the RCCA, a wear rate of approximately 10% probable cause. Both metallurgical of cladding thickness per 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> analysis and dimensional examinations reactor operation was calculated. of the absorber and cladding were performed.
e exa Ina ns s we ence of B Slidin9 Wear Irradiation assisted intergranular The longitudinal scratches observed cracking and stresses due to interaction I
were determined to be the result of between absorber and cladding. The interaction between rodlets and the cladding material was type 304 stainless continuous guide located near the upper steel and the absorber was composed core plate. The wear is caused by slid. of silver (Ag), indium (in), and cadmium l ing movement and has two compon- (Cd) In proportions of 80,15, and 5 e'its. One component is attributed to percent, respectively. Negligible i RCCA trips and the other is due to deterioration of the absorber occurred due to the presence of the crack.
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a Scanning Electron Microscope (SEM) Corrosion tests were also performed to l results confirmed the Intergranular determine the extent of Intergranular nature of the cracks. Cladding tensile attack. The results provided evidence of i tests were also performed to correlate irradiation-enhanced segregation of mechanical properties with the SEM impurities at grain bounderles of the data. These tests provided evidence of stainless steel cladding.
a reduction in cladding ductility i consistent with high neutron fluence Dimensional and density measurements l near the tips. confirmed that irradiation-induced swelling of the absorber was the principal cause of tensile stress in the
- The tensile tests were performed in an cladding, which resulted in cracking inert environment so as to avoid possi- after substantial irradiation.
ble effects due to corrosive agents.
Fracture surfaces of both the tensile test samples and the cracked cladding were intergranular and were consistent with other irradiation-induced embrittlement and stress corrosion cracking experience. Based on a
, number of EPRI programs, such fracture t
characteristics are considered to be associated with migration of sliicon and phosphorus impurities.
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