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Category:CORRESPONDENCE-LETTERS
MONTHYEARIR 05000412/19990071999-10-21021 October 1999
Refers to Special Team Insp 50-412/99-07 Conducted from 990720-29 & Forwards Nov.Two Violations Identified.First Violation Involved Failure to Implement C/A to Prevent Biofouling of Service Water System
ML20217M1591999-10-19019 October 1999
Forwards NRC Rept Number 17, Requal Tracking Rept from Operator Licensing Tracking Sys.Rept Was Used by NRC to Schedule Requalification Exam for Operators & Record Requal Pass Dates
L-99-143, Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct1999-10-11011 October 1999
Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct
L-99-152, Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections1999-10-11011 October 1999
Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections
ML20217C6741999-10-0808 October 1999
Forwards RAI Re Licensee 970128 Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions, . Response Requested within 60 Days of Receipt of Ltr
L-99-151, Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program1999-10-0707 October 1999
Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program
ML20217E0301999-10-0707 October 1999
Forwards Insp Repts 50-334/99-06 & 50-412/99-06 on 990809-13 & 990823-27.Violation Noted Involving Failure to Correctly Translate Design Change Re Pertinent Operating Logs & Plant Equipment Labeling
ML20212M2661999-09-30030 September 1999
Forwards Order Approving Transfer of Licenses for Beaver Valley from Dlc to Pennsylvania Power Co & Approving Conforming Amends in Response to 990505 Application
ML20212K8071999-09-30030 September 1999
Informs That on 990916,NRC Staff Completed mid-cycle Plant Performance Review (PPR) of Facility.Staff Conducted Reviews of All Operating NPPs to Integrate Performance Info & to Plan for Insp Activities at Facility
ML20216J9621999-09-30030 September 1999
Forwards Insp Repts 50-334/99-05 & 50-412/99-05 on 990725-0904.Two Violations Noted & Being Treated as Ncvs.One Violation Re Failure to Follow Operation Manual Procedure Associated with Configuration Control Identified
L-99-149, Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion1999-09-28028 September 1999
Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion
L-99-148, Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 9908171999-09-24024 September 1999
Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 990817
ML20212G0601999-09-23023 September 1999
Forwards Answer of Duquesne Light Co to Petition to Waive Time Limits & Suppl Comments of Local 29, Intl Brotherhood of Electrical Workers.Copies of Answer Have Been Served to Parties & Petitioner by e-mail or Facsimile
ML20212C5521999-09-21021 September 1999
Forwards for Filing,Answer to Firstenergy Nuclear Operating Co & Pennsylvania Power Co in Opposition to Petition to Waive Time Limits & Suppl Comments of Local 29 Intl Brotherhood of Electrical Workers
L-99-144, Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-031999-09-20020 September 1999
Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-03
ML20212B3291999-09-16016 September 1999
Forwards for Filing,Petition to Waive Time Limits in 10CFR2.1305 & Supplemental Comments of Local 29,Intl Brotherhood of Electrical Workers Re Beaver Valley Power Station,Units 1 & 2
L-99-134, Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC1999-09-15015 September 1999
Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC
ML20211Q3431999-09-0808 September 1999
Informs That During 990903 Telcon Between L Briggs & T Kuhar,Arrangements Were Made for NRC to Inspect Licensed Operator Requalification Program at Plant,Unit 1.Insp Planned for Wk of 991115
ML20211Q5601999-09-0707 September 1999
Forwards Insp Rept 50-412/99-07 on 990720-29.Three Apparent Violations Noted & Being Considered for Escalated Ea. Violations Involve Failure to Implement C/As to Prevent bio- Fouling of Svc Water Sys
L-99-138, Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)1999-09-0303 September 1999
Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)
L-99-136, Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls1999-09-0202 September 1999
Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls
L-99-098, Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods1999-09-0202 September 1999
Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods
L-99-137, Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 31999-08-31031 August 1999
Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 3
L-99-022, Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl1999-08-31031 August 1999
Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl
L-99-012, Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B
L-99-037, Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change
L-99-132, Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 21999-08-26026 August 1999
Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 2
05000412/LER-1999-007, Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info1999-08-19019 August 1999
Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info
ML20211A5111999-08-18018 August 1999
Forwards Insp Repts 50-334/99-04 & 50-412/99-04 on 990613- 990724.One Violation Noted & Treated as Non-Cited Violation Involved Failure to Maintain Containment Equipment Hatch Closed During Fuel Movement
L-99-127, Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel1999-08-17017 August 1999
Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel
L-99-124, Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached1999-07-30030 July 1999
Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached
L-99-121, Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements1999-07-28028 July 1999
Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements
L-99-118, Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 20011999-07-25025 July 1999
Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 2001
L-99-120, Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations1999-07-22022 July 1999
Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations
L-99-119, Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 9901221999-07-20020 July 1999
Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 990122
L-99-113, Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by1999-07-15015 July 1999
Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by
L-99-111, Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes1999-07-15015 July 1999
Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes
L-99-112, Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl1999-07-14014 July 1999
Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl
L-99-110, Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.61999-07-14014 July 1999
Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6
ML20209G5701999-07-12012 July 1999
Discusses Closure of TACs MA0525 & MA0526 Re Response to RAI Concerning GL 92-0,Rev 1,Suppl 1, Rv Structural Integrity. Info in Rvid Revised & Released as Ver 2 as Result of Review of Response
ML20207H6621999-07-0808 July 1999
Forwards RAI Re Util 981112 Response to IPEEE Evaluations for Plant,Units 1 & 2.RAI Was Discussed During 990628 Telcon in Order to Ensure Clear Consistent Understanding by All Parties of Info Needed
L-99-105, Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves1999-07-0808 July 1999
Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves
ML20209D8191999-07-0707 July 1999
Forwards Insp Repts 50-334/99-03 & 50-412/99-03 on 990502- 0612.No Violations Noted.Program for Maintaining Occupational Exposures as Low as Reasonably Achievable (ALARA) & for Training Personnel,Generally Effective
L-99-109, Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-62301999-07-0707 July 1999
Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-6230
L-99-108, Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC1999-07-0707 July 1999
Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC
L-99-104, Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl1999-06-29029 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl
L-99-093, Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.51999-06-25025 June 1999
Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.5
L-99-102, Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl1999-06-22022 June 1999
Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl
L-99-101, Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal1999-06-22022 June 1999
Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal
L-99-062, Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages1999-06-17017 June 1999
Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages
1999-09-08
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARL-99-152, Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections1999-10-11011 October 1999
Submits Relief Request BV3-N-533-1,rev 0,requesting Use of ASME Approved Code Case N-553-1, Alternative Requirements for VT-2 Visual Examination of Class 1,2 & 3 Insulated Pressure Retaining Bolted Connections
L-99-143, Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct1999-10-11011 October 1999
Informs That Subsequent Review of Approval Ltrs & SE for Relief Requests 1TYP-2-B5.40-1 (Rev 0),BV1-RV-AUG (Rev 1), BV2-RV-AUG (Rev 1),BV3-IWA-1 (Rev 1) & BV3-IWA-2 (Rev 1) Identified Erroneous Statements Which Dl Wishes to Correct
L-99-151, Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program1999-10-0707 October 1999
Responds to NRC Re Violations Noted in Insp Rept 50-412/99-07.Corrective Actions:Condition Repts Were Written for Listed Issues So That Repts Could Be Addressed Using BVPS Corrective Action Program
L-99-149, Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion1999-09-28028 September 1999
Informs NRC That Items Identified in 990629 Response to GL 98-01 Have Been Completed.Attached Table Submitted with Has Been Updated to Reflect Completion
L-99-148, Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 9908171999-09-24024 September 1999
Notifies NRC of License Withdrawal of M Linch,License SOP-11478 IAW 10CFR50.74.M Linch Resigned from Employment at Bvps,Effective 990817
ML20212G0601999-09-23023 September 1999
Forwards Answer of Duquesne Light Co to Petition to Waive Time Limits & Suppl Comments of Local 29, Intl Brotherhood of Electrical Workers.Copies of Answer Have Been Served to Parties & Petitioner by e-mail or Facsimile
ML20212C5521999-09-21021 September 1999
Forwards for Filing,Answer to Firstenergy Nuclear Operating Co & Pennsylvania Power Co in Opposition to Petition to Waive Time Limits & Suppl Comments of Local 29 Intl Brotherhood of Electrical Workers
L-99-144, Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-031999-09-20020 September 1999
Forwards NRC Form 536 Which Addresses Util Proposed Operator Licensing Exam Schedule,In Response to Administrative Ltr 99-03
ML20212B3291999-09-16016 September 1999
Forwards for Filing,Petition to Waive Time Limits in 10CFR2.1305 & Supplemental Comments of Local 29,Intl Brotherhood of Electrical Workers Re Beaver Valley Power Station,Units 1 & 2
L-99-134, Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC1999-09-15015 September 1999
Provides Addl Info to Support 990617 LAR 127,proposing Mods to Heatup,Cooldown & Overpressure Protection Curves.Info Is Provided to Formally Docket Info Discussed During 990820 Telcon with NRC
L-99-138, Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)1999-09-0303 September 1999
Forwards Rev 37 to Issue 4 for BVPS Physical Security Plan (PSP) (Base Plan,Safeguards Contingency Plan & Training & Qualification Plan).Changes to Plan Are Listed.Encls Withheld Per 10CFR2.790(d)
L-99-136, Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls1999-09-0202 September 1999
Forwards Data Point Library (Dpl) Changes,Iaw Provisions of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6, Administrative Implementation Requirements. Ten Listed Dpls Have Specific Setpoint Values Removed,Per Unit 2 Dpls
L-99-098, Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods1999-09-0202 September 1999
Forwards Proposed Changes to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Change Would Reduce Operations QA Program Description Commitments by Limiting Required Onsite Safety Committee Reviews of Mods
L-99-137, Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 31999-08-31031 August 1999
Forwards Issue 3,rev 3 to Bvps,Unit 1 Inservice Testing Program for Pumps & Valves. Encl 1 Provides Summary of IST Program Changes Which Have Been Incorporated Into Issue 3, Rev 3
L-99-022, Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl1999-08-31031 August 1999
Forwards Issue 2,Rev 1 to Bvps,Unit 2 Inservice Testing (IST) Program for Pumps & Valves. Summary of Changes,Encl
L-99-037, Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2 Operations QA Program Description,Iaw 10CFR50.54(a)(3)(ii).Attachment 1 Further Describes Proposed Change & Identifies Reason for Change
L-99-012, Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B1999-08-30030 August 1999
Forwards Proposed Change to Bvps,Units 1 & 2,operations QA Program Description,Per 10CFR50.54(a)(3)(ii),including Description of Proposed Change,Reason for Change & Basis for Concluding Revised Program Satisfy 10CFR50,App B
L-99-132, Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 21999-08-26026 August 1999
Forwards fitness-for-duty Program Six Month Rept for 990101-990630 for Bvps,Units 1 & 2
05000412/LER-1999-007, Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info1999-08-19019 August 1999
Forwards LER 99-007-00, Forced Shutdown Due to Inoperable EDG, Per 10CFR50.73(a)(2)(i).Rept Is Delayed Due to Util Needing Addl Three Days to Address Event Issues on Reportability & Provide Addl Safety Implications Info
L-99-127, Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel1999-08-17017 August 1999
Provides Response to NRC Ltr Requesting Review & Comment of NRC Reactor Vessel Structural Integrity Database by 990901.Inconsistencies Noted in Way Data Characterizes Condition of Reactor Vessel
L-99-124, Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached1999-07-30030 July 1999
Requests Withdrawal of Editorial Changes That Do Not Pertain to Transfer of Operating Authority or Plant Ownership That Are Proposed in LARs 269 & 144.Revised mark-up License Pages Reflecting Changes,Attached
L-99-121, Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements1999-07-28028 July 1999
Submits Data Point Library (Dpl) Changes,Iaw Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,rev 1,Section 3.6, Administrative Implementation Reqiurements
L-99-118, Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 20011999-07-25025 July 1999
Forwards Response to NRC AL 99-02,request for Info Re Estimate of Number of New Licensing Actions Expected to Be Submitted in Fy 2000 & 2001
L-99-120, Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations1999-07-22022 July 1999
Forwards Annual Rept of Facility Changes,Tests & Experiments for BVPS Unit 1,IAW 10CFR50.59.Rept Provides Brief Description of Each Facility & Procedure Change & Summary of Safety Evaluations
L-99-119, Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 9901221999-07-20020 July 1999
Forwards Rev 17 to UFSAR for Beaver Power Station,Unit 1. Submittal Reflects Changes to Facility & Procedures as Described in UFSAR That Were Completed During Annual Reporting Period Endinig 990122
L-99-111, Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes1999-07-15015 July 1999
Forwards Revised Final Typed Pages for LARs 109 & 115, Previously Submitted by 990615 & 28 Ltrs.Revised Pages Replace Those Previously Provided for LARs 109 & 115.With Summary of Editorial Changes
L-99-113, Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by1999-07-15015 July 1999
Forwards Final,Typed TS Pages for LARs 262 & 135,previously Submitted with Editorial Changes Identified by
L-99-112, Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl1999-07-14014 July 1999
Responds to NRC Questions Made During 990518 Meeting with Util Re LARs 220 & 88,for Bvps,Units 1 & 2.Copy of Ltr DLC-99-743,which Is non-proprietary Version of DLC-96-310 & Westinghouse Technical Bulletin ESBU-TB-96-07-R0 Also Encl
L-99-110, Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.61999-07-14014 July 1999
Forwards Changes to ERDS for Unit 1,IAW Requirements of 10CFR50,App E,Section VI.3.a & NUREG-1394,Rev 1,Section 3.6
L-99-105, Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves1999-07-0808 July 1999
Forwards Response to NRC 990420 RAI Re Previous Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves
L-99-108, Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC1999-07-0707 July 1999
Requests Withdrawal of Change Proposed for TS Bases Page B 3/4 2-2 from LARs 1A-262 & 2A-135,originally Submitted by Licensee to NRC
L-99-109, Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-62301999-07-0707 July 1999
Forwards Inservice Insp Ninety-Day Rept Bvps,Unit 2 Outage 7,Year 1999, IAW ASME Boiler & Pressure Vessel Code,Section Xi,Article IWA-6230
L-99-104, Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl1999-06-29029 June 1999
Responds to GL 98-01,Suppl 1, Y2K Readiness at Nuclear Power Plants. Disclosure Rept Encl
L-99-093, Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.51999-06-25025 June 1999
Provides Response to RAI on Proposed Change to Operations QA Program Description.Attachment 2 Provides Revised Markup for Proposed Changes to Affected UFSAR Section 17.2.5
L-99-102, Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl1999-06-22022 June 1999
Forwards Typed,Final TS Pages for LARs 259 & 131.Summary of Description of Plant Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages,Encl
L-99-101, Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal1999-06-22022 June 1999
Submits Response to NRC Oral RAI Concerning Qualifications for Senior Nuclear Executive Ref in 10CFR50.80 Submittal
L-99-062, Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages1999-06-17017 June 1999
Forwards Final TS Pages for LARs 262 & 135,including Summary Description of Plants Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages
ML20196A6601999-06-16016 June 1999
Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1
ML20195H4651999-06-16016 June 1999
Forwards for Filing Answer of Firstenergy Corp in Opposition to Petition for Leave to Intervene of Local 29, Intl Brotherhood of Electrical Workers. Copies of Answer Have Been Served Upon Parties & Petitioner by e-mail
ML20195J5221999-06-16016 June 1999
Forwards Answer of Duquesne Light Co to Petition to Intervene of Local 29,International Brotherhood of Electrical Workers in Listed Matter.With Certificate of Svc
L-99-100, Forwards Typed,Final TS Pages for LAR 109 Re Rcs.Summary Description of Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages Is Provided in Attachment B-1091999-06-15015 June 1999
Forwards Typed,Final TS Pages for LAR 109 Re Rcs.Summary Description of Editorial Changes Incorporated in Addition to Those Addressed in Previously Provided marked-up Pages Is Provided in Attachment B-109
L-99-095, Provides Addl Info to Support LARs 262 & 135,in Response to NRC 990527 Verbal Request.Info Describes Performance of Unit 1 Rod Position Indication Sys & Provides Some Background Info Re Normal Operation of Sys1999-06-15015 June 1999
Provides Addl Info to Support LARs 262 & 135,in Response to NRC 990527 Verbal Request.Info Describes Performance of Unit 1 Rod Position Indication Sys & Provides Some Background Info Re Normal Operation of Sys
L-99-099, Requests Partial Withdrawal of LAR 120,which Requested Review of USQ Due to Increased Calculated Doses for Locked Rotor Event & Use of Unapproved Methodology for Evaluating Small Break LOCA Doses Involving W Natl Safety Advisory Ltr1999-06-14014 June 1999
Requests Partial Withdrawal of LAR 120,which Requested Review of USQ Due to Increased Calculated Doses for Locked Rotor Event & Use of Unapproved Methodology for Evaluating Small Break LOCA Doses Involving W Natl Safety Advisory Ltr
ML20195H3731999-06-0303 June 1999
Forwards Petition to Intervene of Local 29,Intl Brotherhood of Electrical Workers in Matter of Firstenergy Nuclear Operating Co,For Filing
L-99-090, Forwards Summary Review Completed to Verify Adequacy of Design Basis Accident Thermal Overpressure Protection for BVPS Unit 2 Containment Penetrations,Per Request1999-06-0202 June 1999
Forwards Summary Review Completed to Verify Adequacy of Design Basis Accident Thermal Overpressure Protection for BVPS Unit 2 Containment Penetrations,Per Request
L-99-086, Forwards Bvps,Unit 2 SG Exam Rept for Aug 1998.Rept Provided to Document Results of SG Eddy Current Exams Performed in Aug 1998.Summary of Insps Provided in Encl 11999-05-28028 May 1999
Forwards Bvps,Unit 2 SG Exam Rept for Aug 1998.Rept Provided to Document Results of SG Eddy Current Exams Performed in Aug 1998.Summary of Insps Provided in Encl 1
L-99-089, Forwards Annual Financial Repts,Including Certified Financial Statements,Of Dqe,Firstenergy Corp,Ohio Edison Co,Pennsylvania Power Co,Cleveland Electric Illuminating Co & Toledo Edison Co,Iaw 10CFR50.71(b)1999-05-28028 May 1999
Forwards Annual Financial Repts,Including Certified Financial Statements,Of Dqe,Firstenergy Corp,Ohio Edison Co,Pennsylvania Power Co,Cleveland Electric Illuminating Co & Toledo Edison Co,Iaw 10CFR50.71(b)
L-99-084, Forwards Revised marked-up TS & UFSAR Pages to 990303 LARs 259 & 131 Which Revised Qualifications for Operations Mgt & Incorporated Generic Position Titles in Ts.Encl Pages Incorporate NRC Requested Changes,Per Recent Telcon1999-05-27027 May 1999
Forwards Revised marked-up TS & UFSAR Pages to 990303 LARs 259 & 131 Which Revised Qualifications for Operations Mgt & Incorporated Generic Position Titles in Ts.Encl Pages Incorporate NRC Requested Changes,Per Recent Telcon
L-99-082, Dockets Licensee Plan for Bvps,Unit 1,safety-related Small Bore Piping Evaluation Project Discussed in NRC 990311 Public Meeting at BVPS1999-05-17017 May 1999
Dockets Licensee Plan for Bvps,Unit 1,safety-related Small Bore Piping Evaluation Project Discussed in NRC 990311 Public Meeting at BVPS
L-99-071, Notifies of License Withdrawal for J Scott,License SOP-11481,due to Resignation from Employment at BVPS1999-05-12012 May 1999
Notifies of License Withdrawal for J Scott,License SOP-11481,due to Resignation from Employment at BVPS
1999-09-03
[Table view]
Category:UTILITY TO NRC
MONTHYEARML20059L4271990-09-0707 September 1990
Requests Approval for Use of Steam Generator Tube Plugs for Both Mechanical & Welded Applications
ML20059G0821990-09-0404 September 1990
Forwards Application for Amend to License DPR-66,consisting of License Change Request 180,changing Section 3.3.3.2 to Reduce Required Number of Operable Incore Detector Thimbles for Remainder of Cycle 8
ML20059F7551990-08-29029 August 1990
Responds to Unresolved Item 50-334/90-16-01 Noted in Insp Rept 50-334/90-16.Corrective Actions:Initial Training for Maint Group Personnel Responsible for Maintaining Supplied Air Respirators Will Be Supplemented W/Biennial Retraining
ML20059F1501990-08-29029 August 1990
Advises That Permanent Replacement Chosen for Plant Independent Safety Evaluation Group.Position Will Be Staffed Effective 900829
ML20028G8731990-08-29029 August 1990
Forwards fitness-for-duty Program Performance Data for Jan-June 1990,per 10CFR26.71
ML20059D3761990-08-24024 August 1990
Describes Cycle 3 Reload Design,Documents Util Review Per 10CFR50.59 & Provides Determination That No Tech Spec Changes or Unreviewed Safety Questions Involved.Reload Core Design Will Not Adversely Affect Safety of Plant
ML20028G8881990-08-24024 August 1990
Withdraws Operator License SOP-10731 (55-60749) Issued to K Gilbert,Who Resigned
05000412/LER-1990-007, Responds to NRC Re Deviation 50-412/90-12-01 Noted in Insp Rept 50-412/90-12.Corrective Actions:Procedure OM 2.20-4.I Revised to Require Removal of Flanges Following Drain Operations & LER 90-007-00 Issued1990-08-23023 August 1990
Responds to NRC Re Deviation 50-412/90-12-01 Noted in Insp Rept 50-412/90-12.Corrective Actions:Procedure OM 2.20-4.I Revised to Require Removal of Flanges Following Drain Operations & LER 90-007-00 Issued
ML20058P7651990-08-14014 August 1990
Provides Info on Acceptability of Rescheduling Response to Reg Guide 1.97 Ser,Item 4b, Neutron Flux Monitoring Instrumentation. Rescheduling of Util Response Will Be Determined on or Shortly After Meeting W/Nrc
ML20059E0571990-08-10010 August 1990
Forwards Suppl 3 to Nonproprietary WCAP-12094 & Proprietary WCAP-12093, Evaluation of Pressurizer Surge Line Transients Exceeding 320 F for Beaver Valley Unit 2, for Review by 900901.Proprietary Rept Withheld (Ref 10CFR2.790(b)(4))
ML20059E7631990-08-0101 August 1990
Provides Results of Util Evaluation of Licensed Operator Requalification Exam Conducted During Wks of 900709 & 16. Crew That Failed to Meet Expected Performance Level Has Been Successfully Upgraded & re-evaluated to Be Satisfactory
ML20059B8141990-08-0101 August 1990
Requests Exemption from 10CFR26 Re Fitness for Duty Program & 10CFR73 Re Physical Protection of Plants & Matls Concerning Unescorted Access Requirements for Nuclear Generating Stations
ML20056A3471990-07-31031 July 1990
Responds to NRC Bulletin 90-001.Items 1 Through 5 of Requested Actions for Operating Reactors Completed
ML20056A1841990-07-27027 July 1990
Forwards Revised Methodology for Achieving Alternate Ac for Plant,Per 900720 Telcon
ML20055H2581990-07-25025 July 1990
Forwards Decommissioning Rept, Per 10CFR50.33(K) & 50.75(b)
ML20055F7061990-07-0909 July 1990
Responds to NRC Re Dcrdr Requirements as Specified in Suppl 1 to NUREG-0737.DCRDR Corrective Actions Implemented & Mods Determined to Be Operational Prior to Startup Following Seventh Plant Refueling Outage
ML20055D3871990-07-0202 July 1990
Provides Info Re long-term Solution to Action Item 3 of NRC Bulletin 88-008,per 890714 & s.Util Will Continue to Monitor Temp in Affected Lines & Evaluate Results
ML20058K5031990-06-29029 June 1990
Discusses Use of Emergency Diesel Generators as Alternate Ac Source at multi-unit Sites,Per Licensee .Emergency Diesel Generator Load Mgt Methodology Evaluated to Meet Listed Criteria
ML20044A3661990-06-21021 June 1990
Forwards Application for Amend to License NPF-73,consisting of Tech Spec Change Request 44,changing Stroke Time to 60 for Inside Containment Letdown Isolation Valves.Change Determined Safe & Involves No Unreviewed Safety Issue
ML20043G6811990-06-14014 June 1990
Forwards Application for Amends to Licenses DPR-66 & NPF-73, Revising Tech Specs Re Electrical Power Sys - Shutdown & Ac & Dc Distribution - Shutdown
ML20043H9341990-06-14014 June 1990
Forwards Issue 1 to Rev 4 to Inservice Testing Program for Pumps & Valves. Issue 1 Removes Relief Requests Requiring Prior NRC Approval & Adds Certain Program Changes Permitted by ASME XI & Generic Ltr 89-04
ML20043G5981990-06-12012 June 1990
Forwards Monthly Operating Repts for May 1990 for Beaver Valley Units 1 & 2 & Revised Rept for Apr 1990 for Beaver Valley Unit 1
ML20043G6851990-06-12012 June 1990
Forwards Application for Amend to License DPR-66,consisting of Proposed OL Change Request 176,revising Tech Specs to Replace Current Single Overpressure Protection Setpoint W/ Curve Based on Temp
ML20043G7941990-06-12012 June 1990
Responds to NRC 900524 Request for Addl Info Re Proposed Operating License Change Request 156.Clarification of Magnitude of Confidence Level of Westinghouse Setpoint Methodology,As Specified in WCAP-11419,encl
ML20043G8001990-06-11011 June 1990
Forwards Application for Amend to License NPF-73,consisting of Proposed Operating License Change Request 41.Amend Deletes Surveillance Requirement 4.4.9.3.1.d
ML20043H0291990-06-11011 June 1990
Forwards Application for Amend to License NPF-73,consisting of Proposed OL Change Request 40,modifying Heatup & Cooldown Curves Applicable to 10 EFPYs Per WCAP-12406 Re Analysis of Capsule U from Radiation Surveillance Program
ML20043F5251990-06-0707 June 1990
Requests Temporary Waiver of Compliance from Tech Spec Limiting Condition for Operation Re Operability of Containment Isolation Valves During Quarterly Slave Relay Testing.Evaluation to Support Request Encl
ML20043F1361990-06-0404 June 1990
Advises That Chemistry Manual Chapter 5P1, Enhanced Primary to Secondary Leakrate Monitoring Program for Unit 1,per 880328 Request to Recommit to Item C.1 of NRC Bulletin 88-002
ML20043B5971990-05-18018 May 1990
Advises of Delay in Hiring Independent Safety Evaluation Group Replacement to Maintain Five Permanent Personnel Onsite,Per Tech Spec 6.2.3.2.Replacement Will Be Provided within 30 Days of Retirement of Engineer on 900531
ML20043B0511990-05-15015 May 1990
Responds to Telcon Request for Addl Info Re Elimination of Snubbers on Primary Component Supports.Probability of Case B/G Event Extremely Small & Does Not Represent Realisitic Scenario
ML20043B1921990-05-11011 May 1990
Forwards Cycle 8 & Cycle 2 Core Operating Limits Rept,Per Tech Spec 6.9.1.14
ML20042G9761990-05-0808 May 1990
Responds to NRC 900226 Ltr Re Violations Noted in Insp Repts 50-334/89-80 & 50-412/89-80.Corrective Action:Maint Work Request Program Being Upgraded to Include Responsibilities of Nuclear Const Dept & Will Be Issued by 900601
ML20042G8541990-05-0303 May 1990
Forwards Technical Review,Audit Summary & Operability Assessments Re Potentially Invalid Leak Detection Tests Used as Alternative for Amse Section XI Hydrostatic Tests
ML20042G9071990-05-0101 May 1990
Forwards Annual Financial Repts for Duquense Light Co,Ohio Edison Co,Pennsylvania Power Co,Centerior Energy Corp & Toledo Edison Co,Per 10CFR50-71(b)
ML20042F1381990-04-30030 April 1990
Advises That Final SER for Implementation of USI A-46 Will Be Delayed Until Late 1990
ML20042F0991990-04-20020 April 1990
Forwards Response to Request for Addl Info Re Second 10 Yr ISI Program
ML20012F5951990-04-10010 April 1990
Forwards Monthly Operating Repts for Mar 1990 & Revised Operating Data Rept & Unit Shutdown & Power Reductions Sheets for Jan 1990
ML20042E1471990-04-0404 April 1990
Forwards Application for Amends to Licenses DPR-66 & NPF-73, Consisting of License Change Request 174/36,updating Staff Titles to Reflect Nuclear Group Organization
ML20012F6021990-03-30030 March 1990
Submits Supplemental Response to Station Blackout Rule for Plant,Per NUMARC 900104 Ltr.Summary of Changes to Condensate Inventory of Dhr,Effects of Loss of Ventilation, Control Room HVAC & Reactor Coolant Inventory Listed
ML20012E3091990-03-23023 March 1990
Forwards Response to 900308 Request for Addl Info on Reg Guide 1.97 Re Variable for Steam Generator wide-range Level Instrumentation
ML20012E3451990-03-23023 March 1990
Submits Addl Info for Exemption from General Design Criteria GDC-57,including Background Info Describing Sys Operation & Addl Bases for Exemption Request.Simplified Recirculation Spray Sys Drawings Encl
ML20012D6491990-03-19019 March 1990
Requests Retroactive NRC Approval of Temporary Waiver of Compliance Re Tech Spec Limiting Condition for Operation 3.8.2.1 on Ac Vital Bus Operability.Sts Will Be Followed When Inverters Not Providing Power to Vital Bus
ML20012E4091990-03-16016 March 1990
Forwards Inservice Insp 90-Day Rept,Beaver Valley Power Station Unit 1,Outage 7, for 880227-891221,per Section XI of ASME Boiler & Pressure Vessel Code 1983 Edition Through Summer 1983 Addenda,Section XI
ML20012D6181990-03-15015 March 1990
Responds to NRC 900215 Ltr Re Violations Noted in Insp Repts 50-334/89-23 & 50-412/89-22.Corrective Actions:Safety Injection Signal Reset & Plant Returned to Presafety Injection Conditions & Crew Members Counseled
ML20042D7401990-03-14014 March 1990
Forwards Corrected Annual Rept of Number of Personnel Receiving Greater than 100 Mrem & Associated Exposure by Work Function at Plant for CY89.
ML20012D5801990-03-13013 March 1990
Forwards Correction to First 10-yr Inservice Insp Program, Rev 2 to Relief Request BV2-C6.10-1 Re Recirculation Spray Pump - Pump Casing Welds & Relief Request Index
ML20012D6221990-03-13013 March 1990
Forwards Response to Generic Ltr 89-19, Resolution to USI A-47. Recommends All Westinghouse Plant Designs Provide Automatic Steam Generator Overfill Protection to Mitigate Main Feedwater Overfeed Events
ML20012C1791990-03-0909 March 1990
Responds to NRC 900207 Ltr Re Deviations Noted in Insp Repts 50-334/89-25 & 50-412/89-23.Corrective Actions:Written Request Initiated to Identify Unit 2 post-accident Monitoring Recorders in Control Room & Recorders Labeled
ML20012E0911990-03-0505 March 1990
Lists Max Primary Property Damage Insurance Coverages for Plant,Per 10CFR50.54(w)(2)
ML20012B7051990-03-0202 March 1990
Responds to NRC 900131 Ltr Re Violations Noted in Insp Repts 50-334/90-05 & 50-412/90-04.Requests Withdrawal of Violation Re Stated Transport Problem & Reclassification as Noncompliance,Per 10CFR2,App C,Section G
1990-09-07
[Table view] |
Text
_ _ - _ _ _ _ . - _ _ _ - _ - - - - _ _ _ - _ - _ - - . - . - _ - _ - .-
Beaver Valley Power Stanon Shippingport. PA 15077-0004 u Pr e n Nucteer Group (4121643 52 %
April 27, 1989 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Response to Questions on the Unit No. 1 IST Program Gentlemen:
The enclosed attachments have been prepared in response to questions telecopied to my staff regarding the Beaver Valley Unit No. 1 Inservice Test (IST) Program. On January 19, 1989, a conference call between my staff and the NRC's staff addressed six issues regarding the Unit No. 1 IST Program. The attachments to this letter provide the NRC's questions and our responses.
As a result of these questions, several IST Program changes have been developed. Attachment 3 is a copy of our program changes for you to incorporate into your copy of our IST Program for Unit No. 1. We are presently following revision 3 of our program, which is the same revision currently under review by the NRC. The above referenced questions, which are a result of this review, identify the reviewer's concerns. Since we reached verbal agreement during our conference call regarding the reviewer's concerns, these subsequent IST program changes will be implemented prior to Unit No. 1 entering our seventh refueling outage, scheduled to begin September 1, 1989.
If you have any questions regarding this submittal, please call me or members of my staff.
Very truly yours, a Km
. D. Sieber hVicePresident Nuclear Group Attachments (3)
- 1. Response to NRC Questions
- 2. LHSI Full Flow Acceptance Criteria Methodology
- 3. Revisions to Inservice Test Program for Unit No. 1 (Rev. 4) 8905110072 890427 PDR ADOCK 05000334
, nC y
lBanvar VOllcy Powsr Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 Page 2 I
cc: Mr. J. Beall, Sr. Resident Inspector Mr. W. T. Russell, NRC Region I Administrator Mr. P. Tam, Sr. Project Manager Director, Safety Evaluation & Control (VEPCO) l 1
l l
l
- _ - - - - - - - _ --- m
i 1
l l ATTACHMENT 1 Response to NRC Questions Recardina the Unit No. 1 IST Procram
- 1. Valve relief request 6 asks relief from quarterly exercising and timing of category A valve MOV-CH-142, residual heat a removal letdown to the chemical and volume control system. It {
1 appears that this valve has only a passive safety function.
If so, this valve is not required to be exercised and timed.
The valve was categorized A passive in the previous program submittal and it appears that the licensee has incorrectly ;
changed the categorization. If indeed this valve is passive, then the licensee should delete this relief request.
Response
Valve MOV-CH-142 had been conservatively classified as active in Revision 3 of the IST Program since it can be used in a flow path for bringing the plant to a cold shutdown. During the conference call, the NRC consultant, EG&G, questioned the need for this classification. We agreed to classify the valve as passive. We have revised our program (Revision 4) to change the valve classification to passive on the Valve Outline Sheet and delete relief request (RR) 6 which was based on the requirement for stroke testing an active valve.
- 2. Valve relief request 17 asks relief from quarterly exercising of category A/C valves S1-10, SI-11, SI-12, SI-15, SI-16, and SI-17, low head safety injection cold and hot leg check valves. The licensee has proposed to exercise these valves during refueling outages. How is the licensee verifying that these valves. individually full stroke exercise?
Response
As stated during the conference call, check valves SI-15, SI-16 and SI-17 are located in the three parallel recirculation branch lines from the Low Head Safety Injection (LHSI) system to the Reactor Coolant System hot legs. The valves are located upstream of the connection of the High Head Safety Injection (HHSI) system branch lines to the LHSI system branch lines. The LHSI pumps are no longer used to deliver flow via these branch lines to the Reactor Coolant System hot legs during the post-LOCA hot leg recirculation mode on Beaver Valley Power Station Unit 1; only the HHSI system is used for this function. Consequently, when operating in the hot leg recirculation mode, LHSI check valves SI-15, SI-16 and SI-17 are not required to open. Agreement was reached to change the classification of these valves in the Unit No. 1 IST Program from Active to Passive. Our program has been revised to change the valve classification on the Valve Outline Sheet and we have deleted reference to these valves in RR 17 since stroke testing is not necessary.
f
- i
- Attachmsnt 1 Response to NRC Questions Regarding the Unit No.1 IST Program Page 2 Check valves SI-10, SI-11 and SI-12 are located in the three parallel injection branch lines from the Low Head Safety Injection (LHSI) system to the Reactor Coolant System cold legs. To satisfy the stroking requirements, these valves will be exercised during refueling outages as part of a full flow test of the LHSI system. A description of the test methodology and justification that the test is suitable for full-stroke exercising the subject check valves is provided in l the response to NRC Question 4.
- 3. Valve relief request 18 asks relief from quarterly exercising of category C valves SI-20, SI-21, and SI-22, combined low head and high head safety injection hot leg check valves. The licensee has proposed to exercise these valves during refueling outages. How is the licensee verifying that these valves individually full-stroke exercise and is both high head and low head combined flow necessary for a full-stroke exercise (maximum design accident flow rate)?
Note: Valve relief request 18 also includes valves SI-100, SI-101, and SI-102. These valves are in flow rate instrumented high head safety injection lines and can be individually verified to full-stroke exercise.
Response
During the conference call, we explained that check valves SI-20, SI-21 and SI-22 are located in the three parallel recirculation branch lines from the Low Head Safety Injection (LHSI) system to the Reactor Coolant System hot legs. The valves are located downstream of the connection of the High Head Safety Injection (HHSI) system branch lines to the LHSI system branch lines. The LHSI pumps are no longer used to deliver flow via these branch lines to the Reactor Coolant System hot legs during the post-LOCA hot leg recirculation mode on Beaver Valley Power Station Unit 1; only the HHSI system is used for this function. Consequently, when operating in the hot leg recirculation mode, LHSI check valves SI-20, SI-21 and SI-22 will function similar to HHSI check valves SI-100, SI-101, and SI-102 and will experience only HHSI flow.
l l
- Attachment 1 ;
Response to NRC Questions Regarding the Unit No.1 IST Program Page 3 As a result of further review, we proposed that since check valves SI-20, SI-21 and SI-22 will only experience HHSI flow, they will be individually verified to full stroke exercise in a manner similar to HHSI check valves SI-100, SI-101 and SI-102 us part of the HHSI full flow test utilizing the flow instruments in the individual HHSI lines. Since the maximum design accident flow rate for the subject valves is HHSI flow, only HHSI flow will be used to full-stroke exercise these check valves.
We have provided a revised Basis for Relief in RR 18 to a basis used for other HHSI branch line check valves (ie: SI-83, 84 in RR22).
- 4. Valve relief request 19 asks relief from quarterly exercising of category A/C valves SI-23, SI-24, and SI-25, combined low head and high head safety injection cold leg check valves.
The licensee has proposed to exercise these valves during refueling outages. How is the licensee verifying that these valves individually full-stroke exercise and is both high head and low head combined flow necessary for a full-stroke exercise (maximum design accident flow rate)?
Response
We presented our methodology for full-stroke exercising these valves during the conference call. We believe our testing provides acceptable valve stroking to satisfy code requirements. Due to the complex discussion that took place, and since it represents a new approach for defining check valve operability in parallel flow paths, it was agreed that we would submit our methodology for NRC review.
Check valves SI-10, SI-11 and SI-12, SI-23, SI-24 and SI-25 are located in the three parallel injection branch lines of the Low Head Safety Injection (LHSI) system to the Reactor Coolant System cold legs. The LHSI system is not designed with flow instruments in the individual branch lines but does include flow instruments in the discharge header of each LHSI pump for use in determining total flow delivery from a LHSI pump. Total flow instruments are typically used during plant startup testing to provide system flow test data that is evaluated against acceptance criteria to verify that the LHSI system performance exceeds the design basis upon which the FSAR accident analyses are based. Since plant startup programs typically utilize LHSI full flow test data to verify compliance with system design bases, Duquesne Light Company initiated an evaluation to determine if a periodic LHSI full flow surveillance test could be used to verify continued acceptable performance of the LHSI system, including acceptable operation of the subject check valves.
.________-______-____ - _ _ _ N
L l
- Attachment 1 ;
Response to NRC Questions Regarding the Unit No.1 IST Program )
Page 4 Based on this evaluation, it is concluded that a full flow test of the LHSI system is an acceptable method to determine continued acceptable operation of the LHSI system, including acceptable operation of the subject check valves. The basis for this conclusion is that a quantitative full flow acceptance criteria can be established that, if satisfied, will confirm that the flow rates in the LHSI injection branch lines exceed the minimum safeguards flow rate requirements assumed in FSAR accident analyses. Test results that satisfy the FSAR accident analyses requirements also verify acceptable operation of the subject check valves.
Duquesne Light Company's evaluation has revealed additional insight into the use of a full flow test to verify acceptable branch line flows. The evaluation has shown that for a full '
flow test to be acceptable, the test must utilize quantitative flow acceptance criteria that verifies acceptable flow distribution in the three LHSI system injection branch lines as well as total system flow. Furthermore, the evaluation has shown that the requirement to verify acceptable flow distribution in the LHSI injection branch lines is the most limiting with respect to the development of a full flow acceptance criteria. The resultant full flow acceptance criteria is very restrictive relative to the best estimate LHSI full flow. Although very restrictive, an acceptance criteria can be developed that can function as a screening criteria for acceptable system performance, including check valve operation. Obtaining test results that satisfy the acceptance criteria will verify that actual LHSI system flow rates (i.e., both total flow and branch line flows) exceed the minimum LHSI flow rates taken credit for in the FSAR accident analyses. Since the flows in the branch lines exceed the FSAR accident analyses requirements, it can be concluded that the check valves in each branch line are opening acceptably.
Alternatively, failure to satisfy the acceptance criteria only indicates that actual LHSI flow rates may not satisfy design basis requirements. In this situation, the test is inconclusive since the acceptance criteria has been developed as a screening criteria based on worst case assumptions with respect to adverse effects on flow distribution. LHSI flow rates and operation of the subject check valves may still be acceptable, however, additional testing and/or evaluation is needed to determine compliance with FSAR accident analyses requirements.
,Attachm@nt 1 Response to NRC Questions Regarding the Unit No.1 IST Program Page 5 Attachment 2 provides a detailed description of the methodology developed to define full flow test acceptance criteria that will verify acceptable operation of the subject check valves. Based on the information presented in the ,
attachment, it is concluded that full flow test results that satisfy test acceptance criteria will serve to verify that ,
both total LHSI system and branch line flow rates are greater than the minimum safeguards flow requirements ascumed in the FSAR accident analyses. Verification that branch line flows exceed FSAR requirements confirms that the subject check valves are individually opening acceptably to satisfy design basis requirements.
The full flow test and associated acceptance criteria have been developed such that only LHSI flow is required to test for acceptable opening of the subject check valves. The methodology used to dr vlop the acceptance criteria takes into consideration the int raction of High Head Safety Injection (HHSI) flow and LHSI flow in the piping sections that contain check valves SI-23, SI-24, and SI-25. Although the test is performed with only LHSI flow, the acceptance criteria is developed to ensure that both HHSI and LHSI minimum safeguards flow requirements will be satisfied when both pumps are operating.
Finally, Duquesne Light Company is now performing a review program for check valves installed in safety related systems of BVPS Units 1 and 2. This review, which utilizes the guidelines of EPRI Project RP-2233-20, Application Guidelines for Check Valves in Nuclear Power Plants, includes a determination of the minimum flow velocity required for full disc lift. For the subject 6-inch check valves, the preliminary results of the Check Valve Review Program for minimum flow velocity for full disc lift were compared against the individual LHSI branch line flow rates that are required .
to satisfy the LHSI full flow acceptance criteria. The !
conclusion of this comparison is that the LHSI system branch line flow rates that are required to satisfy the LHSI full ,
flow acceptance criteria are greater than the flow rates l required to satisfy the minimum flow velocity requirement for check valve full disc lift. Consequently, test results that satisfy the LHSI full flow test acceptance criteria will not only verify acceptable check valve operation from an accident anlyses standpoint but should also be sufficient to individually full-stroke exercise the subject check valves.
This preliminary conclusion will be confirmed following completion of the Check Valve Review Program.
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.' Attachment 1-Response to NRC Questions Regarding the Unit No.1 IST Program Page 6
- 5. Valve relief request 21 asks relief from quarterly exercising of category A/C valves SI-48, SI-49, SI-50, SI-51, SI-52, and SI-53, safety injection accumulator series discharge check valves. The method that the licensee is utilizing to full-stroke exercise these valves does not appear to be adequate to verify a full-stroke exercise. The data obtained from this testing appears to contain an excessive amount of data scatter.
ResDonse:
A discussion about the data scatter revealed one data point which stood out from the rest. This was explained as being a result of test conditions being different from previous tests since the reactor coolant system (RCS) was not verified to be solid in the loops. As a result, when the 'C" accumulator flow test was performed, the pressurizer did not measure the total insurge to the RCS. Subsequent testing improved test conditions which resulted in closely correlated test data.
This explained the apparent data scatter to the reviewer.
Since the reviewer's concern was addressed, our method of full stroke exercising these valves was to be further evaluated.
The reviewer did not request any further information pertaining to this question.
We would like to reiterate our position stated in earlier correspondence regarding the determination of valve operability. This test method eliminates the need to operate in a reduced RCS inventory condition which is necessary if valve disassembly is considered. Loss of RHR events in the industry have illustrated a need to reduce the duration and frequency of activities that require a reduced RCS inventory condition. Considerable man-rem dose savings also exist through the use of our test method versus valve disassembly.
- 6. Valve relief request 36 for the emergency diesel generator air start solenoids does not have sufficient technical justification to be granted relief. Also, the proposed timing method on a refueling outage frequency is inadequate. Would it be possible for the licensee to utilize the same relief request for which Unit 2 was granted relief (Unit 2 SER dated January 25, 1988)?
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. Attachment l' Response to NRC-Questions Regarding the Unit No.1 IST Program Page 7 Response: i During the conference call, the design differences between the two units diesel generators was discussed. Unit No.1 cannot perform testing in a manner identical to Unit No. 2. We proposed revising. the monthly diesel generator test to stroke and time the solenoid operated valves using the. alarm circuitry of the diesel generators. The NRC agreed with this approach.
Our program has been revised to change the wording of the relief request to use the diesel generator start failure alarm to time'the valve strokes.
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, ATTACHMENT 2 Page 1 of 10 i
DESCRIPTION OF l LHSI FULL FLOW ACCEPTANCE CRITERIA METHODOLOGY l This attachment describes the methodology for developing acceptance criteria for the Low Head Safety Injection (LHSI) system full flow test. Acceptance criteria developed to this methodology ensure that ,
actual LHSI system flow rates are greater than the minimum )
safeguards design basis flow rates used in FSAR accident analyses. ]
Satisfying the acceptance criteria also verifies acceptable !
operation of the LHSI system check valves, including check valves '
SI-10, SI-11, SI-12, SI-23, SI-24 and SI-25 on Beaver Valley Power Station Unit 1.
l Attached Figure A provides a simplified schematic of the Unit 1 j LHSI system that identifies the subject check valves. Figures B J and C illustrate the application of this methodology to Beaver I Valley Power Station Unit 1. Figures B and C are based on work currently being performed on Beaver Valley Power Station Unit 2 to develop LHSI full flow acceptance criteria utilizing this methodology. Following completion of the Unit 2 acceptance criteria, the methodology will be applied to Unit 1.
The activities and elements that comprise the LHSI Full Flow Acceptance Criteria Methodology are itemized and described below:
Item 1) Identify FSAR Accident Analyses Requirements for LHSI Flow The first activity was to review the functional and performance design basis requirements for the LHSI system to identify the requirements for LHSI flow. These requirements consist of a number of cases which identify the minimum safeguards flow requirements which are used in the FSAR accident analyses.
Several of the cases take credit for total injection flow from all three LHSI branch lines. Other cases take credit for injection flow from only two out of the three injection branch !
lines, assuming that the injection line with the largest flow rate spills to containment via the Reactor Coolant System pipe break. These two types of cases are referred to as the no spilling line cases and the spilling line cases, respectively.
For a full flow test to verify acceptable operation of the LHSI system, acceptance criteria must ensure that both the no spilling line cases (i.e., total flow from all three injection branch lines) and the spilling line cases (i.e., flow from only two out of three injection branch lines) are satisfied.
Spilling line flow requirements are also referred to as branch line flow distribution requirements.
This review also identified the assumptions, analyses and analytical model of the LHSI system used to calculate the LHSI minimum safeguards flow rates used in the FSAR accident analyses.
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, ATTACHMENT 2 Page 2 of 10 l l
Item 2) Identify Actual LHSI Flow Performance Plant performance test data was then reviewed to identify the actual performance of the LHSI system. This review included LHSI pump performance and system flow performance. Since flow performance is dependent on system test configuration, the alignment of the LHSI system as well as system elevation heads and static pressures during the performance of the tests were also reviewed.
Item 3) Develoo Analytical Model for LHSI Flow Evaluation )
l An analytical model for evaluation of LHSI flows was developed based on the model identified in Item 11 for the minimum safeguards design basis flow calculation. This model was benchmarked relative to the actual plant performance data of Item 2) in order to define the actual LHSI system resistance.
The actual LHSI system resistance curve is'shown on Figure B along with the actual LHSI pump performance data obtained from Item 2). Actual LHSI flow for the test configuration is shown on Figure B as the intersection of the actual pump curve and the actual system resistance curve.
Item 4) Evaluate Sensitivity of LHSI Flow to Degradation in Check Valve Performance Sensitivity analyses were performed using the model developed in Item 3) to determine the sensitivity of total flow and branch line flow distribution to the location of a check valve that does not open properly. This was done by adding additional piping resistance at each location where there was a check valve to simulate additional head loss due to a check valve that was not opening properly. These analyses were performed with one LHSI and one High Head Safety Injection (HHSI) pump running to properly account for the interaction of HHSI and LHSI flow in that portion of the LHSI system in which j check valves SI-23, SI-24 and SI-25 are located. This arrangement is consistent with that used to develop the minimum safeguards flow cases used in FSAR accident analyses. The sensitivity analyses served to locate the most critical LHSI check valve from the standpoint of adverse affect on LHSI flow distribution. This evaluation resulted in the conclusion that the most critical check valve location was:
a) in the injection branch line that has the highest piping resistance. This location is critical due to the spilling line assumption. In developing the injection flows for the spilling line cases, the flows that are taken credit for are those from the two lines delivering the smallest flows (i.e., the lines with the highest piping resistance). The line with the highest flow (i.e., the line with the lowest piping resistance) is assumed to spill and is not taken credit for in the minimum safeguards flow. By adding additional resistance to the path with the highest resistance, the flow in this branch
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'. ATTACHMENT 2 Pago 3 of 10 line is further reduced while the flows in the other two injection branch lines are increased. This results in the minimum safeguards injection flow (i.e. , the sum of the two lines delivering the smallest flows) being reduced while the spilling line flow is increased. Thus, the addition of resistance to the branch line with the highest resistance has the largest adverse affect on LHSI flow distribution with respect to compliance with minimum safeguards flow requirements.
b) downstream of the point where the HHSI injection lines connect to the LHSI injection branch lines. This location is critical since the check valves in this location receive a higher flow (i.e., LHSI and HHSI combined flow) than do the check valves located upstream of the HHSI connection (s) to the LHSI injection branch line(s). Thus, for a given additional resistance, the branch line will experience additional head loss which will have the most adverse effect on LHSI flow distribution and compliance with the minimum safeguards flow requirements.
Item 5) Determine Amount of Allowable Degradation in Check Valve Performance Sensitivity analyses were performed using the model developed in Item 3) to determine the amount of additional resistance that could be added to the worst case check valve location of Item 4) and not reduce either total flow or branch line flow distribution below the minimum safeguards flow requirements assumed in the FSAR accident analyses. Since the minimum safeguards flow requirements are developed with one LHSI and one HHSI pump running to account for the interaction between LHSI and HHSI flow in the SI system, the sensitivity analyses to define allowable additional resistance also utilized one LHSI and one HHSI pump running. The sensitivity analyses utilized minimum safeguards pump performance curves consistent with the pump performance curves used to develop the minimum safeguards flow requirements assumed in the FSAR accident analyses.
The sensitivity analyses showed that additional resistance at the worst case check valve location resulted in a reduction in the associated LHSI branch line flow while the flow in the other two branch lines increased. The net effect of this flow redistribution was a relatively large reduction in associated branch line flow but only a relatively small reduction in total LHSI flow. Due to this relationship between branch line flows and total flow, the spilling line minimum safeguards cases were determined to be the most limiting with respect to the amount of additional resistance that could be added and still satisfy the FSAR accident analyses requirements. Based on the sensitivity analyses, a limiting amount of acceptable additional resistance was identified for the worst case check valve location. The new system resistance curve that includes this additional resistance is shown on Figure B.
I
. ATTACHMENT 2 Page 4 of 10 This new system resistance curve is the maximum system )
resistance that can be tolerated by the LHSI system and still satisfy the minimum safeguards flow requirements of the FSAR accident analyses. It is conservative in that it assumes that all additional resistance is added at the worst case check valve location with respect to adverse effect on LHSI branch line flow distribution.
Item 6) Define Acceotance Criteria for Check Valve Operation Since LHSI system resistance cannot be directly measured and compared against the requirement defined in Item 5), it is necessary to define an appropriate acceptance criteria that can be used to verify LHSI system resistance. This can be accomplished through defining a full flow acceptance criteria based on the system resistance curve of Item 5) in combination with pump performance. Acceptable system resistance will thus be determined by measuring system flow rates and pressures and comparing them with the full flow acceptance criteria.
The LHSI full flow test configuration is one in which one LHSI pump is running and the valves are in the cold leg injection alignment. The HHSI pumps will not be operating during the LHSI full flow test. Consequently, acceptance criteria for the LHSI full flow test is developed through performing analyses using the maximum allowable system resistance determined in Item 5) with one LHSI pump running. The full flow acceptance criteria is the intersection of the maximum system resistance curve and the pump performance curve.
Since flow is a function of pump head as well as system resistance, the flow acceptance criteria can shift along the maximum allowable resistance curve as a function of pump performance. LHSI pump performance is acceptable if it is greater than the minimum safeguards pump performance curve used to develop minimum safeguards flows for the FSAR accident analyses. Thus, analyses are performed to determined acceptance criteria for the 100% pump performance case and the minimum safeguards pump performance case. The results of the analyses are shown on Figure B as the intersection of these pump performance curves with the maximum system resistance curve. For the 100% pump performance curve, flow must be i
greater than the intersection of the pump head curve and the maximum system resistance curve. Should the pump performance degrade from the 100% performance curve, the intersection of the pump performance curve with the maximum system resistance curve will move down the resistance curve. If pump performance degrades to the minimum safeguards performance curve, the full flow acceptance criteria slides along the maximum system resistance curve to the intersection of the maximum system l
resistance curve and the minimum safeguards pump performance curve. Thus, the intersection of the pump performance curve and the maximum system resistance curve determines the full flow acceptance criteria. The acceptance criteria decreases with decreasing pump performance.
o -
, ATTACHMENT 2 Page 5 of 10 I
Item 7) Present Full Flow Acceptance Criteria for Test Configuration The acceptance criteria developed in Item 61 is based on a l maximum system resistance curve and LHSI pump performance between the 100% performance curve and the minimum safeguards performance curve. It is also based on the assumption that the differential head on the LHSI system is equal to zero (i.e.,
the suction elevation head and static pressure is equal to the '
discharge elevation head and static pressure). Since the full flow test will be performed with different water levels in the Refueling Water Storage Tank and the Reactor Coolant System, the acceptance criteria can be adjusted to account for different suction and discharge conditions. This is accomplished analytically by running the LHSI model with different system differential heads (i.e., sucLion elevation head and static pressure minus discharge elevation head and static pressure) and presenting the resulting full flow acceptance criteria as curves that are a function of system differential head. Full flow acceptance criteria curves for the 100% and minimum safeguards LHSI pump performance curves are shown in Figure C.
Item 8) Evaluation of LHSI Full Flow Test Data The LHSI system valve alignment used in the full flow surveillance test is the same as the alignment used in the LHSI flow analyses used to develop test acceptance criteria. One LHSI pump is run at a time as the pumps are tested separately.
As the test is performed, the LHSI flow is obtained as well as the LHSI pump developed head, LHSI system suction elevation head and static pressure and LHSI system discharge elevation head and static pressure. The elevation heads and static pressures are used to obtain a system differential head corresponding to the obtained LHSI flow. The values for total flow, pump developed head and system differential head are then used to determine the acceptability of LHSI flow and check valve operation as follows:
a) The point value for total flow / system differential head is plotted on Figure C. If this point is to the right of the acceptance criteria curve for the 100% pump performance curve, acceptance criteria are satisfied. All subject check valves are opening adequately for the LHSI system flow to satisfy the minimum safeguards flow requirements for FSAR accident analyses.
b) If the test point is to the left of the acceptance criteria curve for the minimum safeguards performance curve, the cubject full flow test cannot be used to verify acceptable LHSI flow and check valve performance. The check valves may or may not be operating properly; additional testing and/or evaluation are required to evaluate LHSI system flow distribution and check valve I
1 i
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- ATTACHMENT 2 Pagn 6 of 10 performance. No further conclusions can be drawn from the full flow surveillance test since the acceptance criteria has been developed as a conservative screening criteria based on the assumption that all additional system resistance is added to the system in the check valve location most critical to system flow distribution. If the failure to satisfy the acceptance criteria is due to additional resistance added elsewhere in the system, the ,
adverse affect on flow distribution will be less severe 1 and the reduction in branch line flow with respect to total flow will be less severe. Thus, minimum safeguards flow requirements as assumed in the FSAR accident analyses may still be satisfied. i l
c) If the test point is between the full flow acceptance i criteria curves for the minimum safeguards pump !
performance and 100% pump performance cases, the LHSI flow may be acceptable depending on actual pump performance.
The following process is used to determine the appropriate acceptance criteria:
o The Pump Performance Ratio is first determined from Figure B based on the actual pump head test data and the corresponding 100% and minimum safeguards pump heads at the same flow. The pump Performance Ratio is equal to:
Actual Pump Head - Minimum Safeguards Pump Head 100% Pump Head - Minimum Safeguards Pump Head o The Flow Acceptance Criteria Range is then determined from Figure C for the system differential head. This range is determined by subtracting the acceptance criteria value for the minimum safeguards pump performance case from the acceptance criteria value for the 100% pump performance case.
o The actual full flow acceptance criteria is then determined by multiplying the Pump Performance Ratio times the Flow Acceptance Criteria Range and adding the result to the acceptance criteria value for the minimum safeguards performance case. This actual acceptance criteria is applicable to the actual pump performance head and can be used to evaluate the acceptability of the total flow / system differential head test point. I,
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ - . _ _ _ _ _ _ . i
', ATTACHMENT 2 Pago 7 of 10 If the actual flow is greater than the applicable acceptance criteria; LHSI flow satisfies FSAR accident analyses assumptions and the subject check valves are opening adequately. This is the same situdtion as in Item 8a) above. If the actual flow is less.than the acceptance criteria, the test results are inconclusive relative to the performance of the LHSI system and subject check valves. Additional testing and/or evaluation are required to evaluate LHSI system and check valve performance. This is the same situation as in Item 8b) above.
As an example of this case, evaluate the acceptability of a total flow / system differential head test point of 4300 gpm/0 feet that is obtained with a pump developed head of 183 feet. The pump developed head of 183 feet at 4300 gpm is greater than the minimum safeguards pump performance curve on Figure B, thus, pump performance is acceptable.
The total flow / system differential head test point falls between the full flow acceptance criteria curves on Figure C for the minimum safeguards pump performance case and the 100% pump performance case, thus Item 8c) applies.
Applicable full flow acceptance criteria based on actual LHSI pump performance is determined as follows:
183 feet - 179 feet o Pump Performance Ratio = -------------------
= 0.31 192 feet - 179 feet o Flow Acceptance Criteria Range = 4335 gpm - 4220 gpm
= 115 gpm o Acceptance Criteria = 4220 gpm + ((0.31) * (115 gpm)]
= 4256 gpm At a system differential head of 0 feet, the example flow ,
of 4300 gpm is greater than the applicable full flow acceptance criteria of 4256 gpm based on actual LHSI pump performance. Thus, the LHSI full flow acceptance criteria .
are satisfied. The LHSI system flow rates are greater ;
than assumed in the FSAR accident analyses and the subject check valves are opening acceptably.
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- ATTACHMENT 3 DUQUESNE LIGHT COMPANY Beaver Valley Power Station INSERVICE TESTING PROGRAM FOR PUMPS AND VALVES APPROVAL SHEET Unit 1 - IST Program ISSUE 2 Unit Nuclear OSC Oper. Supvr. App,,,,g R:v. Ravtav Ravtsw Effeeeive Ns. Pages Issued Initial /Dete Date $1gnature Date Date 4 Valve Testing Outlines - [ BV-OSC- 82 Va e est n Relief Requests -
RR6, RR17, RR18 &
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RELIEF REQUEST S Valve No.: CH-31 Category A/C- Class 2 Function: Charging header inside con *.ainment isolation check valve.
Test Requirements: Quarterly full stroke.
Basis for Relief: This normally open check valve must close to fulfill its safety function. . Valve. closure can only be checked by a leak test and there is no instrumentation to monitor upstream pressure. Therefore, relief is requested from-quarterly and cold' shutdown stroke tests.
Alternate Test: Leak rate test during refueling outages per BVT 1.3 - 1.47.11.
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RELIEF REQUEST 17 Valve No.: SI-10 SI-11 SI-12 Category A/C Class 1 Function: To open on a safety injection and to prevent reverse flow from the higher pressure RCS and HHSI system to the LHSI low pressure system.
Test Requirements: Quarterly stroke.
Basis for Relief: These check valves are normally shut during power operation but are required to open in l the event of a safety injection. Due to the lack of installed instrumentation, and the relative system pressures, relief from quarterly full- and part-stroke. exercising is requested. In addition, relief is requested from full- or part-stroke exercising at cold !
shutdown because testing would require j full-flow injection to the RCS where there is !
insufficient expansion volume to receive the additional inventory. ;
Alternate Test: Full-stroke exercise per OST 1.11.14 and I reverse flow stroke per leak test OST 1.11.16 during refueling outages.
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ISSUE 2 REVISION 4 ;
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RELIEF REQUEST 18 l
Valve No.: SI-20 SI-100 SI-21 SI-101 SI-22 SI-102 Category C Class 1 Function: To provide isolation between the RCS and the HHSI and LHSI systems.
Test Requirements: Qaarterly full stroke exercise.
Basis for Relief: .These check valves cannot be full or part stroked at power at any frequency due to the potential for a premature failure of the injection nozzles caused by the thermal shock from a cold water injection. Relief from stroke testing at cold shutdowns is also requested since this could result in a low temperature overpressurization of the RCS.
Alternate Test: Full-flow stroke exercise per OST 1.11.14 during refueling outages.
ISSLT 2 REVISION 4
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RELIEF REQUEST 36 i Valve No.: SOV-DA-101 SOV-DA-103 SOV-DA-102 SOV-DA-104 Category B . Class 3 Function: Diesel Generator Air Start SOVs. j
. Test Requirements: Quarterly stroke and time. !
Basis for Relief: These valves are quick acting and d, , at have i position indication. The operation of these valves wil be monitored by cach individual diesel generator's start failure alarm circuit. Malfunctions which will cause the j annunciator panel START FAILURE light to come on and the alarm bell to ring are.
- 1. Engine fails to crank above 40 RPM within l 3 seconds after a start signal is j i received or j j
- 2. Engine cranks above. 40 RPM within 3 ,
seconds, but fails to exceed 100 RPM l within 4 sconds after a start signal is received.
f Individual-valves will be tested monthly on an ;
alternating fraquency by using a different set l L
of air starting motors.each month to crank the engine. #51s will ensure each bank is capable of starting the diesel generator in .the required time and that the air start SOVs are not degrading. l 1
Alternate Test: Stroked and indirectly timed by the START '
- j. on alternating FAILURE annunciator an frequency in conjunction with monthly diesel generator OSTs 1.36.1 & 2 to ensure compliance with the ASHE XI requirement for stroke testing on a quarterly frequency. O i
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