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MONTHYEARML20155E6801986-04-0909 April 1986
Forwards Request for Addl Info Re Util 830415 Response to Generic Ltr 82-33 Concerning Spds.Pass/Fail Acceptance Criteria for Each Type of Device Should Be Defined.Response Requested within 45 Days of Ltr Receipt
Project stage: RAI
ML20150B5401988-03-0202 March 1988
Insp Repts 50-334/88-03 & 50-412/88-02 on 880125-29. Violation Noted.Major Areas Inspected:Licensee Implementation of Radiological Controls Program During Unit 1 Outage,Diving Operations,Posting & Labeling & ALARA
Project stage: Request
ML20151G3461988-07-18018 July 1988
Forwards Proprietary & Nonproprietary Suppl 7,Rev 0 to WCAP-10170, Westinghouse SPDS Design & Verification... Process for Beaver Valley Unit 1 Nuclear Station
Project stage: Other
ML20154G0961988-09-12012 September 1988
Grants 880718 Request for Withholding WCAP-10170 Suppl 7, Westinghouse SPDS Design & V+V Process for Beaver Valley Unit 1 from Public Disclosure (Ref 10CFR2.790)
Project stage: Other
1988-03-02
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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
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O Beaw Power Staton Shopogport. PA 15077@04
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July 18, 1988 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 Request for Plant Specific Information on SPDS (TAC 51221)
Gentlemen:
By a letter dated April 8, 1987, the NRC requested that DLC submit a plant specific safety analysis report and implementation plan in accordance with the requirements of supplement 1 to NUREG-0737. Attached is the requested document prepared for your review in accordance with the plan sul.nitted to the NRC by DLC letter dated March 2, 1988.
As this submittal contains information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the commission's regulations.
i Accordingly, it is respectfully requested that the information :
which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.790 of the Commission's regulations. Correspondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse affidavit should reference CAW-88-059 and i should be addressed to R. A. Wiesemann, Manager Regulatory and ,
Legislative Affairs, Westinghouse Electric Corporation, P. O. Box <
355, Pittsburgh, Iennsylvania 15230.
If you have any questions or comments regarding this submittal, please contact me or members of my staff.
Sincerely yours, CW.PM , , op . .
00.3 L? 0 NC f$ I e $40 n- ;
$ SIC l g$$ J. D. Sieber Ill0WCAf*iC00(^h10 Men h to WCA8. (0 Vi':e Pre 11 dent '
Nuclear Group cc: Mr. J. Beall, Sr. Resident Inspector .
Mr. W. T. Russell, NRC Region I Admir + .
Mr. P. Tam, Project Manager Director, Safety Evaluation & Contruf .i 8807280304 sso7tg gDR ADOCK 05000334 PDC d ;
M Bhaver Vallcy Power Station, Unit No. 1 Dockot No. 50-334, License No. DPR-66 Page 2 bcc: W. S. Lacey J. O. Crockett N. R. Tonet W. R. Hunter R. F. Balcerek R. J. Swiderski T. P. Noonan R. D. Hecht J. V. Vassello G. S. Sovick G. L. Beatty K. D. Grada P. A. Cadena D. Schmitt R. E. Martin J. Matsko V. Palmicro E. Humer M. W. Rencheck R. J. Druga T. G. Zyra A. Lerczak F. D. Schuster T. W. Burns J. G. Proven
!. . R . Freeland R, Cavaliere l
t l
Safety Analysis Report For Beaver Valley Unit 1 Safety Parameter Display System (SPDS)
Supplement 1 to NUREG 0737 requires that each licensee prepare a written Safety Analysis Report (SAR) which describes the basis upon which the parameters were selected for display on the SPDS. This submittal provides that basis and further documents DLC's evaluation of the SPDS to the requirements of Supplement 1 to NUREG 0737.
DLC's purchase of the BV-1 SPDS in May 1980 pre-dated the NRC design verification audit of the generic Westinghouse SPDS performed in October 1982. This audit was based upon activities on the genaric Westinghouse SPDS after 1980. This audit was performed in parallel with the design / implementation activities of the BVPS-1 SPDS. The WCAP used as the basis for the generic Westinghouse SPDS design (WCAP-10170) and appendices (A through E) document the Westinghouse SPDS design and verification process with regard to the BVPS-1 SPDS system. However, because of the parallel activities associated with the generic Westinghouse SPDS and BVPS-1 SPDS, certain of the activities associated with the generic Westinghouse SPDS do not apply to BVPS-1 SPDS. The activities that do not directly apply to BVPS-1 SPDS (differ from the generic Westinghcuse SPDS design), are; the combination of the Factory and Site Acceptance tests, the non-independence of software verifiers and the NRC Verification and Validation process.
Supplement 7 to WCAP-10170 is appended hereto and depicts the design and verification processes applicable to BVPS-1 SPDS. The document provides an overview of the design process used for BVPS-1 SPDS and delineates, witn an asterisk (*) in the margin, the differences between the BVPS-1 SPDS design activities from those described in the generic Westinghouse SPDS WCAP.
Referring to Supplement 7 to WCAP-10170 figure 1 (page 8) illustrates the ERF Design Process and Table 1 (pages 9 & 10) identify the BV-1 SPDS documentation associated with each step of the design process.
A summary description of each process step is provided in section 2.5 of Supplement 7 to WCAP-10170.
An assessment of the SPDS design implementation as it supports the functional guidelines of Supplement 1 to NUREG-0737 must be done in context with the overall philosophy of operation and use of the SPDS in the BV-1 :ontrol room. The following discussion of the BV-1 SPDS philosophy of operation and use is provided to supplement and clarify DLC's application to the functional guidelines.
- - _ _ - - - - - - - - - - - - - - - - - J
2" S2fcty An21yaic R: port for' B'e2 Var VollGy Unit 1 Snfcty Param3tcr Dicplcy Syst;m (SPDS)
SPDS Philosophy of Operation and Use The Safety Parameter Display System (SPDS) provides personnel in the Technical Support Center (TSC), and Emergency Operating Facility (EOF) with immediate access to critical plant indications. Additionally, the SPDS provides the same concise display of critical plant variables to the control room operators, Shift Technical Advisors (STAS), and plant personnel to aid them in rapidly and reliably verifying the safety status of the plant. The SPDS, by virtue of its installation in the TSC and EOF, reduces the number of staff personnel in the control room, thereby reducing the potential for confusion in the centrol room during a plant transient or emergency condition. The principal purpose and function of the SPDS is to aid the control room personnel during an abnormal or emergency condition in determining the plant safety, status and to assess whether the abnormal condition (s) warrant corrective actions by the operators to avoid a degraded core condition. The licensed operators, STAS, and selected plant personnel are trained on SPDS operations, cognizant of the available SPDS information, and abic to interpret the information provided by the SPDS to understand the plant safety status.
The primary sources of indication for the operation of the plant are the control room indications, including post accident monitoring indication, and plant equipment. The control room indication provides the Operators with the necessary information for safe reactor operations of the plant under normal, transient, and accident conditions. The SPDS is used as an aid to enhance the control room indications. If the SPDS is not available, the control room operators are trained to mitigate the transient or emergency condition by using their control room indications and the Emergency Operating Procedures (EOPS). The control room operators are trained to respond to the transient or accident condition (s) with and without the SPDS.
The EOPs are written to mitigate the consequences of various accidents. The operation of the plant is maintained and controlled using the EOPs until the plant conditions are stabilized. With this EOP philosophy in mind, the SPDS philosophy and SPDS procedure guideline are written as an aid a..d an enhancement to the Emergency Operating Procedures during abnormal, transient, or accident conditions. The SPDS is also used during normal plant evolutions as an aid to the control room staff, but the SPDS and the SPDS procedure guidelines are not considered primary information for the operation of the plant during accident conditions.
SLfGty An0ly010 R2 port for Benv0r VcllGy Unit 1 o S2fsty P2rameter Display System (SPDS)
The primary operators of e 1. re the STAS. The STAS have the available use of the Sirs ;< . tor the plant safety status at several terminals. SPE: Crals are located in the Unit 1 Control Room, Technical Supperr Center and at the Emergency Operating Facility (EOF). In addition to the STAS, all licensed operators and selected plant personnel are trained to be able to operate, understand, and interpret the information from these SPDS terminals. SPDS users are capable of interpreting the color coding and status flags associated with the SPDS parameters. The SPDS users are trained to the level where the users are cognizant on how to use the SPDS terminals as an available reference tool to enhance existing indications. The SPDS is and will be considered an enhancament to operation of the plant and a secondary source of plant safety status information. The SPDS will provide diagrostic information during normal and accident conditions. A tlained SPDS operator, at the various locations, can aid the controi room staff by monitoring plant status changes for the plant transient or emergency conditions and provide plant safety status information to personnel outside the control room.
The SPDS operator has the ability to call up the Top Level, Map Menu, Trends, or history displays by dedicated SPDS pushbuttons.
In addition to the dedicated display pushbuttons, the SPDS can call up a display by locating the cursor in any display predefined poke field area and executing the Display Page pushbutton. This action will display on the SPDS terminal screen the associated screen display related to the poke field area.
All displays except history trends can be called upon by sequential paging, display poke fields, or entry of a specific screen page number. History trend displays can only be replayed following a reactor trip.
The SPDS screen displays are divided into six distinct categories. The initial group consists of the SPDS Map and its associated Menu screens. The Menu screens identify the first, second, third and fourth level displays. By visual observations of the SPDS menu screens, the fourth level displays can be associated with the third level displays. The fourth level displays are tabular listings of all the analog sensor inputs used in the third level displays. The last SPDS display group available is the individual sensor input computer points called the Point Detail Displays. These point detail displays are all the computer points used by the higher level displays in the SPDS.
r S foty An0lyaia R: port for B'c;v r Vollcy Unit 1 S2fcty Parametcr Dicploy Syatcm (SPDS)
There are 29 SPDS displays installed in SPDS display levels one, two and three. Level one (also known as Top Level) contains two abstract iconic displays. These two displays represent plant conditions referenced to the optimum operating points and operating limits. The two screens have eight spokes directed outward from the center of the screen. Each spoke represents a normalized plant parameter to provide an octagonal geometric shape. The symmetric octagonal diagram occurs when the active parameters are at or near the optimum operating level. As the plant parameters deviate from the optimum operating limits, the octagonal shape will distort informing the SPDS operator of the parameter deviation.
Top level display 1TL1 Narrow Range Display (NAR RNG) is structured for normal plant operation without a reactor trip condition. Top level display 1TL2 Wide Range Display (WID RNG) is used to show plant operations from full power to plant shutdown after a reactor trip. If the Narrow Range display is on the screen prior to a reactor trip, the Wide Range display will automatically be substituted for the narrow range display after initiation of the reactor trip. Both Top Level displays 1TL1 and 1TL2 are useful to display plant parameter conditions before and soon after a reactor trip. However, the iconic displays in the third and fourth Level provide more useful detail information to analyze / diagnose a plant transient or emergency conditions.
Therefore, after a reactor trip, normal operations of the SPDS would be in the third and fourth levels. Additional information can be acquired using the point detail displays if more specific information is required for the computer point in the SPOS.
In addition to the normal iconic display, Level Two contains the history trend graphs and history iconics. The history trend graphs and history iconics can ba called upon using the dedicated SPDS history pushbutton. The Iconic trends (both iconic replay and values and references vs. time) are available at all times and are not trip dependent. The Iconic histories (both iconic replat and values and references vs. time) are available after a trip. The history iconics and trend graphs provide a recorded history replay before and after the reactor trip breakers are opened of 30 minutes at 1 minuto intervals and 5 minutes at 10 second intervals. The SPDS stores the data on memory disks for the respective trend graphs or iconic history.
The SPDS procedure guideline is a reference procedure for the SPDS operator. The procedure will be written in three parts.
The first section is a reference section of the SPDS terminal operational controls, parameter color codings and parameter status flags for the SPDS operator. The second section is a grouping of SPDS parameters that are helpful in analyzing hnd diagnosing plant conditions which may be U7ed with the EOP procedural steps. A table is provided 11 ting all the SPDS parameter groupings in the second section for quick reference to operate the SPDS following the EOPs. The last section of the SPDS procedure guideline provides information that is not available in the first two sections that may be useful to the SPDS operator.
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c SCfcty Analyaio R2 port for 8env;r Vallcy Unit 1 SOfQty Parametcr Dicploy Syatim (SPDS)
The procedure guideline allows the SPDS operator to manipulate the SPDS to gather information as required to diagnose or analyze plant safety status conditions. The procedure guideline provides instructions for the SPDS operator to call up any SPDS screen displays available in the SPDS. The guideline allows the operator to follow plant conditions as specified by the EOP procedural steps and determine plant performance by viewing plant safety status as the EOP steps are accomplished.
When the SPDS is used with the EOP procedure, the EOP procedure is the driving force in controlling the SPDS. The SPDS operator will normally use the SPDS to follow plant conditions identified in the EOP steps. In addition, the SPDS operator may observe other plant status to determine the changes influenced by the EOP procedural steps. Operating the SPDS as described will provide plant personnel with information to determine plant conditions to aid the operating staff in determining the conditions the EOPs were entered.
To ensure the procedure guideline and SPDS philosophy objectives are accomplished, a procedure validation and verification program has been incorporated into the SPDS philosophy. The procedure validation and verification program ensures the procedure guideline and SPDS philosophy are impleme?ted correctly and fulfills the requirements for which they were developed. The procedure validation and verification are the last phase in the development process before the SPDS philosophy and procedure guideline are approved and recommended for use in the plant.
DLC has assessed the SPDS design to functional guidelines of Supplement 1 to NUREG 0737. The following discussions address these considerations.
- 1. The SPDS should provide a concise display of critical plant variables.
The top level displays (narrow and wide range iconics) present the critical plant variables utilizing distortion of an octagonal pattern and color coding to illustrate critical / abnormal plant conditions. These displays provide a top level assessment of plant conditions. Second level graphic displays provide the overall plant status, third level displays provide a graphic representation of plant systems and the fourth level displays provide alphanumeric format displays of sensor data. The display methodology is identical to the generic design reviewed and accepted by the NRC and has also been endorsed in Supplement 7 to WCAP-10170.
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S2fCty AnalyJia R: port for BcCvCr Vallcy Unit 1 SOfoty ParametOr Dicplcy Sy0tcm (SPDS)
- 2. The SPDS shall be located convenient to the control room operators.
Two SPDS terminals are located in the BV-1 control room. One terminal is located on the operators console and the other is located on the center section of the vertical control board with the control keyboard located on the center island console near the control board. The displays are convenient to the SPDS operators when the use is considered in the context of the overall SPDS philosophy of operation and use. All STAS, licensed operators and selected control room personnel will be trained to operate, interpret and understand the information displayed on the SPDS terminals. The STA is the primary user of the SPDS in the control room (as indicated in the philosophy) and the application of a team concept in the control room to mitigate transients illustrates the usefulness to the STA of the terminal at the operator console. The operators have a clear view of the SPDS CRT located on the conter section of the vertical control board. To clarify the use of the SPDS, the primary source of indication for the operators is the control room indicators and the operators are trained to mitigate transient or emergency conditions using these indicators in conjunction with the Emergency Operating Procedures (EOPs).
- 3. The SPDS shall continuously display information from which the safety status of the plant can be assessed.
The top level iconic displays (terminate and mitigate mode) provide a conciso display of the critical plant parameters. The lower level displays provide more useful detailed information to analyze and diagnose a plant transient or emergency condition.
Discussions with plant operators indicate that the overall plant status display (level 2) and the reactor coolant system flow diagram display (level 3) have proven to be a very useful display on the SPDS CRT located on the vertical board.
The SPDS produces 29 displays that provide control room personnel with graphic representations of plant system conditions during normal and emergency plant operations. These displays provide the SPDS user with optional means of obtaining information necessary to assess the plant safety status. Any one of these 29 displays will be administratively maintained on the SPDS screen in the control room while the SPDS is in service.
S2fCty Annlysis R port for Bc; var Vallcy Unit 1 Safety Parameter Display System (SPDS) s
- 4. The SPDS should aid the operators in rapidly and reliably determining the safety status of the plant.
Validation testing of the SPDS system was performed as a combined factory / site acceptance test procedure which included hardware diagnostic testing, man-machine interface tests, (i.e.: data update and response time) input processing performanco, and SPDS algorithm and display coding verification.
Digital input indication was verified relative to device position by simulated testing and by exercising the actual field device.
Analog input verification tests were conducted to verify process loop integrity after tic-in of the computer. Testing was conducted to determine that the SPDS tie-in did not degrado existing Plant Variable Computer indication. SPDS analog input accuracy testing was conducted to verify the I/O cabinet input with the SPDS display and to verify proper indication scaling.
The SPDS display indication was also subjected to an accuracy verification with the corresponding control room indication.
Other miscellaneous SPDS testing included ARTEL fiber optic verification ERF-BVPS data multiplexor testing and Avanti repeator/ driver verification.
The test specification for the SPDS computer specification number 8700-DES-004, Revision 1 defined the engineering test requirements and acceptance criteris and is the basin for the test procedures. The specification requirements for the following parameters are: update rate (2 seconds), display refresh rate (60Hz), availability (greater than 99% in the control room), and response times (2 second background start and 10 second maximum to build a display).
The satisfactory performance of the test procedures and test results verified that the SPDS met the specification requirements. All open items which were identified during testing have been resolved and DLC considers the functional guidelines of Supplement 1 to NUREG 0737 to be satisfied.
The SPDS designed availability estimate is greater than 99%. An administrative procedure will be developed and implemented prior to start-up following the seventh BV-1 refueling outage to monitor the SPDS unavailability. Computer downtime will be tracked using a problem log which will be reviewed for trending purposes to reduce downtime and improve overall system availability.
The SPDS is designed such that the processing of any variable shall not increase the error of the displayed value by more than 1% of the span of that variable. The accuracy of the SPDS process combined with the accuracy of the analog indication yields an overall accuracy of the SPDS displayed variable of 2%.
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Safcty Analisic Raport for B3avar Vclley Unit 1 Safety Parameter Display System (SPDS)
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The SPOS displayed information will be periodically verified primarily by performing checks with other qualified indication.
A program to accomplish these verification checks will be implemented prior to start-up following the seventh refueling outage.
Data validity on the SPDS displays is illustrated by four quality attributes associated with each variable. The qualities are
,d, Manual, Poor and Bad. Manual data will indicate that the C ae has been manually entered into the data base rather than i . ..m a scanned sensor. Poor data will be used to indicate that one or more sensors of a redundant or diverse set of sensors are no longer Good. Bad data will be used to indicate that a sensor value iF either missing or detected by the system as resulting from failed input devices. Data quality will be retained as an attribute of any calculated variables and will be determined based on the quality of the individual inputs to that calculation.
The SPDS also utilizes the following color convention in conjuction with the qua ity codes: yellow is used to represent normal conditions, magents to represent suspect (bad, poor, or manually substituted) values and red to represent alarm conditions.
This color convention is the standard applied to the generic W SPDS; however, when the SPDS was evaluated during the BV-1 control room design review an inconsistent color 'oplication was identified and documented as a human engineering ) ,repancy. As a result, the SPDS color application will be ch. 7d and green will be substituted for yellow where yellow is t x1 to depict normal conditions and yellow will be substitutea for magenta where magenta is used to depict suspect conditions. This change will be subjected to the same design process as the delivered system and the change is scheduled to be implemented prior to start-up following the seventh BV-1 refueling outage.
Security is maintained on the SPDS through several means. The SPDS software can only be modified from the programmers console using a number of sequential steps. These steps involve unloading the operating disk, loading the source code disk, performing a system rebuild to include the modifications, and transferring the rebuilt system to the operating disk. The SPDS database can have approved online changes performed. These online changes are temporary since they are not part of the database master disk. Modification of the database master disk l requires an approved database change i. em and several steps similar to the above. Procedures for software and database modifications are covered in the operating manual and I&C manual.
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3 SOfsty Analycia Report-for 53cvar Valley Unit 1 Safcty Paramator Display System-(SPDS)
- 5. The SPDS shall be suitably isolated from electrical and electronic interference with equipment and sensors that are in use for safety systems.
In a letter dated April 9, 1986, the NRC requested additional information regarding the isolation devices utilized with the BV-1 Safety Paramci.er Display Systems. In a letter dated February 17, 1987, DLC submitted a response to this request for addjeional' information and to date no further information on the sub,- c has been requested.
- 6. The SPDS shall be designed to incorporate accepted human factors principles so that displayed information can be readily perceived and comprehended by the SPDS users.
Section 2.2.1(b) of Supplement 7 of WCAP-10170 describes the human factors engineering process performed on the SPDS design.
In addition, a human factors review of the SPDS was performed using the guidelines of Section 6.7, Process Computers, of NUREG-0700. The review has been conducted and the HEDs will be processed in accordance with the BVPS-1 Detailed Control Room Design RevinW (DCRDR) Program. The details of the SPDS human factors review will be provided in a DCRDR Supplemental Report which is currently scheduled to be issued on November 18, 1988.
A summary of ene results is provided below.
Total Number of 0700 Guidelines in Section 6.7 224 Number Complied With 124 Number Not Complied With 19 Number Remaining To Be Checked 7 Number Determined To Be Not Applicable 74 The more significant discrepancies address the abbreviations used in the screen text, no CRT displays of the files being processed and no storage of sequential files of operator entries, the use of color codes and the CRT location on the vertical board. All the items to be checked except one pertain to measurements of the screen luminance in the control room.
Additionally, the BVPS-2 SPDS Computer Survey results were reviewed in order to determine a consistent application of the NUREG-0700 criteria. Finally, 30 of the guidelines which were determined to be "not applicable" are included in the Printer, Alarm Messages, and Graphs & Tables Requirements Sections of Section 6.7 of NUREG 0700. These guidelines were determined to be "not applicable" because the SPDS utilizes a video copier and because the SPDS does not function as an alarm printer for the Annuniciator System.
Snfsty Annly3ic Rsport for B3nvar Valley Unit 1 Snfoty Param3 tor Display Systcm (SPDS)
- 7. The SPDS should display critical plant variables.
The parameters displayed on the SPDS are required to provide the operator with sufficient information regarding the following five critical safety functions identified in NUREG 0696 and in section 4.1.f of Supplement 1 to NUREG 0737.
I. Reactivity Control II. Reactor Core Cooling anu Heat Removal from the Primary System III. Reactor Coolant System Integrity IV. Radioactivity Control V. Containment Conditions Westinghouse and DLC selected those plant specific parameters necessary to evaluate the critical safety functions listed above.
The parameters used in the two top level iconic displays (terminate and mitigate mode) have been placed in these five safety functions and are listed in Tables C.1 & C.2 of Supplement 7 to WCAP-10170. Table C.3 maps the individual system displays into the five safety function categories and Table C.4 maps the individual plant parameters into the five safety function categories.
- 8. Procedures which describe the timely and correct safety status assessment when the SPDS is and is not available will be developed by tne licensee in parallel with the SPDS.
Furthermore, operators should be trained to respond to accident conditions both with and without the SPDS availcble.
The SPDS philosophy identifies the principle users of the SPDS to be the STAS. Additionally, all STAS, licensed operators and selected control room personnel are trained in the use of the SPDS. A procedure guideline is being developed as a reference for the SPDS operator to instruct the user on the use and effective manipulation of the system. With this procedure guideline, the SPDS aids plant operatians and emergency personnel in monitoring and assessing plant safety status, however, the
. guideline maintains that the primary sources of indication for the operation of the plant are control room and control board indication and plant equipment. If the SPDS is not available, the control room crew is already trained to mitigate the transient or emergency condition by using the control room indications, normal operating procedures and Emergency Operating Procedures. The SPDS philosophy and the procedure guideline verification and validation process are conducted to determine that the philosophy and guideline are correctly implemented to l demonstrate that the SPDS fulfills the purpose for which it is intended. SPDS training to this philosophy and procedure guideline will be completed prior to start-up following the seventh BV-1 refueling outage.
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