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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217D5211999-09-30030 September 1999 Informs That Remediating 3D Monicore Sys at Pbaps,Units 2 & 3 & 3D Monicore/Plant Monitoring Sys at Lgs,Unit 2 Has Been Completed Ahead of Schedule ML20216J3981999-09-29029 September 1999 Submits Comments for Lgs,Unit 1 & Pbaps,Units 2 & 3 Rvid,Rev 2,based on Review as Requested in GL 92-01,rev 1,suppl 1, Reactor Vessel Structural Integrity ML20212J6561999-09-29029 September 1999 Informs of Completion of mid-cycle PPR of Limerick Generating Station on 990913.Identified No Areas in Which Licensee Performance Warranted Addl Insp Beyond Core Insp Program.Historical Listing of Plant Issues Encl ML20212H6401999-09-24024 September 1999 Forwards Revised Epips,Including Rev 11 to ERP-101 & Rev 18 to ERP-800.Copy of Computer Generated Rept Index Identifying Latest Revs of LGS Erps,Encl ML20212E8081999-09-22022 September 1999 Provides Notification That Listed Operators Have Been Permanently Reassigned to Duties That Do Not Require Maintaining Licensed Operator Status,Per 10CFR50.74 ML20212E7941999-09-22022 September 1999 Requests Authorization for Listed Licensed Operators to Temporarily Suspend Participation in Licensed Operator Requalification Program at LGS ML20212F5481999-09-20020 September 1999 Forwards Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing, for Pbaps,Units 2 & 3 & Lgs,Units 1 & 2 ML20212F8991999-09-17017 September 1999 Provides Written Confirmation That Thermo-Lag 330-1 Fire Barrier Corrective Actions at Lgs,Units 1 & 2 Have Been Completed ML20216F7821999-09-16016 September 1999 Forwards Insp Repts 50-352/99-05 & 50-353/99-05 on 990713-0816.One Violation Noted & Being Treated as NCV, Consistent with App C of Enforcement Policy.Violation Re Inoperability of Automatic Depression Sys During Maint ML20212A8751999-09-13013 September 1999 Forwards Safety Evaluation of First & Second 10-year Interval Inservice Insp Plan Request for Relief ML20212A0091999-09-0909 September 1999 Provides Notification That Licenses SOP-11172 & SOP-11321, for SO Muntzenberger & Rh Wright,Respectively,Are No Longer Necessary as Result of Permanent Reassignment ML20211N5061999-09-0909 September 1999 Forwards TSs Bases Pages B 3/4 10-2 & B 3/4 2-4 for LGS, Units 1 & 2,being Issued to Assure Distribution of Revised Bases Pages to All Holders of TSs ML20211P8571999-09-0808 September 1999 Forwards Reactor Operator Retake Exams 50-352/99-303OL & 50-353/99-303OL Conducted on 990812 ML20211P3891999-09-0303 September 1999 Informs That During 990902 Telcon Between J Williams & B Tracy,Arrangements Were Made for NRC to Inspect Licensed Operator Requalification Program at Plant.Insp Planned for Wk of 991018 ML20211H2571999-08-26026 August 1999 Informs of Individual Exam Result on Initial Retake Exam on 990812.One Individual Was Administered Exam & Passed ML20211E9191999-08-24024 August 1999 Forwards fitness-for-duty Program Performance Data for Jan-June 1999 for PBAPS & LGS IAW 10CFR26.71(d).Data Includes Listed Info ML20211E9731999-08-23023 August 1999 Forwards LGS Unit 2 Summary Rept for 970228 to 990525 Periodic ISI Rept Number 5, Per TS SRs 4.0.5 & 10CFR50.55a(g) ML20211D6761999-08-20020 August 1999 Forwards non-proprietary Revised Emergency Response Procedures (Erps),Including Rev 29 to ERP-110, Emergency Notification & Rev 17 to ERP-800, Maint Team & Proprietary App ERP-110-1.App Withheld Per 10CFR2.790(a)(6) ML20210T4271999-08-13013 August 1999 Informs That NRC Revised Info in Rvid & Releasing Rvid Version 2 as Result of Review of 980830 Responses to GL 92-01 Rev 1,GL 92-01 Rev 1 Suppl 1 & Suppl Rai.Tacs MA1197 & MA1198 Closed ML20211B7881999-08-10010 August 1999 Transmits Summary of Two Meetings with Risk-Informed TS Task Force in Rockville,Md on 990514 & 0714 05000353/LER-1999-005, Forwards LER 99-005-00,re Actuation of Primary Containment & Reactor Vessel Isolation Control Sys,Esf.Fuse Failed Due to Mechanical Failure of Cold Solder Joint1999-08-10010 August 1999 Forwards LER 99-005-00,re Actuation of Primary Containment & Reactor Vessel Isolation Control Sys,Esf.Fuse Failed Due to Mechanical Failure of Cold Solder Joint ML20210U2211999-08-10010 August 1999 Forwards Insp Repts 50-352/99-04 & 50-353/99-04 on 990525-0712.One Violation Occurred & Being Treated as NCV, Consistent with App C of Enforcement Policy.Violation Re Late Performance of off-gas Grab Sample Surveillance ML20210P4191999-08-0404 August 1999 Forwards Initial Exam Repts 50-352/99-302 & 50-353/99-302 on 990702-04 (Administration) & 990715-22 (Grading).Six of Limited SRO Applicants Passed All Portion of Exam ML20210M7571999-08-0404 August 1999 Forwards Response to Requesting Addl Info Re Status of Decommissioning Funding for Lgs,Pbaps & Sngs. Attachment Provides Restatement of Questions Followed by Response NUREG-1092, Informs J Armstrong of Individual Exam Results for Applicants on Initial Exam Conducted on 990702 & 990712-14 at Facility.All Six Individuals Who Were Administered Exam, Passed Exam.Without Encls1999-08-0303 August 1999 Informs J Armstrong of Individual Exam Results for Applicants on Initial Exam Conducted on 990702 & 990712-14 at Facility.All Six Individuals Who Were Administered Exam, Passed Exam.Without Encls ML20210L2011999-07-28028 July 1999 Forwards Final Personal Qualification Statement (NRC Form 398) for Reactor Operator License Candidate LB Mchugh ML20211F2641999-07-27027 July 1999 Forwards Three Copies of Rev 12 to LGS Physical Security Plan, Rev 4 to LGS Training & Qualification Plan & Rev 2 to LGS Safeguards Contingency Plan. Without Encls 05000353/LER-1999-004, Forwards LER 99-004-00 Re 990701 Discovery of Pressure Setpoint Drift of Thirteen Mss SRV Due to Corrosion Induced Bonding within SRVs1999-07-23023 July 1999 Forwards LER 99-004-00 Re 990701 Discovery of Pressure Setpoint Drift of Thirteen Mss SRV Due to Corrosion Induced Bonding within SRVs ML20210E6211999-07-22022 July 1999 Submits Rev to non-limiting Licensing Basis LOCA Peak Clad Temps (Pcts) for Limerick Generating Station (Lgs),Units 1 & 2 & Pbaps,Units 2 & 3 ML20216D3081999-07-19019 July 1999 Requests Renewal of OLs for Listed Individuals,Iaw 10CFR55.57.NRC Forms 398 & 396,encl for Applicants.Without Encl ML20216D8041999-07-19019 July 1999 Submits Summary of Final PECO Nuclear Actions Taken to Resolve Scram Solenoid Pilot Valve Issues Identified in Info Notice 96-007 ML20209G9121999-07-0909 July 1999 Informs That Ja Hutton Has Been Appointed Director,Licensing for PECO Nuclear,Effective 990715.Previous Correspondence Addressed to Gd Edwards Should Now Be Sent to Ja Hutton ML20209F6341999-07-0909 July 1999 Submits Supplemental Response to GL 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Bwrs, for Unit 2.Rev 0 to 1H61R & GE-NE-B13-02010-33NP Repts & Revised Pages to Summary Rept Previously Submitted,Encl ML20210B4441999-07-0808 July 1999 Forwards Preliminary NRC Form 398 & NRC Form 396 for Reactor Operator for License Candidate LB Mchugh.Candidate Failed Category B Portion of Operating Exam Given at LGS During Week of 990315.Tentative re-exam Has Been Scheduled 990812 ML20209C9041999-07-0808 July 1999 Forwards Monthly Operating Repts for June 1999 for Limerick Generating Station,Units 1 & 2 & Revised Monthly Repts for May 1999 ML20209D8821999-07-0707 July 1999 Submits Estimate of Number of Licensing Actions Expected to Be Submitted in Years 2000 & 2001,as Requested by Administrative Ltr 99-02.Renewal Applications for PBAPS, Units 2 & 3,will Be Submitted in Second Half of 2001 05000353/LER-1999-003, Forwards LER 99-003-00,re Bypass of RW Cleanup Leak Detection Sys Isolation Function on Three Separate Occasions.Bypass of Safety Function Was Caused by Inadequate Review & Approval of Change to Procedure1999-07-0707 July 1999 Forwards LER 99-003-00,re Bypass of RW Cleanup Leak Detection Sys Isolation Function on Three Separate Occasions.Bypass of Safety Function Was Caused by Inadequate Review & Approval of Change to Procedure ML20209D2671999-07-0202 July 1999 Responds to NRC 990322 & 0420 RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves 05000352/LER-1999-004, Forwards LER 99-004-00,re Inoperability of Automatic Depressurization Sys Portion of Eccs.Condition Resulted from Incomplete Impact Review of Isolating Portion of ADS Nitrogen Backup Supply on Operability of ECCS Sys1999-07-0101 July 1999 Forwards LER 99-004-00,re Inoperability of Automatic Depressurization Sys Portion of Eccs.Condition Resulted from Incomplete Impact Review of Isolating Portion of ADS Nitrogen Backup Supply on Operability of ECCS Sys ML20196J6301999-07-0101 July 1999 Requests Addl Info Re Status of Decommissioning Funding for Limerick Generating Station,Units 1 & 2,Peach Bottom Atomic Power Station,Units 1,2 & 3 & Salem Nuclear Generating Station,Units 1 & 2 ML20209B7001999-06-30030 June 1999 Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20212J5401999-06-28028 June 1999 Discusses Completion of Licensing Action for NRC Bulletin 96-003, Potential Plugging of ECC Suction Strainers by Debris in Bwrs. Bulletin Closed for Unit 2 by NRC ML20207H8271999-06-24024 June 1999 Informs NRC That Util Has Completed Core Shroud Insps for LGS Unit 2.Proprietary Rept GE-NE-B13-02010-33P & non-proprietary Rev 0 to 1H61R,encl.Proprietary Rept Withheld,Per 10CFR2.790(a)(4) ML20196G7041999-06-24024 June 1999 Forwards Insp Repts 50-352/99-03 & 50-353/99-03 on 990413- 0524.No Violations Noted.Nrc Concluded That Licensee Staff Continued to Operate Both Units Safely ML20196A5641999-06-15015 June 1999 Provides Info Re Util Use of Four Previously Irradiated LGS, Unit 1,GE11 Assemblies in Unit 2 Cycle 6.Encl 990518 GE Ltr Provides Objective of Lead Use Assemblies Program & Outlines Kinds of Measurements That Will Be Made on Assemblies ML20195J6831999-06-11011 June 1999 Provides Proprietary Objectives for Lgs,Units 1 & 2,1999 Emergency Preparedness Exercise Scheduled to Be Conducted on 990914.Licensee Identifies Which Individuals Should Receive Copies of Info.Proprietary Info Withheld ML20195G4591999-06-10010 June 1999 Forwards MORs for May 1999 & Revised Repts for Apr 1999 for LGS Units 1 & 2 ML20195H0531999-06-0909 June 1999 Forwards Revised Bases Pages B3/4 10-2 & B3/4 2-4 for LGS Units 1 & 2,in Order to Clarify That Requirements for Reactor Enclosure Secondary Containment Apply to Extended Area Encompassing Both Reactor Enclosure & Refueling Area ML20195E7701999-06-0707 June 1999 Provides Notification of Change to NPDES Permit PA0052221, for Bradshaw Reservoir Facility Which Supports Operation of Lgs,Units 1 & 2,per EPP Section 3.2 ML20195C7631999-06-0101 June 1999 Notifies NRC That PECO Energy Has Completed Installation of New Large Capacity,Passive Strainers on RHR & Core Spray Sys Pump Suction Lines at Lgs,Unit 2,in Response to Ieb 96-003 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217D5211999-09-30030 September 1999 Informs That Remediating 3D Monicore Sys at Pbaps,Units 2 & 3 & 3D Monicore/Plant Monitoring Sys at Lgs,Unit 2 Has Been Completed Ahead of Schedule ML20216J3981999-09-29029 September 1999 Submits Comments for Lgs,Unit 1 & Pbaps,Units 2 & 3 Rvid,Rev 2,based on Review as Requested in GL 92-01,rev 1,suppl 1, Reactor Vessel Structural Integrity ML20212H6401999-09-24024 September 1999 Forwards Revised Epips,Including Rev 11 to ERP-101 & Rev 18 to ERP-800.Copy of Computer Generated Rept Index Identifying Latest Revs of LGS Erps,Encl ML20212E7941999-09-22022 September 1999 Requests Authorization for Listed Licensed Operators to Temporarily Suspend Participation in Licensed Operator Requalification Program at LGS ML20212E8081999-09-22022 September 1999 Provides Notification That Listed Operators Have Been Permanently Reassigned to Duties That Do Not Require Maintaining Licensed Operator Status,Per 10CFR50.74 ML20212F5481999-09-20020 September 1999 Forwards Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing, for Pbaps,Units 2 & 3 & Lgs,Units 1 & 2 ML20212F8991999-09-17017 September 1999 Provides Written Confirmation That Thermo-Lag 330-1 Fire Barrier Corrective Actions at Lgs,Units 1 & 2 Have Been Completed ML20212A0091999-09-0909 September 1999 Provides Notification That Licenses SOP-11172 & SOP-11321, for SO Muntzenberger & Rh Wright,Respectively,Are No Longer Necessary as Result of Permanent Reassignment ML20211E9191999-08-24024 August 1999 Forwards fitness-for-duty Program Performance Data for Jan-June 1999 for PBAPS & LGS IAW 10CFR26.71(d).Data Includes Listed Info ML20211E9731999-08-23023 August 1999 Forwards LGS Unit 2 Summary Rept for 970228 to 990525 Periodic ISI Rept Number 5, Per TS SRs 4.0.5 & 10CFR50.55a(g) ML20211D6761999-08-20020 August 1999 Forwards non-proprietary Revised Emergency Response Procedures (Erps),Including Rev 29 to ERP-110, Emergency Notification & Rev 17 to ERP-800, Maint Team & Proprietary App ERP-110-1.App Withheld Per 10CFR2.790(a)(6) 05000353/LER-1999-005, Forwards LER 99-005-00,re Actuation of Primary Containment & Reactor Vessel Isolation Control Sys,Esf.Fuse Failed Due to Mechanical Failure of Cold Solder Joint1999-08-10010 August 1999 Forwards LER 99-005-00,re Actuation of Primary Containment & Reactor Vessel Isolation Control Sys,Esf.Fuse Failed Due to Mechanical Failure of Cold Solder Joint ML20210M7571999-08-0404 August 1999 Forwards Response to Requesting Addl Info Re Status of Decommissioning Funding for Lgs,Pbaps & Sngs. Attachment Provides Restatement of Questions Followed by Response ML20210L2011999-07-28028 July 1999 Forwards Final Personal Qualification Statement (NRC Form 398) for Reactor Operator License Candidate LB Mchugh ML20211F2641999-07-27027 July 1999 Forwards Three Copies of Rev 12 to LGS Physical Security Plan, Rev 4 to LGS Training & Qualification Plan & Rev 2 to LGS Safeguards Contingency Plan. Without Encls 05000353/LER-1999-004, Forwards LER 99-004-00 Re 990701 Discovery of Pressure Setpoint Drift of Thirteen Mss SRV Due to Corrosion Induced Bonding within SRVs1999-07-23023 July 1999 Forwards LER 99-004-00 Re 990701 Discovery of Pressure Setpoint Drift of Thirteen Mss SRV Due to Corrosion Induced Bonding within SRVs ML20210E6211999-07-22022 July 1999 Submits Rev to non-limiting Licensing Basis LOCA Peak Clad Temps (Pcts) for Limerick Generating Station (Lgs),Units 1 & 2 & Pbaps,Units 2 & 3 ML20216D3081999-07-19019 July 1999 Requests Renewal of OLs for Listed Individuals,Iaw 10CFR55.57.NRC Forms 398 & 396,encl for Applicants.Without Encl ML20216D8041999-07-19019 July 1999 Submits Summary of Final PECO Nuclear Actions Taken to Resolve Scram Solenoid Pilot Valve Issues Identified in Info Notice 96-007 ML20209F6341999-07-0909 July 1999 Submits Supplemental Response to GL 94-03, Intergranular Stress Corrosion Cracking of Core Shrouds in Bwrs, for Unit 2.Rev 0 to 1H61R & GE-NE-B13-02010-33NP Repts & Revised Pages to Summary Rept Previously Submitted,Encl ML20209G9121999-07-0909 July 1999 Informs That Ja Hutton Has Been Appointed Director,Licensing for PECO Nuclear,Effective 990715.Previous Correspondence Addressed to Gd Edwards Should Now Be Sent to Ja Hutton ML20209C9041999-07-0808 July 1999 Forwards Monthly Operating Repts for June 1999 for Limerick Generating Station,Units 1 & 2 & Revised Monthly Repts for May 1999 ML20210B4441999-07-0808 July 1999 Forwards Preliminary NRC Form 398 & NRC Form 396 for Reactor Operator for License Candidate LB Mchugh.Candidate Failed Category B Portion of Operating Exam Given at LGS During Week of 990315.Tentative re-exam Has Been Scheduled 990812 05000353/LER-1999-003, Forwards LER 99-003-00,re Bypass of RW Cleanup Leak Detection Sys Isolation Function on Three Separate Occasions.Bypass of Safety Function Was Caused by Inadequate Review & Approval of Change to Procedure1999-07-0707 July 1999 Forwards LER 99-003-00,re Bypass of RW Cleanup Leak Detection Sys Isolation Function on Three Separate Occasions.Bypass of Safety Function Was Caused by Inadequate Review & Approval of Change to Procedure ML20209D8821999-07-0707 July 1999 Submits Estimate of Number of Licensing Actions Expected to Be Submitted in Years 2000 & 2001,as Requested by Administrative Ltr 99-02.Renewal Applications for PBAPS, Units 2 & 3,will Be Submitted in Second Half of 2001 ML20209D2671999-07-0202 July 1999 Responds to NRC 990322 & 0420 RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves 05000352/LER-1999-004, Forwards LER 99-004-00,re Inoperability of Automatic Depressurization Sys Portion of Eccs.Condition Resulted from Incomplete Impact Review of Isolating Portion of ADS Nitrogen Backup Supply on Operability of ECCS Sys1999-07-0101 July 1999 Forwards LER 99-004-00,re Inoperability of Automatic Depressurization Sys Portion of Eccs.Condition Resulted from Incomplete Impact Review of Isolating Portion of ADS Nitrogen Backup Supply on Operability of ECCS Sys ML20209B7001999-06-30030 June 1999 Responds to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20207H8271999-06-24024 June 1999 Informs NRC That Util Has Completed Core Shroud Insps for LGS Unit 2.Proprietary Rept GE-NE-B13-02010-33P & non-proprietary Rev 0 to 1H61R,encl.Proprietary Rept Withheld,Per 10CFR2.790(a)(4) ML20196A5641999-06-15015 June 1999 Provides Info Re Util Use of Four Previously Irradiated LGS, Unit 1,GE11 Assemblies in Unit 2 Cycle 6.Encl 990518 GE Ltr Provides Objective of Lead Use Assemblies Program & Outlines Kinds of Measurements That Will Be Made on Assemblies ML20195J6831999-06-11011 June 1999 Provides Proprietary Objectives for Lgs,Units 1 & 2,1999 Emergency Preparedness Exercise Scheduled to Be Conducted on 990914.Licensee Identifies Which Individuals Should Receive Copies of Info.Proprietary Info Withheld ML20195G4591999-06-10010 June 1999 Forwards MORs for May 1999 & Revised Repts for Apr 1999 for LGS Units 1 & 2 ML20195H0531999-06-0909 June 1999 Forwards Revised Bases Pages B3/4 10-2 & B3/4 2-4 for LGS Units 1 & 2,in Order to Clarify That Requirements for Reactor Enclosure Secondary Containment Apply to Extended Area Encompassing Both Reactor Enclosure & Refueling Area ML20195E7701999-06-0707 June 1999 Provides Notification of Change to NPDES Permit PA0052221, for Bradshaw Reservoir Facility Which Supports Operation of Lgs,Units 1 & 2,per EPP Section 3.2 ML20195C7631999-06-0101 June 1999 Notifies NRC That PECO Energy Has Completed Installation of New Large Capacity,Passive Strainers on RHR & Core Spray Sys Pump Suction Lines at Lgs,Unit 2,in Response to Ieb 96-003 ML20195D5381999-05-26026 May 1999 Forwards 1998 Occupational Exposure Tabulation Rept for LGS Units 1 & 2. Encl Is Diskette & Instructions.Rept Is Being re-submitted to Reset 12 Month Time Period.Without Disk ML20195B2821999-05-24024 May 1999 Requests That NRC Distribution Lists for LGS Be Updated. Marked-up Distribution List Showing Changes Is Attached ML20196L2891999-05-20020 May 1999 Provides Status Update of Thermo-Lag 330-1 Fire Barrier Corrective Actions,Iaw Commitments Made in ML20195B2951999-05-20020 May 1999 Forwards Rev 0 to LGS Unit 2 Reload 5,Cycle 6 COLR, IAW TS Section 6.9.1.12.Values Listed Have Been Determined Using NRC-approved Methodology & Are Established Such That All Applicable Limits of Plants Safety Analysis Are Met 05000352/LER-1999-003, Forwards LER 99-003-00,re Rps,Pcrvics Actuations.Ler Contains Special Rept Info for HPCI & Reactor Core Isolation Cooling Sys Injections Into Rv1999-05-19019 May 1999 Forwards LER 99-003-00,re Rps,Pcrvics Actuations.Ler Contains Special Rept Info for HPCI & Reactor Core Isolation Cooling Sys Injections Into Rv 05000353/LER-1999-002, Forwards LER 99-002-00,automatic Actuations of Primary Containment & Reactor Vessel Isolation Control Sys & Other Common Plant ESF Due to Loss of Power to a Rps/Ups Power Distribution Panel on 9904191999-05-18018 May 1999 Forwards LER 99-002-00,automatic Actuations of Primary Containment & Reactor Vessel Isolation Control Sys & Other Common Plant ESF Due to Loss of Power to a Rps/Ups Power Distribution Panel on 990419 ML20206E2001999-04-28028 April 1999 Forwards 1998 Annual Environ Operating Rept (Non- Radiological) for Limerick Generating Station,Units 1 & 2. Rept Submitted IAW Section 5.4.1 of App B of Fols,Epp (Non- Radiological) & Describes Implementation of EPP for 1998 ML20206D8801999-04-27027 April 1999 Forwards Rev 2 to LGS Unit 1 Reload 7,Cycle 8 COLR, IAW TS Section 6.9.1.12.COLR Provides cycle-specific Parameter Limits for Noted Info ML20206A5461999-04-21021 April 1999 Responds to Conference Call Between Util & NRC on 990420,re TS Change Request 98-07-2,revising TS Section 2.0 to Incorporate Revised MCPR Safety Limits.Attached Ltr Contains Info Requested ML20205T0441999-04-17017 April 1999 Forwards 1998 Annual Radiological Environ Operating Rept 15, IAW TS Section 6.9.1.7.REMP for 1998,confirmed That LGS Environ Effects from Radioactive Release Were Well Below LGS TSs & Other Applicable Regulatory Limits ML20205Q7581999-04-15015 April 1999 Forwards Response to RAI Re ISI Program First & Second 10-Yr Interval Relief Requests.Revs to Identified by Vertical Bar in Right Margin ML20205Q4821999-04-12012 April 1999 Requests Change to Co Which Modified Licenses NPF-39 & NPF-85 to Incorporate Commitment to Complete Implementation of Thermo-Lag 330-1 Fire Barrier Corrective Actions by 990930 ML20205F8981999-03-31031 March 1999 Provides Info Re Status of Decommissioning Funding for LGS, Units 1 & 2,PBAPS,Units 1,2 & 3 & Sgs,Units 1 & 2,per Requirements of 10CFR50.75(f)(1) ML20205J9821999-03-31031 March 1999 Forwards Rev 26 & 27 of ERP-110 with Sensitive Info Excluded & Insert Sheet Indicating That Certain Info Was Intentionally Omitted ML20204D0431999-03-16016 March 1999 Forwards Safeguards Event Rept 99-001-00,providing 30 Day Written Rept Re Compromise of Safeguards Info in That Old Rev of LGS PSP Was Discovered on Hard Drive of self- Contained Computer Sys within Security Ofc Area 1999-09-09
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059K3671990-09-14014 September 1990 Informs of Revised Commitments Re Crud Induced Localized Corrosion Related to Fuel Cladding Failures.Deep Bed Demineralizers Installation Activities Will Be Performed in Unit 1 Subsequent to Third Refueling Outage ML20065D4421990-09-14014 September 1990 Responds to Generic Ltr 90-07, Operator Licensing Natl Exam Schedule. Proposed Schedules for Operator Licensing Exams, Requalification Exams & Generic Fundamental Exams Encl ML20064A5831990-09-0707 September 1990 Responds to Violations Noted in Insp Repts 50-352/90-17 & 50-353/90-16 Re Differential Pressure for Pumps.Corrective Actions:Licensee Will No Longer Use Expanded Ranges as Acceptance Criteria for Inservice Testing Program Tests ML20064A4821990-08-31031 August 1990 Forwards Rev 20 to Emergency Plan.Changes Necessitated by Annual Emergency Plan Update & Administrative in Nature ML20059E6071990-08-29029 August 1990 Forwards Semiannual Effluent Release Rept,Jan-June 1990 & Rev 8 to Odcm ML20064A6471990-08-24024 August 1990 Forwards Public Version of Revised Epips,Consisting of Rev 10 to EP-101,Rev 2 to EP-112,Rev 13 to EP-208,Rev 11 to EP-230 & Rev 22 to EP-291 ML20059E9861990-08-24024 August 1990 Provides Justification for Applicability of Reload Methodology Topical Repts to Facility & Requests NRC Approval for Application of Reload Analysis Methodologies ML20059B0751990-08-24024 August 1990 Forwards Rev 0 to Updated FSAR for Limerick Generating Station,Units 1 & 2,Vols 1-19.W/one Oversize Encl. Proprietary Vol 7A (App 3B) Withheld (Ref 10CFR2.790) ML20058N9591990-08-13013 August 1990 Forwards Revised Response to Violations Noted in Insp Repts 50-352/90-13 & 50-353/90-13.Corrective Actions:Ltr Issued to All Plant Personnel Providing Instructions on Proper Use & Handling of Controlled Documents in Controlled Locations ML20058N1771990-08-10010 August 1990 Responds to NRC Re Unresolved Items Noted in Insp Repts 50-352/90-80 & 50-353/90-80.Plant-specific Technical Guideline Has Been Revised to Ref Contingency Numbers Rather than Transient Response Implementation Plan Procedures ML20063P9461990-08-10010 August 1990 Provides Plans for Ultimate Disposition of Recirculation Inlet Nozzle to Safe End Weld Indication.Alternative Corrective Actions to Disposition Nozzle to Safe End Weld Indication Include Repair by Weld Overlay W/O Monitoring ML20058N1281990-08-0909 August 1990 Forwards Correction to Rev 10 to EPIP EP-234, Obtaining Containment Gas Samples from Containment Leak Detector During Emergencies ML20058N1991990-08-0909 August 1990 Advises of Change of Address for Correspondence Re Util Operations.All Incoming Correspondence Must Be Directed to One of Listed Addresses ML20058P1261990-08-0909 August 1990 Forwards Monthly Operating Repts for Jul 1990 for Limerick Units 1 & 2 & Rev 1 to June 1990 Rept ML20058M9951990-08-0808 August 1990 Responds to NRC Re Violations Noted in Insp Repts 50-352/90-15 & 50-353/90-14.Corrective Actions:Personnel Counseled on Importance of Procedure Compliance & Operations Manual Revised ML18095A3761990-07-26026 July 1990 Forwards Decommissioning Repts & Certification of Financial Assurance for Plants ML20055J0241990-07-26026 July 1990 Forwards Response to NRC Regulatory Effectiveness Review Rept for Plant.Response Withheld Per 10CFR73.21 ML20056A9731990-07-25025 July 1990 Forwards Facility Written Exam Comments for NRC Insp Repts 50-352/90-10 & 50-353/90-11.Written Exam for Reactor Operator & Senior Reactor Operator Considered Comprehensive & Thorough ML20055H8511990-07-24024 July 1990 Responds to NRC 900720 Request for Addl Info Re Util 900516 Request for Exemption from Full Participation During 1990 Onsite/Offsite Emergency Exercise.Nrc Region I & FEMA Support Feb 1991 Exercise,Per 900718 Telcon ML20055H8331990-07-20020 July 1990 Submits Change of Addresses for Correspondence Re Util Nuclear Operations ML20055H0231990-07-12012 July 1990 Forwards Public Version of Revised Epips,Including Rev 10 to EP-210,Rev 19 to EP-231 & Rev 13 to EP-237 ML20044A1041990-06-22022 June 1990 Forwards Application for Amends to Licenses NPF-39 & NPF-85, Consisting of Tech Spec Change Requests 90-03-0 & 90-04-0, Revising Surveillance Requirement 4.9.6.1 for Section 3.9.6 Refueling Platform Re Main Hoists/Auxiliary Hoists ML20043J0371990-06-20020 June 1990 Forwards Description,Scope,Objectives for Plant 1990 Annual Emergency Exercise Scheduled for 900920,per 890809 Ltr.Util Will Submit Revised Objectives for Exercise to Reflect Limited Participation,If Exemption Request Approved ML20043H6081990-06-19019 June 1990 Corrects 900427 Response to Generic Ltr 87-07, Info Transmittal of Final Rulemaking for Revs to Operator Licensing - 10CFR55 & Conforming Amends ML20055C7621990-06-18018 June 1990 Informs NRC of Plans Re Licensing of Senior Reactor Operators (Sros) Limited to Fuel Handling at Plants.Util in Process of Implementing New Program for Establishment & Maint of Licensed SROs Limited to Fuel Handling at Plants ML20055C7471990-06-15015 June 1990 Requests That Listed Operator Licenses Be Discontinued ML20043G1331990-06-14014 June 1990 Responds to NRC Re Violations Noted in Insp Repts 50-352/90-13 & 50-353/90-12.Corrective Actions:Boxes of Completed Procedures Improperly Stored Shipped to Util Storage Vault by 900406 ML20043G9981990-06-12012 June 1990 Forwards, Core Operating Limits Rept for Unit 1 Reload 2, Cycle 3 & Core Operating Limits Rept for Unit 2,Cycle 1. Repts Submitted in Support of Tech Spec Change Request 89-13 Re Parameter Limits,Per Generic Ltr 88-16 ML20043G7311990-06-0808 June 1990 Provides Addl Response to Generic Ltr 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping. Welds Examined During Last Refueling Outage Addressed ML20043G7501990-06-0808 June 1990 Requests Withdrawal of 900516 Tech Spec Change Request 90-11-1 Re Extension of Snubber Visual Insp Period.Change No Longer Needed Since Unit Shutdown on 900605 & Visual Insp of Three Affected Snubbers Performed on 900607 ML20043F8021990-06-0808 June 1990 Forwards Monthly Operating Repts for May 1990 for Limerick Units 1 & 2 & Revised Pages to Mar 1990 Rept for Unit 2 & Apr 1990 Rept for Units 1 & 2 ML20043D8101990-05-29029 May 1990 Forwards Application for Amends to Licenses NPF-39 & NPF-85, Consisting of Tech Specs Change Request 89-07 to Relocate Radiological Effluent Tech Specs to ODCM or Process Control Program,Per Generic Ltr 89-01 ML20043E6571990-05-25025 May 1990 Forwards Public Version of Rev 135 to Epips,Including Rev 11 to EP-202,Rev 14 to EP-282,Rev 12 to EP-284,Rev 8 to EP-312 & Rev 9 to EP-410.W/DH Grimsley 900607 Release Memo ML20055C5121990-05-18018 May 1990 Provides Info Inadvertently Omitted in Re Property Insurance Coverage for Plants.Limerick Generating Station Unit 2 Should Have Been Ref as Being Included Under Insurance Coverage ML20043A7881990-05-16016 May 1990 Requests Exemption from Requirement to Perform Biennial full-participation Onsite/Offsite Emergency Exercise for Plant During 1990 ML20055C4851990-05-15015 May 1990 Forwards Annual Financial Repts for 1989 for Philadelphia Electric Co,Pse&G,Atlantic Energy,Inc & Delmarva Power & Light Co ML20043B1501990-05-14014 May 1990 Forwards Public Version of Rev 134 to Epips,Consisting of Rev 10 to EP-230,Rev 4 to EP-255,Rev 1 to EP-302,Rev 7 to EP-304 & Rev 3 to EP-314.Release Memo Encl ML20043A2361990-05-14014 May 1990 Responds to NRC Re Violations Noted in Insp Repts 50-352/90-07 & 50-353/90-06.Corrective Actions:Sampling Review of Plant Baseline Data Will Be Performed to Ensure Product Code Number Correctness for Components ML20042F4481990-05-0101 May 1990 Advises That Plant Transient Response Implementing Plan Procedures & Related Ref Matls Provided to Dj Florek,Nrc Region I,On 900430.Documents Provided in Response to NRC 900327 Ltr Re Preparation for Planned NRC Insp of Procedure ML20042E8741990-04-27027 April 1990 Responds to Generic Ltr 87-07, Info Transmittal of Final Rulemaking for Revs to Operator Licensing. Certifies That Limerick Operator Requalification Training Program Renewed on 900125 & Peach Bottom Subj Program Renewed on 890622 ML20034C7381990-04-26026 April 1990 Forwards Application for Amends to License NPF-39 & NPF-85, Consisting of Tech Spec Change Request 89-16 Re Reactor Protection Sys & Eccs.Ge Proprietary Rept RE-019, Tech Spec Improvement... Encl.Ge Rept Withheld (Ref 10CFR2.790) ML20034A0781990-04-0909 April 1990 Provides Supplemental Info Re 10CFR50.63, Loss of All AC Power, Including Station Blackout Duration.Hpci or RCIC Sys Capable of Supplying Sufficient Condensate Inventory to Maintain Core Covered Assuming Leak Rate of 25 Gpm Per Pump ML20042E0881990-04-0909 April 1990 Forwards Addl Info Re 891011 Tech Spec Change Request 89-09 to Reduce Number of Suppression chamber-to-drywell Vacuum Breakers Required to Be Operable ML20042E0201990-04-0606 April 1990 Forwards Vols 1-3 to Preservice Insp Summary Rept, & Books 1-3 to Form NIS-2 for Preservice Insp Interval 1985-1990, Per 10CFR50.55a(g) & ASME Code Section Xi,Paragraph IWA-6230 ML20033G6291990-03-23023 March 1990 Forwards Application for Amends to Licenses NPF-39 & NPF-85, Consisting of Tech Spec Change Request 14,revising Section 3/4.0 & Bases to Remove 3.25 Limit on Extending Surveillance Intervals,Per Generic Ltr 89-14 ML20012E2151990-03-20020 March 1990 Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants,' for Peach Bottom.Response for Limerick Generating Station Will Be Provided by 900504 ML20012C2931990-03-12012 March 1990 Responds to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey, Per 900118 Request ML20033H1241990-03-12012 March 1990 Lists Levels & Sources of Property Insurance for Facilities, Per 10CFR50.55(w) ML20012D9511990-03-0909 March 1990 Forwards Public Version of Revised Epips,Including Rev 10 to EP-203,Rev 12 to EP-317 & Rev 18 to EP-292.W/DH Grimsley 900322 Release Memo ML20012A3631990-03-0101 March 1990 Responds to NRC Re Violations Noted in Insp Rept 50-353/89-32 on 891211-15.Corrective Action:Util Will Document Both Receipt & Shipment of Fuel Loading Chambers on Next Semiannual Doe/Nrc Form 742 1990-09-07
[Table view] |
Text
,
PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 PHILADELPHIA, PA.19101 JOHN S. KEMPER April 5, 1984 vicencsioc,.y ENG6ha E RONG AND RESE ARCH Docket Nos. 50-352 50-353 Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing U. S. Nuclear Regulatory Commiseion Washington, D.C.
20555
Subject:
Limerick Generating Station, Units 1 and 2 Pump and Valve Operability Review Team (PVORT)
Audit of January 17-20, 1984 - PECO Resolution of Open Items
Dear Mr. Schwencer:
We are pleased to provide in the enclosures the information necessary to resolve the finding identified during the subject audit. The finding is summarized below. Uc trust that the enclosed information will assist you in the closeout of SER Open Issue #6 concerning operability qualification of mechanical equipment.
As a result of the subject audit conducted at Limerick Generating Station, the following finding was issued:
"During the PVORT audit, it was noted that the original design parameters for some components were exceeded by the current expected normal operating or accident parameters." From discussions at the audit exit interview, the design parameters were identified as temperature and pressure.
The following is " Action Required for Open Items" from the finding:
"1.)
Review all safety related pumps and valves.
2.)
Identify those equipment items for which the original design parameters were exceeded by the current accident or normal values.
3.)
In each case'for which the original design parameters were exceeded, the applicant should provide justification that pump and valve operability is not adversely affected."
In accordance with the above " Action Required for Open Items", we have completed our review; and the results are included on the enclosures attached.
8404100189 840405 gDRADOCK 05000352 l '
._^
I
. Mr. A. Schwencer, Chief Page 2.
In addition, appropriate portions of the LGS FSAR, including Tables 3.9-8 and 3.9-19, are in the process of being revised to provide an up-to-date list of safety related, active mechanical equipment to be consistent with AE/NSSS active safety related equipment lists. These revisions to the LGS FSAR will be submitted as soon as they are available.
. Should any additional.information be required, please do not hesitate to contact us.
Very truly yours, A$r2/- r Enclosures Copy to: See Atta:hed Service List.
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r-cca Judge Lawrence Brenner (w/o enclosure)
Judge Peter A. Morris (w/o enclosure)
Judge Richard F. Cole (w/o enclosure)
Troy B. Conner, Jr., Esq.
(w/o enclosure)
Ann P. Hodgdon, Esq.
(w/o enclosure)
Mr. Frank R. Ranano (w/o enclosure)
Mr. Robert L. Anthony (w/o enclosure)
Mr. Marvin I. Iewis (w/o enclosure)
Charles W. Elliot, Esq.
(w/o enclosure)
Zori G. Ferkin, Esq.
(w/o enclosure)
Mr. Thanas Gerusky (w/o enclosure)
Director, Penna. Emergency (w/o enclosure)
Managerent Agency Mr. Steven P. Hershey
.(w/o enclosure)
Angus Inve, Esq.
(w/o enclosure)
Mr. Joseph H. White, III (w/o enclosure)
David Werson, Esq.
(w/o enclosure)
Robert J. Sugannan, Esq.
(w/o enclosure)
Spence W. Perry, Esq.
(w/o enclosure)
Jay M. Gutierrez, Esq.
(w/o enclosure)
Atctnic Safety & Licensing (w/o enclosure)
Appeal Board Atanic Safety & Licensing (w/o enclosure)
Board Panel Docket & Service Section (w/o enclosure)
Martha W. Bush, Esq.
(w/o enclosure)
James Wiggins (w/o enclosure)
Mr. Tinothy R. S. Campbell (w/o enclosure)
Phyllis Zit=er.
(w/o enclosure)
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l 1
ENCLOSURE
- - Review Summary of BOP Active Pumps - Review Summary of BOP Valves
. - Review Summary of NSSS Active Essential Pumps & Valves
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ENCLOSURE 1 4
-Review Summary of' BOP Active Pumps 1.-
Emergency Service Water Pumps Tag No.'s O(A-D) P548 have a maximum -
shut-off -head.of 137 PSI,: and the ESW piping system and the pumps are designed for this pressure. This system is satisfactory.
2.;
RHR Servihe Water Pumps Tag No..'s O(A-D) P506 have a maximum shut-
.off head of 169 PSI'with' the piping system design pressure of 155 PSI. The. system will only experience the 169 PSI less than 17. of its operating time, which is. acceptable in accordance with'ASME Section III-.1973 Ed. ND-3612.3, applicable to Limerick.
3.
Control ' Room Chilled. Water Pumps Tag No. 's O(A&B) P162 have a maximum shut-off head of.110 PSI; however, the piping system is
- designed for 120 PSI,'which the system will never experience during operation.. The. piping system was hydrotested with the pump isolated to protect the pump from any overpressure.
Diesel Oil Transfer Pumps, Tag No. 's 1-(A-D) P514 have a maximum 4.
shut-off head of 31 PSI whereas the associated ~ piping system was designed to 50 PSI. When the piping system was hydrotested the-pumps were isolated to prevent overpressure. The maximum that this system could experience during operation is 31 PSI.
i 5..
Safeguard Piping Fill Pumps Tag No.'s 1(A&B) P256 have a ma::imum
~
shut-off head of 69 PSI. The safeguard, piping system is designed.
to;150 PSI but will only experience 69 PSI, the shut-off head of the pump. When the piping systems were hydrotested, the pumps were isolated to prevent overpressurization.
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Review Sununary of BOP Valves SH 1/4
. Valve Dat6 ' Valve Valve Data
. System Design /
- - Sheet Tag
.Line Sheet Maxinum Cond.
- No..
.No.'s' No.
PD To Pmax max.
T PD To Pmax max T
Remads P -102 49-1F010 HBB-130 125 100 125 100 125 170 125 170 Analyzed to 750 PSI @
-2F010 SCO*F See Notes 1, 7&8 1
55-lF004 HB&-110 125 100 125 100 125 170 125 170 3ee Notes'l & 2, 7
-2F004 2
Sl-1F006A,B HBB-118 140 320 140 320 190 360 190 360
-2F006A,B, 3
51-1F075.
GBB-11 500 320 500 320_
420 360 500 360 Analyzed to 500 PSI @
-2F075 480'F See Notes 3,7
.4 52-127,128
_HBB-134 150 200 150 200 125 212 125 212 See Notes 1 & 2,7
-227,228 18 51-1F040 GBB-104 500 320 500 320 420 360 500 360
. Analyzed to IS00Tsr
-lF049
@ 500*F See Notes 2, 3, 7 & 8
-2F040 GB&-104 500 320 500 320 420 360 500 360
-2F049 18 51-105A,B GBB-109 500 320 500 320 420 360 500 360
-205A,B 20 1F068A,B 2B-103 455 150 455 150 420 480 500 480 See Notes 2 & 3, 7
-2F068A,B 21-52-lF031A,B. GBB-112 420 212 475 212-500 212 550 212 See Notes 3,7
-2F031A,B 27 Sl-1F073 GBC-106 455 150 500 150 420 320 500 320 See Notes 2 & 3,7
-2F073 27 51-1F014A,B GBC-102 455 150 500 150 420 480 500 480
-2F014A,B 35 55-1F095 HBB-144 30 212 30 212 185 366 185 366 See Notes 1,2 & 7
-2F095
'36-55-1F093 HBB-144 65 360 65 360 185 366 185 366
-2F093 36 49-lF080 HBB-145 65 360 '65 360 160 268 160 268
-2F080 36 49-lF084 HBB-145 30 212 30 212 160 268 160 268
-2F084
~
38.
51-1F007A-D GB-109 420 320 500 320 420 360 500 360 Analyzed to 151TO PSI ~
-2F007A-D
@ 500'F See Notes 2, 368 39 61-111 HBB-164 150 140 150 140 65 340 65 340 See Notes 1,2 & 7
-211 39 61-131-HBB-165 150 140 150 140 65 340 65 340
-231 40 57-160A,B-HBD-161 :100.150 100 150 150 150 150 150 Analyzed for 175 PSI
-260A,B;
@ 100'F See Notes 1, 2 & 7 T-23/6(pg1) -
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SH 2/4 Valve Data Valve Valve Data System Design /
Sheet Tag Line-Sheet Maximum Cond.
.No.
~No.'s No.
Po To Pmax max T
PD To Pmax max T
Remarks P-103 1
55-lF042 HBB-109 70 212 70 212 125 170 125 170 See Notes 1,2 & 7
-2F042 1
51-lF004A-D HBB-ll7 70 212 70 212 190 360 190 360
-2F004A-D
~
1
.51-125A,B GBB-108 430 170 430 170 420 320 500 320 See Notes 2,3 & 7
-225A,B 2
52-lF015A,B GBB-ll4 445 212 445 212 500 212 550 212
-2F015A,B 2
49-1F060 HBB-101 25 268 25 268 160 268 160 268 See Notes 1,2 & 7
-2F060 2
55-lF072 HBB-108 25 268 25 268 185 366 185 366
-2F072
~P-104 1
49-1F022 EBB-133 1300 170 1500 170 1325 170 1575 170 Valve Analyzed foF
-2F022 max Pd=2160 See Notes 4, 7 2
55-1F007 EBB-129 1396 170 1625 170 1423 170 1707 170 Valve analyzed foF
-2F007 max Pd=2160 See !btes 4, 7 2
55-1F008 EBB-134 1396 170 1625 170 1423 170 1707 170 Valve analyzed for
-2F008 max Pd=2160 See Notes 4, 7 2
55-1F006 EBB-129 1396 170 1625 170 1423 170 1707 170
-2F006 3
55-lF071.
EBB-134 1396 170 1625 170 1423 170 1707 170
-2F071 8
49-lF012 EBB-135 1300 170 1500 170 1325 170 1575 170
-2F012 19 55-lF0ll EBB-134 1396 170 1625 170 1423 170 1707 170 Valve analyzed for
-2F011 max Pd=2160 See Notes 4, 7 P-106 2
61-130 HCB-107 150 140 150 140 65 291 65 291 See Notes 1, 7
-230 2
61-110' HCB-106 150 140 150 140 65 291 65 291
-210
~ T-23/6(pg2)
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SH 3/4 V21ve Data Valve Valve Data System Design /
Sheet Tag Line Sheet Maxinum Cond.
No, No 's No.
_ PD To Pmax max Pp TD P T
h ds T
max max P-114A
^9_
61-102,112, HCB-123 150 140 150 140 55 291 55 291 See Notes 6A, 7
-132, 202
-212, 232 10 48-lF006A,B DCA-102 1680 150 1680 150 1500 598 1500 598 See Notes 6B, 7
-2F006A,B 12 49-lF002 HBB-101 75 190 75 190 160 268 160 268 See Notes 6A, 7
-2F002 15 57-116 HBB-125 62 340 62 340 150 340 150 340
-216 26 Sl-156A,B GBC-104 155 95 182 95 420 480 500 480 See Notes 6A, 7
-157A,B
-256A,B (EC-104 155 95 182 95 420 480 500 480
-257A,B 26 Sl-158A,B GBC-103 155 120 182 120 420 480 500 480
-258A,B 28 57-165,166 HBD-161 125 150 125 150 150 150 150 150
-167,169 i
- -N s' T-23/6(pg3)
/
SH 4/4 General Note: All valves not listed have been reviewed against postulated accident conditions of temperature and pressure and are acceptable.
~ NOTES: P.eferred to under " Remarks" column:
~1)
Valves are designed to standard pressure rating of 150# per ANSI B16.5 which allows a maximum pressure of 275 PSI @ 100 F and 150 PSI @ 500 F.
(Intermediste pressures / corresponding temperatures are per Table 2 of B16.5.)
2)
Material allowable stresses do not vary thru temperature range 0
from 100 F to 650 F for ASME Class' 2 and 3 components.
3)
Valves are designed _t'o standard pressure rating of 300#_ per ANSI B16.5 which allows a maximum pressure of 720 PSI @ 100 F and 625 PSI @ 500 F.
(Intermediate pressures are per Table 3
.of B16.5.)
4)
Valves are designed to standard pressure rating of 900# per ANSI B16.5 which allows a maximum pressure of 2160 PSI @
10G F.
(Alternate pressures / temperatures are per Table 6 of B16.5.)
~
5)
Valves are designed to standard pressure rating of 600# per ANSI'B16.5 which allows a maximum pressure of 1440 PSI @
100 F.
(Alternate pressures / temperatures are per Table 5 of B16.5.)
-6)
Valves are designed to standard pressure rating of 1500# per ASME Section III -- 1974 Ed. Table NB-3531-6.
(Flanged end valves 'for 2" and smaller per NB-3513.)
t a)
For carbon steel valves maximum pressure is 3750 PSI
@ 1000F with alternate values per above Table.
b)
For stainless steel (Type 316L) maximum pressure is 2570 PSI @ 1000F with alternate values per above Table.
7)
LAll changes were checked against applicable stress analyses and determined to have no impact. In addition, where pressures y
have'bcen revised, the " actual operating" differential pressure
~
requirements for the valve / operator were confirmed to be
'within the qualified range for the valve / operator as supplied.
8)
Valve body is actually a higher pressure rating than required;-
therefore enveloping pressure was ueed in the analysis.
T-23/6 L(pg 4)
ENCLOSURE 3 1 of 8 REVIEW
SUMMARY
OF NSSS ACTIVE ESSENTIAL PUMPS AND VALVES i
NRC PVORT AUDIT GENERIC FINDING:
Findina:
During the PVORT audit it was noted that the original design parameters for some components were exceeded by the current expected normal operating or accident parameters.
Required Action:
1)
Review all safety related pumps and valves.
2)
Identify those equipment items for which the original design parameters were exceeded by the current accident or normal values.
3)
In each case for which the original design parameters were exceeded, the applicant should provide justification that pump and valve operability is not adversely affected.
GE Response:
1)
GE has completed a review of all active essential pumps and valves in the GE NSSS scope of supply.
2)
Table 1 lists all of the active essential pumps and valves in the GE NSSS scope of supply, with the corresponding component design parameters (pressure and temperature, per GE purchase specifications) and the maximum service conditions (i.e., worst case normal or accident pressures and temperatures, per GE system design specifications and process diagrams, except the ATWS transient conditions are excluded).
Under the " REMARKS" column of Table 1, "O.K." means the maximum service conditions are equal to or less than the component design parameters, and no further evaluation effort was required.
3)
For those components which the maximum service conditions exceed the design parameters, the enclosures to Table 1 provide justifi-cation for the exceedance.
ATWS TRANSIENT The GE design basis for ATWS (Anticipated Transient Without Scram) requires that the pressure integrity of the primary pressure boundary components shall be assured for the initial (short-term) peak ATWS conditions under ASME code
-rules for these conditions using Service Level C limits.
None of the components are required to perform any active safety related functions during the initial peak ATWS transient. The subsequent (long-term) peak ATWS conditions are
2 of 8
~
considered for the active safety related functions of applicable components; however, these ATWS conditions are within the component design conditions.
Consequently, the peak ATWS transient conditions do not affect the operability of~the active safety related pumps and valves listed in Table 1, and these peak ATWS conditions are not included in the maximum service conditions listed in Table 1.
All components exposed to ATWS transient conditions have been evaluated as acceptable under these conditions in accordance with the GE design basis for ATWS.
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3 of 8 TABLE 1 LIMERICK GE NSSS PUMPS & VALVES COMPONENT PARAMETERS - DESIGN VS. MAX SERVICE CONDITIONS I~
COMPONENT MAX SERVICE DESIGN (1)
CONDITIONS (2)
MPL DESCRIPTION PRESS.
TEMP.
PRESS.
TEMP.
REMARKS B21-F013 MS Safety Relief Valve 1250 575 1250 575
- 0. K.
821-F022/28-Main Steam Isolation Valves 1250
-575 1250 575 0.K.
C11-F009/182 CRD Solenoid Valves 500 180 105 200 Enc. A.1 C11-F010/180 SDV Vent Valves 1250 280 1250 500 Enc. A.2&B C11-F011/181 SDV Drain Valves 1250 280 1250 500 Enc. A.2&B C11-F160/162/163 ARI Valves 250 215 105 200
- 0. K.
C41-C001' SLC Pump / Motor 1400 150 1400 150
- 0. K.
C41-F004 SLC Explosive Valve 1400 150 1400 150
- 0. K.
C41-F029 ~
SLC Relief Valve 1540 150 1400 150
- 0. K.
E11-C002 RHR Pump / Motor 500 360 290 360
- 0. K.
E11-F015 RHR MO Valve 1500 575 1500 575
- 0. K.
~E11-F016 RHR MO Valve.
500 360 500 360 0.K.
E11-F017 RHR MO Valve 1250 575 1250 575
- 0. K.
E11-F021 RHR MO Valve 500 360 500 360
- 0. K.
E11-F027
-RHR M0 Valve 500 360 500 360
- 0. K.
E11-F041 RHR A0 Check Valve 1250 575-1250 575
- 0. K.
E11-F050 RHR A0 Check Valve 1250 575 1250 575
- 0. K.
t E21-C001 Core Spray Pump / Motor 500 212 290 212
- 0. K.
E21-F001 CS MO Valve 150=
500 100 212
- 0. K.
E21-F005:
CS MO Valve 1360 575 1360 575
- 0. K.
E21-F006 CS A0 Check Valve 1250 575 1250 575 0.K.
-E21-F037 CS MO Valve
.1360 575 1360 575-0.K.
E41-C001 HPCI Pump 1500 140 1670 140 Enc. A.3 E41-C002 HPCI Turbine 1250 SAT. - 1140 SAT.
- 0. K.
E41-F005 HPCI Swing Check Valve 1500 100 1670 170 Enc. A.4 E41-F012 HPCI MO Valve 2200 100 1670-170 Enc. A.5 E41-F021 HPCI_Stop Check Valve 150 366_
185 366-Enc. A.6 L
E51-C001 RCIC Pump.
-1500 140 1575 140 Enc. A.3 E51-C002 RCIC Turbine 1250 SAT.
1140 SAT.
- 0. K.
E51-F001 RCIC Stop Check Valve
-275 100 160-267 Enc. A.6 E51-F014 RCIC. Swing Check Valve
~1500 100-1575 170 Enc. A.6 E51-F019 RCIC M0 Valve 1500 100 1575 170 Enc. A.5 NOTES: '(1) Component Design Parameters per GE Purchase Specifications.
(2) Max Service Condition l Parameters per GE System Design I
Specifications and Process Diagrams, except ATWS conditions are excluded.
(3) Pressure = psig, Temperature = 'F 1 :
(4)' Max. service conditions are shown = component design conditions in many-cases where actual expected max serv. ice conditions are less than~ component design conditions.
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4 of 8 l
ENCLOSURE A JUSTIFICATION FOR EXCEEDANCE OF DFSIGN PARAMETERS 1.
C11-F009/182 - CRD Solenoid Valves Design pressure is not exceeded.
The design temperature of 180*F is exceeded by the maximum service temperature of 200*F.
A qualification test was successfully performed on the valve using an environment of 200*F plus margin.
Hence the design adequacy of this valve has been demonstrated at the higher maximum service temperature.
2.
C11-F010/180 & C11-F011/181 - SDV Vent & Drain Valves Design pressure is not exceeded.
The design temperature of 280*F is exceeded by the maximum service temperature of 500*F.
The valves supplied are 2,500 # valves, with a body material of SA352LCB.
According to ANSI B16.34-1977, the allowable working pressure for a 2,500 lb. of material SA352LCB at 500*F is 4,850 psig (See Enclosure B).
Further, the 500*F condition occurs after the scram (280*F design temperature) when the valve would have already operated.
3.
~E41-C001 & E51-C001 - HPCI & RCIC Pumps Design temperature is not exceeded. The design pressure of 1500 psig is exceeded by the maximum' service pressure of 1,670 psig for the HPCI pump and 1,575 psig for the RCIC pump.
The basis for the pressure transient is a worst case accumulation of conditions (105% turbine drive overspeed, maximum suction head, maximum discharge pressure at pump shutoff, etc.)
which occurs less than 1% of the operating time.
Paragraph 102.2.4 of the ANSI B31.1.0-1967 code allows 20% overpressure for events which occur less than 1% of the operating time. The 1,670 psig and 1,575 psig peak pressures are within 120% of design pressure (1.2 x 1500 = 1800 psig) as allowed by the code.
Further, operability on the HPCI and RCIC pumps can be addressed with respect to the following criteria:
1)
During any loading or pressurization event the deflections of the l:
pump' case and shaft must not be such that contact at close clearance
[
locations, such as wear rings, occurs due to the deflections.
l-2)
No permanent plastic deformation of the pump case and other parts shall take place which cause misalignment of the bearing and seal centerlines and other_ factors affecting shaft and case alignment.
3) 0verpressurization shall not cause failure of the shaft mechanical seal parts such as elastomeric seal-rings, carbon seal parts, or failure of metallic structural parts.
The HPCI and RCIC pumps are assessed as follows with respect to these criteria:
p I!
/
5 of 8 1)
Review of the pump designs show that their cases are basically symmetrical with shaf t centerlines.
Therefore, deformation of the pump cases and shafts due to overpressure would not be expected to affect bearing or seal clearance.
2)
If it were possible to have some local plastic yielding due to an overpressure condition, the case symmetry would prevent yielding to cause loss of bearing or seal clearances.
Further the pumps have been subjected to vendor hydrostatic tests at 150% of design pressures and subsequently operated satisfactorily.
Hence,'no detrimental plastic deformation occurred at these more severe overpressure conditions.
3)
The seal manufacturer for both the HPCI and RCIC pumps states that the mechanical seals see only suction pressure + 25 psi (i.e.,
s100 psig maximum) and the seal is designed for 1000 psig operating pressure.
Consequently, the pump seals are not affected by these maximum pump discharge overpressure transients.
4.
E41-F005 - HPCI Swing Check Valve The design temperature of 100'F is exceeded by the maximum service temperature of 170*F.
The design pressure of 1500 psig is exceeded by the maximum service pressure of 1670 psig.
Based on.the valve actual wall thickness and the 1968 Nuclear Pump and Valve Code, this valve cannot meet the code allowable pressure at these maximum service conditions.
However, the following rationale is used to justify the exceedance of the design conditions.
This is a Class 2, 600 #
valve, with a body material of A216, Gr. WCB.
a.
The allowable stress is 17,500 psi for temperatures up to 650'F.
Thus, the 170*F service temperature is justified.
b.
The. valve was designed according to the 1968 NP&V Code which did not specifiy a design method for any condition that is a variation from normal conditions.
Therefore, the criteria for Power Piping ANSI B31.1.0 - 1967 are adopted.
Paragraph 102.2.4 of this code allows up to 20% increase above the allowable stress during 1% of the operating time.
Since the maximum pressure of 1670 psig at 170*F is expected to occur less than 1% of the operating time, this allowance is applicable to valve E41-F005.
Based on the valve wall thickness and the provision of Paragraph 104.1.2(a) of B31.1.0, the valve stress at'the maximum service pressure is shown to be within the allowable stress value (17500 psi).
Therefore, the exceedance of the design pressure is also justified.
It can be concluded that valve E41-F005 is justified to maintain its structural integrity under the peak transient conditions.
6 of 8 The internal pressure does not affect the operability of ~ the check valves.
Therefore, operability of the valves is assured even though the peak pressure exceeds the design condition.
5.
E41-F012 & E51-F019 - HPCI & RCIC Motor Operated Valves a.
Pressure Integrity Table 5.1 lists the design conditions, the maximum service conditions and the maximum allowable pressure at the service temperature, of the valves.
The allowable pressures were based on the valves wall thickness, material, and the NP&V Code, 1968, pressure and tempera-
From 1968 NP&V Code, the maximum allowable stress of the valve material (A216, Gr. WCB) is 17500 psi, unchanged for all temperatures up to 650*F.
Therefore, the exceedance in temperature of the valves does not affect the pressure integrity of the valves.
As shown in Table 5.1, the maximum allowable pressure corresponding to the service temperature of the above valves is higher than the maximum service pressure. Thus, the pressure integrity of the valves is assured.
b.
Operability of Actuator Under Peak Pressure
'The Limitorque actuator motor capability is 25 ft-lb for valve
'~
E41-F012, and 5 ft-1b. for E51-F019. These motor sizes were sele'cted based on a design AP of 1400 psi and 1300 psi, respectively. The Limitorque motor-sizing procedure was used to calculate the maximum required torque to operate against the maximum service pressures of 1670 psig for valve E41-F012 and 1575 psig for valve E51-F019.
This calculation.showed that the motor capability exceeds the required torque by a 2 to 1 margin, approximately. Therefore.the maximum service pressure does not affect the actuator operability.
Table 5.2 summarizes the above results.
7 of 8 TABLE 5.1 PRESSURE-TEMPERATURE RATINGS Temperature Pressure Allowable Press.
Valve Design Service Design Service At Service Temp.
E41-F012 100*F 170*F 2200 psig 1670 psig 2110 psig E51-F019 100*F 170*F 1500 psig 1575 psig 1849 psig TABLE 5.2 ACTUATOR MOTOR SIZING Torque Due to Actuator Motor Valve Design AP-Maximum AP Maximum AP Torque Capability E41-F012 1400 psi 1670 psig 13.4 ft-lb 25 ft-lb E51-F019 1300 psi 1575 psi 2.04 ft-lb 5 ft-lb
8 of 8 6.
E41-F021, E51-F001 & E51-F014 - HPCI & RCIC Check Valves Table 6.1 lists the design conditions, the maximum service conditions and the maximum allowable pressure at the service temperature, of the valves.
The allowable pressures were based on the valve wall thicknesses, material, and the NP&V Code, 1968 pressure and temperature rating tables.
'From the 1968 NP&V Code, the maximum allowable stress of the valve material (A216,13r. WCB) is 17500 psi, unchanged for all temperatures up to 650*F.
Therefore, the exceedance.in temperature of the valves does not affect the pressure integrity of the valves.
As shown in Table 6.1, the maximum allowable pressure corresponding to the service temperature of the valves-is higher than the maximum service pressure.
Thus, the pressure integrity of the valves is assured.
The internal pressure does not affect the operability of the check valves.
Therefore, operability of the valves is assured even though the maximum service pressures exceed the design conditions.
TABLE 6.1 Temperature (*F)
Pressure (psia)
Allowable Press.
Valve Desian Service Desian Service At Service Temp.
E41-F021 366' 366 150' 185 190 psig E51-F001 100 267 275 160 220 psig E51-F014 100 170 1500 1575 2539 psig L
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