Forwards Responses to Listed 831207 Requests for Info Re Environ Qualification of Electrical Equipment,Including Resolution of Deficiencies & TMI Action Plan Items (NUREG-0737) to Be Installed at End of Cycle 10 OutageML20087H873 |
Person / Time |
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Site: |
Oyster Creek |
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Issue date: |
03/16/1984 |
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From: |
Fiedler P GENERAL PUBLIC UTILITIES CORP. |
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To: |
Crutchfield D Office of Nuclear Reactor Regulation |
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References |
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RTR-NUREG-0737, RTR-NUREG-737, TASK-03-05.B, TASK-2.B.3, TASK-2.D.3, TASK-2.E.4.1, TASK-2.F.1, TASK-2.K.3.19, TASK-3-5.B, TASK-RR, TASK-TM NUDOCS 8403220012 |
Download: ML20087H873 (18) |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217J3341999-10-19019 October 1999 Forwards Request for Addl Info Re Sale of Portion of Land Part of Oyster Creek Nuclear Generating Station Site Including Portion of Exclusion Area ML20217E0181999-10-0606 October 1999 Provides Nj Dept of Environ Protection Comments on Oyster Creek Nuclear Generating Station TS Change Request 267 Re Clarifications to Several TS Sections ML20216J7591999-09-30030 September 1999 Informs NRC That Remediation Efforts for Software Sys Etude & Rem/Aacs/Cico Have Been Completed According to Schedule & Now Y2K Ready 05000219/LER-1998-011, Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts1999-09-30030 September 1999 Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts ML20212J6721999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Oyster Creek Nuclear Generating Station on 990913.No Areas Identified in Which Licensee Performance Warranted Addl Insp Beyond Core Insp Program.Historical Listing of Plant Issues Encl ML20216K1421999-09-29029 September 1999 Provides NRC with Name of Single Point of Contact for Purpose of Accessing Y2K Early Warning Sys,As Requested by NRC Info Notice 99-025 ML20217D1661999-09-27027 September 1999 Forwards Proprietary Completed NRC Forms 396 & 398,in Support of License Renewal Applications for Listed Individuals,Per 10CFR55.57.Encl Withheld ML20217B2531999-09-24024 September 1999 Informs That on 980903,Region I Field Ofc of NRC Ofc of Investigations Initiated Investigation to Determine Whether Crane Operator Qualification/Training Records Had Been Falsified at Oyster Creek Nuclear Generating Station ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212A7921999-09-13013 September 1999 Forwards Second RAI Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, Issued on 950817 to Plant ML20212B5571999-09-10010 September 1999 Forwards Rev 11 to Oyster Creek Emergency Dose Calculation Manual, IAW 10CFR50,App E,Section V ML20211N2941999-09-0303 September 1999 Responds to NRC 990802 Telcon Request for Environ Impact Assessment of TS Change Request 251 Concerning Movement of Loads Up to 45 Tons with RB Crane During Power Operations ML20211J9831999-09-0202 September 1999 Discusses 990804 Telcon Re Sale of Portion of Oyster Creek Nuclear Generating Station Land.Requests Info Re Location of All Areas within Property to Be Released Where Licensed Radioactive Matl Present & Disposition of Radioactive Matl ML20211J6771999-08-30030 August 1999 Submits Response to NRC 990802 Telcon Request for Gpu to Provide Environ Impact Assessment for Tscr 251 ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj ML20211C0161999-08-19019 August 1999 Advises That Info Submitted by Ltr,Dtd 990618, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool, Holtec Rept HI-981983,rev 4,will Be Withheld from Public Disclosure,Per 10CFR2.790 ML20211B9011999-08-18018 August 1999 Forwards Rev 0 to EPIP 1820-IMP-1720.01, Emergency Public Info Implementing Procedure ML20210U4341999-08-17017 August 1999 Responds to to Chairman Dicus of NRC on Behalf of Fm Massari Concern About Oyster Creek Nuclear Generating Station Not Yet Being Fully Y2K Compliant ML20210Q7331999-08-12012 August 1999 Responds to Re TS Change Request (TSCR)264 from Oyster Creek Nuclear Generating Station.Questions Re Proposed Sale of Property within Site Boundary & Exclusion Area ML20210L6311999-08-0606 August 1999 Discusses Licensee Response to GL 92-01,Rev1,Suppl 1, Rv Structural Integrity, for Plant.Staff Has Revised Info in Rv Integrity Database & Releasing as Rvid Version 2 ML20210D2801999-07-22022 July 1999 Submits Response to Administrative Ltr 99-02 Operating Reactor Licensing Action Estimates. Estimate of Licensing Actions Projected for Fy 2000 Encl.No Projection Provided for Fy 2001 ML20209H5001999-07-14014 July 1999 Forwards Revised TS Pages 3.1-15 & 3.1-17 Which Include Ref to Note (Aa) & Approved Wording of Note H of Table 3.1.1, Respectively ML20210U4411999-07-12012 July 1999 Forwards Article from Asbury Park Press of 990708 Faxed to Legislative Officer by Mutual Constitute Fm Massari Indicating That Oyster Creek Nuclear Generating State Not Fully Y2K Compliant ML20209G1451999-07-0909 July 1999 Forwards Rev 1 to 2000-PLN-1300.01, Oyster Creek Generating Station Emergency Plan. Attachment 1 Contains Brief Summary of Changes,Which Became Effective on 990702 ML20209E0821999-07-0707 July 1999 Forwards TS Change Request 269 for License DPR-16,changing Component Surveillance Frequencies to Indicate Frequency of Once Per Three Months ML20209B7501999-07-0101 July 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Generic Ltr 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Licensee Y2K Readiness Disclosure for Ocngs,Encl ML20196G1361999-06-23023 June 1999 Provides Status of Corrective Actions Proposed in in Response to Insp Rept 50-219/98-80 & Revised Schedule for Completion of Actions Which Are Not Yet Complete ML20196E6421999-06-22022 June 1999 Forwards Revised Pages of TS Change Request 261,dtd 990618. Replacement Requested Due to Several Dates Being Omitted on Certain Pages ML20195G6541999-06-0707 June 1999 Discusses 981204 Initiation to Investigate Whether Contract Valve Technician,Was Discriminated Against for Raising Concern Re Use of Untrained/Unqualified Workers Performing Valve Repairs.Technician Was Not Discriminated Against ML20195G6631999-06-0707 June 1999 Discusses 981204 Intiation to Investigate Whether Contract Valve Technician Was Discriminated Against for Raising Concern Re Use of Untrained/Unqualified Workers Performing Valve Repairs.Technician Was Not Discriminated Against ML20209B0561999-06-0404 June 1999 Informs That NRR Has Reorganized,Effective 990328.Forwards Organizational Chart ML20195D0551999-06-0303 June 1999 Forwards TS Change Request 226 to License DPR-16,permitting Operation with Three Recirculation Loops.Certificate of Svc & Tss,Encl ML20195C5511999-05-25025 May 1999 Forwards Book of Controlled Drawings Currently Ref But Not Contained in Plant Ufsar.Drawings Were Current at Time of Submittal ML20206N7711999-05-11011 May 1999 Forwards Rev 0 to Oyster Creek Emergency Plan, IAW 10CFR50.47(b) & 10CFR50.54(q).Changes Became Effective on 990413 ML20206H9441999-04-28028 April 1999 Forwards Application for Amend to License DPR-16,requesting Approval to Handle Loads Up to & Including 45 Tons Using Reactor Bldg Crane During Power Operations,Per NRC Bulletin 96-002 ML20206B6991999-04-26026 April 1999 Forwards Copy of Rev 11 to UFSAR & Rev 10 to Oyster Creek Fire Hazards Analysis Rept. Without Fire Hazard Analysis ML20206D3801999-04-26026 April 1999 Forwards Rev 11 to UFSAR, & Rev 10 to Fire Hazards Analysis Rept, for Oyster Creek Nuclear Generating Station, Per 10CFR50.712(e) ML20206A9931999-04-22022 April 1999 Forwards Number of Personnel & Person Rems by Work & Job Function Rept for Period Jan-Dec 1998. Included in Rept Is Listing of Number of Station,Util & Contractor Personnel as Well as Diskette Reporting 1998 Occupational Radiation ML20206C8261999-04-22022 April 1999 Submits Financial Info IAW Requirements of 10CFR50.71(b) & 10CFR140.21 ML20205P8411999-04-15015 April 1999 Forwards TS Change Request 267 to License DPR-16,modifying Items in Sections 2 & 3 of Ts,Expanding Two Definitions in Section 1 & Modifying Bases Statements in Sections 2,3 & 4. Certificate of Svc Encl ML20205P5381999-04-14014 April 1999 Ack Receipt of Re Request for Exception to App J. Intended Correction Would Need to Be Submitted as Change to TS as Exceptions to RG 1.163 Must Be Listed in Ts,Per 10CFR50,App J ML20205P9401999-04-12012 April 1999 Informs NRC That Gpu Nuclear Is Modifying Oyster Creek FSAR to Reflect Temp Gradient of 60 F & to Correct Historical Record ML20205P0651999-04-0909 April 1999 Discusses 990225 PPR & Forwards Plant Issues Matrix & Insp Plan.Results of PPR Used by NRC Mgt to Facilitate Planning & Allocation of Insp Resources 05000219/LER-1998-015, Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page1999-04-0505 April 1999 Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page ML20205J3281999-04-0101 April 1999 Discusses Arrangements Made on 990323 for NRC to Inspect Licensed Operator Requalification Program at Oyster Creek Nuclear Generating Station During Week of 990524 ML20205H1081999-03-31031 March 1999 Forwards Current Funding Status for Decommissioning Funds Established for OCNPP,TMI-1,TMI-2 & SNEC ML20205F0611999-03-25025 March 1999 Submits Info on Sources & Levels of Property Insurance Coverage Maintained & Currently in Effect for Oyster Creek Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20205E1171999-03-24024 March 1999 Forwards Rev 39 to Oyster Creek Security Plan & Summary of Changes,Iaw 10CFR50.54(p).Rev Withheld ML20207F0331999-03-0404 March 1999 Forwards Insp Rept 50-219/98-12 During Periods 981214-18, 990106-07 & 20-22.Areas Examined During Insp Included Implementation of GL 89-10 & GL 96-05.No Violations Noted ML20207K2471999-02-25025 February 1999 Forwards Fitness for Duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Ny 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217E0181999-10-0606 October 1999 Provides Nj Dept of Environ Protection Comments on Oyster Creek Nuclear Generating Station TS Change Request 267 Re Clarifications to Several TS Sections 05000219/LER-1998-011, Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts1999-09-30030 September 1999 Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts ML20216J7591999-09-30030 September 1999 Informs NRC That Remediation Efforts for Software Sys Etude & Rem/Aacs/Cico Have Been Completed According to Schedule & Now Y2K Ready ML20216K1421999-09-29029 September 1999 Provides NRC with Name of Single Point of Contact for Purpose of Accessing Y2K Early Warning Sys,As Requested by NRC Info Notice 99-025 ML20217D1661999-09-27027 September 1999 Forwards Proprietary Completed NRC Forms 396 & 398,in Support of License Renewal Applications for Listed Individuals,Per 10CFR55.57.Encl Withheld ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212B5571999-09-10010 September 1999 Forwards Rev 11 to Oyster Creek Emergency Dose Calculation Manual, IAW 10CFR50,App E,Section V ML20211N2941999-09-0303 September 1999 Responds to NRC 990802 Telcon Request for Environ Impact Assessment of TS Change Request 251 Concerning Movement of Loads Up to 45 Tons with RB Crane During Power Operations ML20211J6771999-08-30030 August 1999 Submits Response to NRC 990802 Telcon Request for Gpu to Provide Environ Impact Assessment for Tscr 251 ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj ML20211B9011999-08-18018 August 1999 Forwards Rev 0 to EPIP 1820-IMP-1720.01, Emergency Public Info Implementing Procedure ML20210D2801999-07-22022 July 1999 Submits Response to Administrative Ltr 99-02 Operating Reactor Licensing Action Estimates. Estimate of Licensing Actions Projected for Fy 2000 Encl.No Projection Provided for Fy 2001 ML20209H5001999-07-14014 July 1999 Forwards Revised TS Pages 3.1-15 & 3.1-17 Which Include Ref to Note (Aa) & Approved Wording of Note H of Table 3.1.1, Respectively ML20210U4411999-07-12012 July 1999 Forwards Article from Asbury Park Press of 990708 Faxed to Legislative Officer by Mutual Constitute Fm Massari Indicating That Oyster Creek Nuclear Generating State Not Fully Y2K Compliant ML20209G1451999-07-0909 July 1999 Forwards Rev 1 to 2000-PLN-1300.01, Oyster Creek Generating Station Emergency Plan. Attachment 1 Contains Brief Summary of Changes,Which Became Effective on 990702 ML20209E0821999-07-0707 July 1999 Forwards TS Change Request 269 for License DPR-16,changing Component Surveillance Frequencies to Indicate Frequency of Once Per Three Months ML20209B7501999-07-0101 July 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Generic Ltr 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Licensee Y2K Readiness Disclosure for Ocngs,Encl ML20196G1361999-06-23023 June 1999 Provides Status of Corrective Actions Proposed in in Response to Insp Rept 50-219/98-80 & Revised Schedule for Completion of Actions Which Are Not Yet Complete ML20196E6421999-06-22022 June 1999 Forwards Revised Pages of TS Change Request 261,dtd 990618. Replacement Requested Due to Several Dates Being Omitted on Certain Pages ML20195D0551999-06-0303 June 1999 Forwards TS Change Request 226 to License DPR-16,permitting Operation with Three Recirculation Loops.Certificate of Svc & Tss,Encl ML20195C5511999-05-25025 May 1999 Forwards Book of Controlled Drawings Currently Ref But Not Contained in Plant Ufsar.Drawings Were Current at Time of Submittal ML20206N7711999-05-11011 May 1999 Forwards Rev 0 to Oyster Creek Emergency Plan, IAW 10CFR50.47(b) & 10CFR50.54(q).Changes Became Effective on 990413 ML20206H9441999-04-28028 April 1999 Forwards Application for Amend to License DPR-16,requesting Approval to Handle Loads Up to & Including 45 Tons Using Reactor Bldg Crane During Power Operations,Per NRC Bulletin 96-002 ML20206D3801999-04-26026 April 1999 Forwards Rev 11 to UFSAR, & Rev 10 to Fire Hazards Analysis Rept, for Oyster Creek Nuclear Generating Station, Per 10CFR50.712(e) ML20206B6991999-04-26026 April 1999 Forwards Copy of Rev 11 to UFSAR & Rev 10 to Oyster Creek Fire Hazards Analysis Rept. Without Fire Hazard Analysis ML20206C8261999-04-22022 April 1999 Submits Financial Info IAW Requirements of 10CFR50.71(b) & 10CFR140.21 ML20206A9931999-04-22022 April 1999 Forwards Number of Personnel & Person Rems by Work & Job Function Rept for Period Jan-Dec 1998. Included in Rept Is Listing of Number of Station,Util & Contractor Personnel as Well as Diskette Reporting 1998 Occupational Radiation ML20205P8411999-04-15015 April 1999 Forwards TS Change Request 267 to License DPR-16,modifying Items in Sections 2 & 3 of Ts,Expanding Two Definitions in Section 1 & Modifying Bases Statements in Sections 2,3 & 4. Certificate of Svc Encl ML20205P9401999-04-12012 April 1999 Informs NRC That Gpu Nuclear Is Modifying Oyster Creek FSAR to Reflect Temp Gradient of 60 F & to Correct Historical Record 05000219/LER-1998-015, Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page1999-04-0505 April 1999 Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page ML20205H1081999-03-31031 March 1999 Forwards Current Funding Status for Decommissioning Funds Established for OCNPP,TMI-1,TMI-2 & SNEC ML20205F0611999-03-25025 March 1999 Submits Info on Sources & Levels of Property Insurance Coverage Maintained & Currently in Effect for Oyster Creek Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20205E1171999-03-24024 March 1999 Forwards Rev 39 to Oyster Creek Security Plan & Summary of Changes,Iaw 10CFR50.54(p).Rev Withheld ML20207K2471999-02-25025 February 1999 Forwards Fitness for Duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Ny ML20206S2541999-01-20020 January 1999 Confirms Resolution of Thermo-Lag Fire Barriers in Fire Zones OB-FZ-6A & OB-FZ-6B (480 Switchgear Rooms) IAW Previous Commitments Contained in Gpuns Ltrs to NRC & 971001 ML20199J2631999-01-18018 January 1999 Requests That Listed Changes Be Made to Correspondence Distribution List for Oyster Creek Generating Station ML20199D0271999-01-11011 January 1999 Requests Listed Addl Info in Order to Effectively Review TS Change Request 264 Re Ownership of Property within Exclusion Area ML20199A6521999-01-0707 January 1999 Notifies That Reactor Operators G Scienski,License SOP-11319 & D Mcmillan,License SOP-3919-4 Have Terminated Licenses at Oyster Creek Nuclear Generating Station, Effective 990101 ML20198T1061999-01-0606 January 1999 Forwards Rev 15 to Gpu Nuclear Corporate Emergency Plan for TMI & Oyster Creek Nuclear Station. with Summary of Changes Which Reflect Use of EALs Approved in NRC Ltr to Gpun on 980908 & Other Changes Not Related to Use of New EALs ML20198K0331998-12-23023 December 1998 Forwards Change Request 268 for Amend to License DPR-16. Amend Would Change TS to Specify Surveillance Frequency of Once Per Three Months ML20198H0181998-12-22022 December 1998 Forwards Attachment Addressing New Info & Modifying 980505 Submittal Re Request for Change to Licensing Bases for ECCS Overpressure,In Response to NRC Bulletin 96-03, Potential Plugging of ECCS by Debris in Bwrs ML20198H8521998-12-16016 December 1998 Dockets Completion of Physical Inventory Performed in July 1997,as Addl Info to Nuclear Matl Balance Rept Submitted on 980416 ML20196H4461998-12-0202 December 1998 Provides Final Response to NRC GL 96-01, Testing of Safety-Related Logic Circuits ML20196B4471998-11-23023 November 1998 Provides Required Response 2 to NRC Bulletin 96-003, Potential Plugging of ECC Suction Strainers in Bwrs. During Recently Completed 17R Refueling Outage,New Strainers Were Installed ML20195J8451998-11-12012 November 1998 Forwards Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan, as Change Previously Made Without Appropriate Notification to NRC ML20195C7201998-11-11011 November 1998 Forwards 120-day Required Response to GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment, ML20195E1221998-11-10010 November 1998 Notifies NRC of First Time Usage of Code Case N-504 & Inclusion Into OCNGS ISI Program,As Accepted by RG 1.147, Inservice Insp Code Case Acceptability ML20155J6851998-11-0505 November 1998 Forwards TS Change Request 266,to Modify Safety Limits & Surveillances of LPRM & APRM Sys & Related Bases to Ensure APRM Channels Respond within Necessary Range & Accuracy & Verify Channel Operability ML20155H5641998-11-0202 November 1998 Informs That Bne Has No Comments on Proposed Change 259 to Ts,Correcting Required Water Level in Condensate Storage Tank So That Design Basis Is Correctly Implemented ML20155G3741998-10-29029 October 1998 Forwards Response to NRC 980619 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L2551990-09-14014 September 1990 Advises of Preparation for Final Refueling Outage to Complete Second 10-yr Interval for Inservice Insps ML20059F5121990-09-0505 September 1990 Requests Exemption from Filing Requirement of 10CFR55.45(b)(2)(iii) to Allow Submittal of NRC Form 474, Simulator Facility Certification, After 910326 Deadline & to Allow Administering of Simulator Portion of Test ML20059F7441990-08-31031 August 1990 Forwards Util Review of NRC Backfit Analysis for Hardened Wetwell Vent.Nrc Analysis Does Not Support Conclusion That Hardening Existing Vent Is cost-beneficial Mod for Plant ML20059E9061990-08-30030 August 1990 Forwards Response to 900808 Request for Addl Info Re NRC Bulletin 90-002, Loss of Thermal Margin Caused by Fuel Channel Box Bow ML20059G1841990-08-29029 August 1990 Ack NRC Request to Perform Type C Testing During Unscheduled Outage,As Plant Conditions Will Allow.Type C Exemptions Should Remain in Effect Until New Outage Start Date ML20059C8231990-08-27027 August 1990 Advises That SPDS Enhancements Described in Completed,Per 900628 Request.Offline & Online Testing Completed & Enhancements Considered to Be Operational ML20059C8571990-08-24024 August 1990 Provides Results of Evaluation of Ability to Meet Acceptance Criteria for Eccs,In Response to 900804 Notice of Violation. Plant Meets Acceptance Criteria Contained in 10CFR50.46 W/ Valve Logic Design Deficiency in Containment Spray Sys ML20058N0781990-08-0909 August 1990 Submits Info Re pressure-temp Operating Limits for Facility, Per Generic Ltr 88-11.Util Recalculated Adjusted Ref Temp for Each Belt Line Matl as Result of New Displacement Per Atom Values ML20063P9521990-08-0909 August 1990 Advises That Response to NRC 900523 Request for Assessment of Hazardous Matl Shipment Will Be Sent by 910531 ML20058L9521990-08-0303 August 1990 Forwards Rev 2 to Security Contingency Plan.Rev Withheld (Ref 10CFR73.21) ML20058L9551990-08-0303 August 1990 Responds to SALP Rept 50-219/88-99.Although Minor,Several Factual Errors Noted.Dialogue Promotes & Identifies Areas Where Improvements Should Be Made ML20056A2071990-07-30030 July 1990 Forwards Response to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Record Review Performed & Sys Walkdowns Completed to Assemble Requisite List ML20055H7961990-07-20020 July 1990 Advises of Change to Preventive Maint Program for Electromatic Relief Valves.Rebuild Schedule Will Be Modified to Require Rebuilding Two or Three Valves During Refueling Outage & Remaining Valves During Next Refueling Outage ML20058N9911990-07-20020 July 1990 Partially Withheld Response to NRC Bulletin 90-002 Re Loss of Thermal Margin Caused by Box Bow (Ref 10CFR2.790(b)(1)) ML20055J0481990-07-19019 July 1990 Requests 2-wk Extension for Submittal of Response to Re Installation of Hardened Wetwell Vent W/ Appropriate Extension Period to Be Decided Pending Outcome of 900724 Meeting Discussion W/Bwr Owners Group ML20064A1221990-07-11011 July 1990 Discusses 900710 Telcon W/Nrc Re Util Corrective Actions in Response to NRC Finding That Operator Received Passing Grade on Administered Requalification Exam in 1989 Should Have Received Failing Grade.Corrective Actions Listed ML20055F8491990-07-10010 July 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 188,reducing Low Condenser Vacuum Scram Setpoint ML20043H7471990-06-21021 June 1990 Confirms Telcon W/A Dromerick Re Util Plans to Inspect CRD Hydraulic Control Units During Plant Walkdown to Address USI A-46, Seismic Qualification of Equipment in Operating Nuclear Power Plants. Walkdown Planned in Oct 1992 ML20043H2301990-06-14014 June 1990 Documents Licensee Commitment to Improve Seismic Restraints for Diesel Generator Switchgear Encls,Per 900613 Telcon W/ Nrc.Engineering Will Be Finalized & Mods Completed Prior to 900622 ML20043F7581990-06-0707 June 1990 Responds to Request for Info Re Util Compliance W/Generic Ltr 88-01 & Insp Plans for Upcoming 13R Outage.Frequency of Insp of Welds Classified as IGSCC Categories C,D & E Will Not Be Reduced During 13R Outage ML20043C5801990-05-25025 May 1990 Provides Descriptions & Conclusions of Three Remaining Issues of SEP Topic III-7B.Issues Include,Evaluation of Drywell Concrete Subj to High Temp & Thermal Transients ML20043C2461990-05-25025 May 1990 Forwards Rev 7 to EPIP 9473-IMP-1300.06 & Rev 4 to Radiological Controls Policy & Procedure Manual 9300-ADM-4010.03, Emergency Dose Calculation Manual. ML20043B2981990-05-21021 May 1990 Responds to NRC 900420 Ltr Re Violations Noted in Insp Rept 50-219/90-06.Corrective Actions:Incident Critique Rept Incorporated as Required Reading for Appropriate Operations Personnel & Change Made to Procedure 201.1 ML20043D0701990-05-17017 May 1990 Provides NRC W/Addl Info Re SPDS & Responds to Concerns Raised During 900117 & 18 SPDS Audit Documented in 900130 Ltr ML20043B3901990-05-0909 May 1990 Responds to NRC 900408 Ltr Re Violations Noted in Insp of License DPR-16.Corrective Actions:Two Narrow Range Drywell Pressure Monitoring Instruments to Be Provided During Cycle 14R Refueling Outage,Per Reg Guide 1.97,Category 1 ML20042G7071990-05-0808 May 1990 Forwards Summary of Initiatives & Accomplishments Re SALP, Per 891031 Commitment at mid-SALP Meeting.Plant Div Responsibilities Now Include Conduct of Maint Outages & Emergency Operating Procedure Training Conducted ML20042G2291990-05-0707 May 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 180,revising Tech Specs Re Fuel cycle-specific Parameters ML20042G2601990-05-0404 May 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 187,revising Tech Specs to Accommodate Implementation of 24-month Plant Refueling Cycle ML20042E9431990-04-20020 April 1990 Forwards Revised Epips,Consisting of Rev 7 to 9473-IMP-1300.01,Rev 4 to 9473-1300.11 & Rev 2 to 9473-ADM-1319.04.Deleted EPIPs Listed,Including Rev 3 to 9473-1300.19,Rev 2 to 9473-1300.21 & Rev 5 to 9473.1300.24 ML20042E6371990-04-16016 April 1990 Informs of Plans to Install Safety Grade Check Valve in Supply Line Inside Emergency Diesel Generator Fuel Tank Room Coincident W/Replacement or Repair of Emergency Diesel Generator Fuel Oil Tank ML20042E5001990-04-13013 April 1990 Forwards Rev 1 to Topical Rept 028, Oyster Creek Response to NRC Reg Guide 1.97. ML20012E8711990-03-28028 March 1990 Lists Property Insurance Coverage,Effective 900401,per 10CFR50.54(w)(2) ML20012D4391990-03-19019 March 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 186,allowing Idle Recirculation Loop to Be Isolated During Power Operation by Closing Suction,Discharge & Bypass Valves ML20012B6781990-03-0202 March 1990 Requests Exemption of Specified Local Leak Rate Test Intervals to Include Next Plant Refueling Outage Scheduled for Jan 1991,per 10CFR50,App J ML20012A1501990-02-23023 February 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 184,removing 3.25 Limit on Extending Surveillance Intervals,Per Generic Ltr 89-14 ML20011F2571990-02-21021 February 1990 Advises That 891003 Request for Appropriate Tech Specs for Chlorine Detection Re Control Room Habitability,Not Warranted ML20006G0101990-02-21021 February 1990 Discusses 900110 Meeting W/Nrr Re 13R Insp Criteria for RWCU Welds Outboard of Second Containment Isolation Valve. All Welds Required 100% Radiography Based on Review of Piping Spec.Response to Generic Ltr 88-01 Will Be Revised ML20006F5931990-02-20020 February 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 177.Amend Changes Tech Spec 4.7.B to Include Battery Svc Test Every Refueling Outage & Mod of Frequency of Existing Battery Performance Test ML20011F6641990-02-20020 February 1990 Responds to NRC 900122 Notice of Violation & Forwards Payment of Civil Penalty in Amount of $25,000.Corrective Actions:Change Made to Sys Component Lineup Sheets in 125- Volt Dc Operating Procedure to Include Selector Switches ML20006F9181990-02-15015 February 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 183,permitting No Limitation on Number of Inoperable Position Indicators for 16 ASME Code Safety Valves During Power Operation ML20006D2551990-01-30030 January 1990 Forwards Response to Generic Ltr 89-13 Re Plant Svc Water Sys.Insp Program for Intake Structure at Plant Implemented During Past Two Refueling Outages & Emergency Svc Water Currently Chlorinated to Prevent Biofouling ML19354E8571990-01-24024 January 1990 Forwards Omitted Pages of 900116 Ltr Re State of Nj DEP Comments on Draft full-term OL SER & Clarification of Page 10,fourth Paragraph on New Seismic Floor Response Spectra ML20006B2171990-01-23023 January 1990 Responds to Unresolved Items & Weaknesses Identified in Insp Rept 50-219/89-80.Corrective Actions:Procedure Re Containment Spray sys-diagnostic & Restoration Actions Revised to Stand Alone Re Installation of Jumpers ML19354E3891990-01-19019 January 1990 Responds to Violations Noted in Insp Rept 50-219/89-27. Corrective Actions:Procedure 108 Revised to Allow Temporarily Lifting of Temporary Variation ML19354E8441990-01-19019 January 1990 Forwards Revised Tech Spec Table 4.13-1, Accident Monitoring Instrumentation Surveillance Requirements, in Support of Licensee 890630 Tech Spec Change Request 179,per NRC Project Manager Request ML20006B2881990-01-18018 January 1990 Forwards Results from Feedwater Nozzle Exam,In Accordance w/NUREG-0619 Insp Format ML20005G8161990-01-16016 January 1990 Provides Assessment of State of Nj Concerns Re full-term OL Plant,Per NRC 891222 Request.Comments Did Not Raise Any Concerns That Refute Conclusions Reached by NRC That Facility Will Continue to Operate W/O Endangering Safety ML19354D8281990-01-15015 January 1990 Responds to Violation Noted in Insp Rept 50-219/89-21. Corrective Action:Procedure A000-WMS-1220.08, Mcf Job Order Revised to Provide Detailed Guidance for Performance of Immediate Maint ML20042D4891989-12-28028 December 1989 Forwards Response to Generic Ltr 89-10, Safety-Related Motor-Operated Valve Testing & Surveillance, Fulfilling 6-month Reporting Requirement ML20005E1401989-12-22022 December 1989 Forwards Integrated Schedule Semiannual Update for Dec 1989 1990-09-05
[Table view] |
Text
_ _ _ _ - _ _ - _ _
GPU Nuclear Corporation Nuclear
- =:r388 Forked R.ver, New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:
March 16, 1984 Mr. Dennis M. Crutenfield, Chief Operating Reactors Branch #5 U. S. Nuclear Regulatory Commission Washington, D. C. 20055
Dear Mr. Crutchfield:
Subject:
Oyster Creek Nuclear Generating Station (OCNGS)
Docket No. 50-219 Environmental Qualification of Electrical BIuipment During our meeting concerning the subject matter on December 7,1983, the NRC staff requested UPUN representatives to provide:
- 1. Resolution of environmental qualification deficiencies at OCNGS identified in a Technical Evalution Report (TEit), prepared by Franklin Research Center under contract to the NRC.
- 2. Justification for continued op3 ration (JCO) for those items to be replaced, modiiied or evaluated after the current Cycle 10 refueling outage.
- 3. A list of TMI action plan items (NUREG 0737) to be installed by the end of the current (Cycle 10) refueling outage and identification of their qualification documentation.
- 4. Confirmation that all design-basis events at OCNGS which could result in a potentially harsh environment, including flooding outside containment, were addressed in identifying safety related electrical equipment.
- 5. Description of the method used to identify electrical equipment within the scope of paragraph (b)(2) of 10CFP.50.49 (i.e., "Nonsafety-related electrical equipment whose failure under postulated environm7ntal conditions could prevent satisfactory accomplishment of safety functions
.....").
8403220012 840316 . GNCW PDR ADOCK 05000219 .
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GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
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- 6. Identification of electrical equipment and their q.talification documentation for post-accident monitoring system as required by R~julatory Guide 1.97.
- 7. Reasons for extension of ceplacement or modification to the second refueling outage after March,1982.
Our responses to these items are provided below.
Resolution of Deficiencies Identified in TER Attachment I to this letter contains a table which lists proposed solutions for deficiencies identified by the TER for various safety-related components.
The table was reviewed in taa December 7, 1983 meeting. It has been revised to incorporate caninents generated in the meeting and additional information requested by th0 NRC staff.
JustificationforContinuedOperation(JC.i .
JCO's for those safety related equig ent in a harsh environment without complete qualification documentation were provided in Chapter 7 of our report to the NRC dated November 1, 1980. Subsequently, the JCO's were rcviewed by Franklin Research Canter and were found to be adequate. By our letter dated March 16, 1983 GPUN transmitted the revised Chapter 7 with additional JCO's.
The additional JCO's were reviewed in the December 7, 1983 meeting and the NRC staff requested further clarification. W e aforementioned TER concludes that some of the eqaipment which we considered to be qualified, still lacks complete documentation. The staff, therefore, requested JCO's for these equipments. Attachment II to this letter prcvides the required JCO's for your review.
TMI Action Plan Items (NUREG 0737)
GPUN letter dated October 23, 1981 listed the electrical equipment originally scheduled to be installed during the current. Cycle JO refueling outage in compliance with the NUREG 0737. W e letter also listed the clectrical equipment installed during the previous outage. Attachment III provides updated status of the NUREG 0737 items that were previously reported in the October 23, 1981 letter. JCO's for operational equipment without complete documentation are given in Attachment II.
DesignBasisEvents(DBEl Oyster Creek FDSAR analysis of the DBE IDCA shows that peak drywell temperatures do not exceed 285 'F. Wis is for a double-er;3ed rupture of a recirculation line with containment spray. The peak containment pressure is about 38 psig which is the saturation pressure at the peak containment temperature. n ese values are the highest containment temperature and pressure for breaks which occur below the core mixture level. Higher temperature and pressure in a containment atmosphere could be reached when the blowdown is pure steam, since the increased heat capacity of a droplet laden atmosphere reduces superheat conditions. On this basis GPUN performed plant specific analyses for steam line breaks above the core mixture !evel in determining the containtant temperature and pressure profile for environmental qualification of equignent. Methodology and results of the analy.as are provided in Chapter 2 of Environmental Qualification Report dated November 1, 1980 which was sutxaltted to you.
Plant specific high energy line break accidents outside containment were also evaluated to identify safety-related equipment which may te exposed to a harsh environment. Detailed discussion and results of the evaluation are given in Chapter 3 of the November 1, 1980 report. An ana]ysis was performed to evaluate the reactor building flood lewls for the oyster Creek Nuclear Generating Station following a high energy line break outside cotainment.
Description of the analysis and results are included in Cnapter 4 of the November 1, 1980 report. Werefore, the design-basis events and high ener.gy line break accidents at OCNGS were considered in the identification of the saftey related equignents which are essential to mitigate the postulated accidents and to achieve cold shutdown.
Affect of Nonsafety Equignent on Safety BIuipment JCP&I/GPU letter dated October 5,1979 in response to IE Information Notice 79-22 states that our evaluation of interactions between nonsafety systems and safety systems did not identify any adverse impact which would increase the consequences of any accidents analyzed in the FDSAR.
In addition, GPUN plans to conduct verification of proper selective coordination of protective devices or circuit breakers and fuses on vital i
buses to ensure that an electrical fault developed in nonsafety systems due to.
harsh envirotraent will not be transmitted to the safe shutdown systems. mis work will be conducted in the first half of the 1984 calendar year as part of the evaluation for the Fire Protection Program (10CFR50 Appendix R).
Isolation of the reactor protection system from nonsafety systems has been reviewed by the MC staff also in the Systematic Evaluation Program (SEP) for OCNC6 (SEP 'Ibpic No. VII-1A).
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Fosh Accident Monitoring Equignent (Reg. Guide 1.97, Rev. 2) l As indicated in Attachment III (NUREG 0737 Item II.B.3), four new isolation valves will be installed to mee[ the Post Accident Sampling capability requirement. Other modifications required by Reg. Guide 1.97 are still under investigation. As soon as the type of modification and electrical equipent involved are determined, GPUN will transmit SCEW sheet for those equipments.
Replacement Schedule Section (g) of 10CFR50.49 requires licensees to qualify the electric equipment in harsh enviror. ment by the end of the second refueling outage after March 31, 1982 or by March 31, 1985, whichever is earlier. Tne section (g) also states, "Tne Director of the Office of Nuclear Reactor Regulation may grant requests for extensions of this deadline to a date no later than November 30, 1985, As we have indicated in our previous submittals and in the December 7, 1983 meeting with your staff, we plan to replace partc and components without complete documentation during the second refueling outage after March 31, 1982. The second refueling outage will most likely take place after the March 31, 1985 deadline. However, every effort will be made to replace by Maren 31, 1985 those parts and components whose replacement activity does not place plant operation in an unsafe condition .
Very truly yours,
$ A j' Y H 'f)A&
er a. Fledler Vice President and Director Oyster Creek 1r/0122e cc: Administrator Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pa. 19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, N. J. 08731
Pags.1 of 8 ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES O.C.
NRC PP.O POSED TEE CAT. DEFICIENCY SGLUTION NO.- (X)MPONENT 1,3,8 Motorized Valve Actuator II.A Documentation, Similarity, Qustified Units will be replaced this 9,10 V-17-54, V-16-1, V-17-19 Life, Spray, Steam Exposure, Radiation outage w/ Qualified Limitorque
.V-14-30, 31, 32, 33, 34, Model.
35, 36, 37 2,6,7_ Matorized Valve Actuator II.A Documentation, Qualified Life, Limitorque's Report fB0058, V-20-15, 21,<40, 41 Radia tion which will be referenced in the V-21-1, 3, 5, 7, 9, 11 SCEW sheet, describes the generic qualification and qualified life of subject actuator. Limitorque's Report
' d'J 0003 includes Radiation exposures.
II.C Qualified Life V-16-7 & 14 will be replaced 4 Motorized Valve Actuator V-16-2, 14 . this outage 'w/ Qualified Limit-orque Model.
V-16-61 V-16-61 is equipped w/ Reliance Class B Motor, instead of Peer-less Class B as mentioned in the TER. Limitorque's Report fB0058, which will be referenced ir t' e SCEW sheet, addresses th qualified life of subject actuator.
'S- ' Motorized Valve Actuator II.A Qualified Life, Radiation V-17-1,2,3 will be replaced this
-V-17-1, 2, 3 outage w/ Qualified Limitorque V-17-55, 56, 57 Meiel.
V-17-55,56,57 requires additional Documentation for Peerless Class B DC Motor.
' See Attachment 11 for JCO.
Page 2 of @ ,
ATTACHfElT 1 SPECIFIC EQUIPMENP EQ DEFICIENCIES O.C.
TER NRC PROPOSED NO. COMPCNENT CAT. DSFICIENCY SOLUTION 11 Motorized Valve Actuator II.A Documents. ion, Qualified Life, Spray Actuators are equipped with V-5-166, 148 Reliance Class RH A.C. Motors.
Limitorque's Report #B0058, which will be referenced in the SCEW sheet describes the generic qualification and qualification tests. Chemical Solution (Sodium Chromate) will be replaced w/ demineralized water.
12 Motorized Valve Actuator II.A Documentation, Similarity This motor is a Reliance Class V-1-106, 107 RH instead of Class B as men-tioned in the TER. Limitorques Report #'s 600456 and 80058 will ba referenced in the SCEW sheet.
13,15 Motorized Valve Actuator II.A ' Qualified Life, Radiation Additional Documentation for V-21,13, 17 Peerless Class B D.C. Motor is required. See Attachment II for JCO.
'14 Motorized Valve Actuator II.C Qualified Life Limltorque's Report 180058, V-5-167, 147 which will be referenced in the SCEW Sheet, addresses the qualified life of subject actuator.
16,17,18 Solenoid Valve II.C Qualified Life ASCO Report #AQR-67368/Rev. O, 19,20 .V-24-29 which ' vill be referenced in the 4 NS03A, B SCEW Si.eet, addresses the NSO4AL1, L2, L3 qualified life of subject valve..
NSO4BL1, L2, L3 o -
ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES 0.C.
TER NRC PRO POSED
- NO. COMPONENT CAT. DEFICIENCY SOLUTION 21 Solenoid Valve - 1.B Evaluation of Aging Degradation, Replaced Age Sensitive Non-V-38-9, 10, 16, 17 Qualified Life, Program to Identify Metallic Parts this outage.
Aging Degrada t ion
- 27 Solenoid Valve II.C Qualified Life lhe qualified life was covered
- V-26-16,-'18 in reference document, Wyle Report No. 17451-13.TER evalu-ation (on page 5f) does not' apply.
-22,23,24- Solenoid Valve 1.B Documentation Replace w/ Qualified Valve next 25,26,28 -V-23-13, 15, 16, 17, 18 outage.
31,35,36 19, 20,- 21, 22 V-27-1, 2, 3, 4 V-28-17, .18, 4 7 29,30,32 Solenoid Valve 1.B Documentation Replaced this outage with -
33,34 V-38-22, 23 Qua1ified ASCO Va1ve.
V-31-2.
V-24 V-22-1, 2,,28,- 29 V-11-34,--36 37 Solenoid Valve II.A Eva lua tion of Aging Degrada tion, Additional Documentation NR-108 A,B,C,D,E Qualified Life, Radiation and eva lua tion.
38,39,40 Pressure, Level, Flow 1.B Documenta t i on Replace with Qualified
- 41,42,43 Tra nsmi t te r Transmit ter next outage.
44,45 - ID46 A,B; ID45
- IP05'A,B,C,D; IAl2 ID13 A,B; IP07 IG06 A,B; RV26 A,B
- IP03 A,B 1
6 . - .__..__.m.-
~
. + Page 4 of 8 ATTACHMENT I -
SPECIFIC EQUIPMENT EQ DEFICIENCIES 0.C.
TER NRC PR0l'OSED NO. COMPONENT CAT. DEFICIENCY SOLUTION 46,48,49 Pressure, Level Switch I.B Documentation Replace with Qualified 50,53,54 'RE23 A, B,C,D ; RE17 A, B ,C,0 Analog Trip System next -
55,56,57 REIS A,B,C,D ; RE03 A,B,C,D outage 58,60,61 REOS A,B; RE22 A,B,C,D,E
( 62,63 RE18 A,B,C,D; RE05-19 A,B l RE02 A, B, C,D ; RV46 A, B,C,D
( IA83 A,B,C,D,E; IP15 A,B,C,D IBil-Al,A2,B1,B2;
( IB05-Al,A2,B1,B2 I
47 Pressure Suitch II.A Evaluation of Aging Degradation, Environmental Qualification not PS-153 Qualified Life, Rad ia t ion required. The automatic bypass i design feature in the control I I logic for venting the drywell'
! on high N2 Pressure was elimi- l l
nated by replacing the existing l switch with a spring return-switch. This was done in order to ensure that deliberate oper-ator action is required to open the drywell vent valves after l
L the isolation signal has been I reset.
51,52 Pressure Switch .I.B Documentation Replace w/ Qualified Pressure RV29 A,B,C,D Switch next outage.
RV40 A,B,C,D 1 59 Level Switch 1.B -Documentation Replaced this outage w/Qualifici, RD08 A,B,C,D,E,F Level Switch.
64 - Limit Switch I.B Documentation Replaced this outage w/Qualifie'.
NSO4A-1,-2 Limit Switch.
NSO4B-1,-2 l
l)
Page 5 of 8 ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES O.C.
TSR NRC PROPOSED NO. COMPONENT CAT. DEFICIENCY SOLUTION 65 Temperature Swit.ch II.A Similarity, Adequate Pressure, We referenced document, Wyle IB10 A thru P Adequate Steam Exposure Reports No. 17451-18 and No.
43854-1, provides all the neces-sary information which estab-lished the specific relationship between the tested switch model '
and installed unit.
In addition, recent Wyle letter (which will be referenced in the SCEW sheet)-outlined the applicable portions and photo-graphs from the referenced re-ports about the similarity of the switch and verified that subject switch's body and wirings were exposed to steam environment during the I.OCA testing. A re-evaluation of the pressure requirements in the Pain Steam 'Iunnel area, where IB-10E thru P switches are located, reveals that the pressure will be about 18.2 psia following accident con-dition. his information is found in EDS Report No. 02-0990-1085, Rev. O dated July 1981, wil] be referenced in the SCEW Sheet.
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Page 6 of 8 ATTACriMENT 1 ,
SPECIFIC EQUIPMENT EQ DEFICIENCIES O.C.
TER NRC PROPOSED NO. COMPONENT CAT. DEFICIE!CY SOLUTION 66 Temperature Switch I.B Documentation In 1979, Jersey Central Power IB06A,B,C,D and Light 03mpany (JCPEL) per-formed a high energy line break (HELB) analysis of the Oyster Creek Emergency (bndenser System (ECS) piping outside containment and concluded that a pipe break could cause damage to the ECS isolation valves and controls. JCP&L provided thia conclusion to the NRC. The NRC (SEP Branch) performed an on-site inspection as a part of the evaluation of the SEP Tapic No. III-5B, confirmed JCP&L's findings, and requested JCP&L to provide adequate protection against the effects of a postulated HELB. GPUN (now licensed operator for the OCNGS) performed and submitted an analysis to demonstrate that the ECS piping would leak before a significaat break could occur. During the integrated assessment of the SEP topics, the NRC staff accepted the result of the analysis and requested GPUN to provide leak detection capability (i.e., visual inspections or automated devices). GPUN is currently evaluating the type of the capability to be utilized.
Since GPUN will provide improved capability of an early ECS leak detection, the subject temperature switch is no longer needed.
Page 7 of @
A'RTACHMEft' 1 SPECIFIC EQUIPMElff EQ DEFICIENCIES 0.C.
TER NRC PROPOSED NO. COMPONENT CAT. DEFICIENCY SOLUTION 67 Electric Motor II.C Qualified Life Inbrication changeout is NZ01-A,B,C,D covered by Plant PM program.
70 Electric Motor II.A Steam Exposure Motors are located inside the 1-1 corner rocus. Tnere are no 1-2 steam lines in corner rooms.
1-3 These motors will not be ex-1-4 posed to steam environment.
TER evaluation (page SF) does not apply.
71 Terminal Block I.B Documentation, Evaluation of Aging Replaced with Qualified Block TB#'s63-242 Degradation, Qualified Life, steam this outage.63-246 Exposure, Spray, Radiation 22-389 22-390 71-412 22-640 22-641 72 Electrical Penetration II.A Aging Degradation, Qualified Life Additional Analysis and X10, X13, X18 Spray Radiation documentation.
73,74 Motor Control Centers II.A Documentation pdditional Analysis.
DC-1, lA21B, LAB 2, 1821A, 1821B ,
75 fetor (bntrol Center II.A Documentation This equipment is located DC-2. in floor El. 75' in about the same area where the temperature switches under TER No, 66 are located. Therefore, the pro-posed solution described under TER No. 66 also applies.
5 Page 8 of 8 :
ATTACHMENT 1 SPECIFIC EQUIPMENT EQ DEFICIENCIES 0.C.
NRC PROPOSED '
TER
' CAT. DEFICIENCY SOLUTION NO. COMPONENT II.A Documentation The Rockbestos Report No.
80 Electrical Cable Rockbestos Firewall EP QA 1804 will be referenced in the SCEW sheet. ,
i II.A Locumentation, Similarity, Aging The E.I. Dupont Report dated j 81 Electrical Cable Tensolite Deg ra da t i on, Qualified Life, Radiation August 1, 1974, " Tests of Elec-tric Cables Insulated and '
Jacketed With TEFZEL 280 Flouropolyuer Under IEEE 383-1974," will be referenced in the SCEW Sheet.
Documentation Franklins Report d's F-C2770
, 82 ,
Elect rical Cable II.A dated October 1970 and F-C2737 t Kerite dated April 15, 1970 will be l referenced in the SCEW Sheet.
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N/A Temperature Switch N/A See Attachment II for JCO.
IP18A IP18B 4
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(014 7M)
ATTACINEN T II ~l I
JUSTIFICATION OF WNTINUED OPERATION TER NO .
V-17-3 5 Shutdown Cooling Valves V-17-5 6 V-17-57
'lhese valves are installed a t the discharge side of the shutdown heat exchangers respectively. If these valves should fail, the alternate shutdown cooling me thad is to direc t the flew to the main condenser by way of the cleante system lytdown and maintaining reactor water level using the ccadensate system via the feedwater string (s) to makeup to the reactor vessel, Currently, anotiwr shutdown cooling method is being developed whi ch will include the Elec tromatic Relie f Valve s. If this me thod becanes official and supersedes the preceding, we chall advise accordingly.
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TER No. 13, 15 l V-21-13 Containment Spray Test Valves ,
V-21-17 j These are the Containment Spray System dynamic test valves under each system loop respectively. They are normally closed and are used for full-flow '
testing of the Containment Spray System without wetting the drywell. The flow from these valves is directed to the suppression chamber through connections into the vacuum breaker line.
The operating circuits of these valves are interlocked so that both loops can not be in the test position at the same time. .Each loop has the heat removal capacity to hold primary containment pressure below the design
, pressure of 62 psig, and to reduce the pressure to essentially atmospheric within about 8 days following the accident. Also, the control circuitry for the containment spray system provides automatic reset from dynamic test to automatic start readiness when process conditions indicate impending need for the containment spray.
If any of the test valves fail to close while the containment spray system is on, the net flow of that loop will be reduced by about 20%. However, the flow of one containment spray loop in either loop is more than ample to provide the necessary heat removal capacity.
(0147M)
. . . -. . = .. -
ADDITIONAL ITIM (not covered by TER)
I P A Contalment Spray Temperature Switches IP-18-B he containment spray system consists of two independent full-capacity cooling loops. Ei ther loop is capable of removing the fission product decay heat fro:n the contalment.
he containment spray temperature switches, located outside the Drywell a t dif ferent locations, are used to sense the water temperature from the hea t exdiangers in loops I and II respectively, as a failure monitor. Failure of the temperature switch of start loop will autanatically disable the contaiment spray pump of that loop. We contairinent spray pumps of the standby loop can be manually started using their respective controls'in the Control Roan.
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- ADDITIONAL ITEM (Not covered by TER)
IUREG 0737 Item II.D.3
- Direct Indication of Relief and Safety Valve Position 2
I j' These acoustic monitoring systems were operational at the OCNGS prior to this IUREG requirement which includes environmental qualification. The environmental gitalification will be established by the B&W Owner's Group Test Program, in which the OCNGS is a participant.
The testing has been concluded and the final report and recommendations are scheduled for release in the first quarter of 1984. The results of the
, testing show that the basic components are satisfactory for in containment IDCA and MSLB. However, some additional protection of the driver / amplifier and connections to it probably will be required at the OCNGS.
4 For the relief valves there are also position indicating switches with Control Room displays. For the safety valves there are tenperature readouts t
In the Reactor Building which indicate if a safety valve has lifted.
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ADDITIONAL ITEM (not covered by TER)
This is in addition to previously submitted JCO No. 51 - Torus Vacuum Relief System.
a) IP-12 Pressure Transmitters PT-52 Wese transmitters are physically located in different locations outside the Drywell. The cables and conduits run separately, and power supply to these transmitters comes from two different sources.
b) DPS-66A Pressure Switch DPS-66B 2ese switches control vacuum breaker valves V-26-16 and V-26-18 respectively to prevent excessiva vacuum in the suppression chartber.
If both switches should fail under a single failure situation, the Control Room Operator can monitor the pressure in the suppression chamber with the Torus Pressure indicatiaa device (IP13) and operate the vacuum breaker valves manually with the Control Switch at the llF panel.
ATTACHMENT III TMI ACT70N PLAN, NURF 0737 ITEMS NUREG 0737 Item TI.B.3 Post Accident Sampling Capability Of the four solenoid valves :equired, one, V-40-6, has been installed, but is not operable. The remaining valves will be installed consirtent with our post accident system operational comitment.. These four valves are environmentally qualified and SCEW Gnesta will be submitted.
NUREG 0737 Item II.D.3 Direct Indication of Relief and Safety Valve Position The environmental qualification will be established by the B&W Owner's Group Test Program, in which OCNGS is a participant.
The testing has been concluded and the final report and recomendation are scheduled for release in the first quarter of 1984. The re.3ults of the testing show that the basic components are satisfactory for in containment IOCA and MSLB. However, some additional protection of the driver / amplifier and connections to it probably will be required at the OCNGS.
NUREG 0737 Item II.E.4.1
< Dedicated Hydrogen Pet.etrations GPUN required exemption of this item. Refer to the letter dated December 15, 1902.
NUREG 0737 Item II.F.1 Subparts (1) through (6)
Additional Accident Monitoring Instrumentation Rese items will be operable by the end of the Cycle 10 refuel 4ng outage.
They are fully qualified and the applicable Sch:W sheets will be submitted.
NUREG 0737 Item II.K.3.19 Interlock on Recirculation Pump toops mis item will be installed during Cycle 11 refueling outage.
.