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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M2101999-10-19019 October 1999 Forwards NRC Rept Number 17, Requal Tracking Rept from Operator Licensing Tracking Sys.Rept Was Used by NRC to Schedule Requalification Exam for Operators & Record Requal Pass Dates ML18107A5601999-10-18018 October 1999 Submits 30-day Fuel Clad Temp Rept,Iaw 10CFR50.46.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Salem Generating Station Large & Small Break LOCA Analyses ML20217K4391999-10-18018 October 1999 Provides Response to RAI to Support USI A-46 Program Submittal for North Anna Power Station,Units 1 & 2.Rev 10 to BNL Rept 52361,encl ML20217H3301999-10-14014 October 1999 Forwards Rev 0 to COLR for North Anna 2 Cycle 14 Pattern Su. No New Commitments Intended by Ltr ML18152B3541999-10-12012 October 1999 Requests Use of Code Case N-532 & N-619,per Provisions of 10CFR50.55a(a)(3).Detailed Info Supporting Request,Encl. Attachment 2 Includes Technical White Paper That Provides Further Technical Info ML18107A5531999-10-0808 October 1999 Forwards Summary Rept of Plant Startup & Power Ascension Testing for Sgs,Unit 2 Cycle 11,per Requirements of TS 6.9.1.1 ML18107A5561999-10-0707 October 1999 Requests Relief Associated with Containment Examinations at Hope Creek & Salem Generating Stations.Attachment 1 Includes Proposed Alternatives & Supporting Justification for Relief Requests ML18107A5501999-10-0505 October 1999 Provides Current Status of Pse&G Actions Re GL 98-01, Y2K Readiness of Computer Sys at Npps, for Salem Nuclear Generating Station,Units 1 & 2 & Hope Creek Nuclear Generating Station ML18107A5521999-10-0505 October 1999 Encourages NRC to Support Abb Combustion Engineering Nuclear Power Request for Priority Review of Generic TR Re Crossflow Ultrasonic Flow Measurement Sys ML18107A5591999-10-0505 October 1999 Informs That Nj Dept of Environ Protection Has No Comments on License Change Request S99-07 for Sgs,Units 1 & 2 ML20216K1681999-10-0101 October 1999 Forwards Vols I-VIII of Rev 35 to UFSAR for Naps.Rev Also Includes Update to Chapter 17 of Ufsar,Which Contains Operational QA Program.Changes to Program Description Do Not Reduce Commitments Contained Therein ML20212J9101999-10-0101 October 1999 Forwards SE Accepting Licensee 990916 & 27 Relief Requests IWE-3 for Plant.Se Addresses Only IWE-3 Due to Util Urgent Need for Relief.Requests IWE-7 & IWE-8 Will Be Addressed at Later Date IR 05000272/19990071999-09-28028 September 1999 Forwards Insp Repts 50-272/99-07 & 50-306/99-07 on 990721- 0831.One Potentially Safety Significant Issue Identified Dealing with Control Room Special Ventilation System.Four Addl Issues of Low Safety Significance Identified ML18152B3391999-09-27027 September 1999 Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC ML20212G5091999-09-22022 September 1999 Forwards in Triplicate,Applications for Renewal of Licenses for Listed Individuals.Encls Withheld,Per 10CFR2.790(a)(6) ML18107A5341999-09-22022 September 1999 Provides Data Re Operator Licensing Exam for Salem & Hope Creek Station,In Response to NRC Form 536 (7-1999) ML18107A5421999-09-22022 September 1999 Forwards Discharge Monitoring Rept for Salem Generating Station for Aug 1999.Rept Is Required by & Prepared Specifically for EPA & Nj Dept of Environ Protection ML18152B6671999-09-17017 September 1999 Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03 ML18152B3331999-09-17017 September 1999 Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves. ML18152B3341999-09-16016 September 1999 Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations ML20212B3631999-09-14014 September 1999 Forwards Rev 13 to Salem - Hope Creek Security Plan,Iaw 10CFR50.54(p).Summary of Proposed Changes to Plan,Encl. Encl Withheld ML18107A5321999-09-13013 September 1999 Forwards Revised 10CFR50.92 Evaluation to Clarify Util Response to Question Number 1 Re Amend to Modify TS 3/4 8.1, AC Power Sources. ML18107A5331999-09-13013 September 1999 Provides Notification That PSEG Intends to Utilize ASME Code Case N-481 During Second ISI Interval at Sgs Units 1 & 2 ML18107A5351999-09-13013 September 1999 Informs That NRC Has Reviewed Pse&G Request Proposing to Modify TS Which Allow EDG to Be Operated for 24 Surveillance Test During Any Mode,Iaw 10CFR50.91(b) & Has No Comments ML18107A5221999-09-0808 September 1999 Requests Approval to Use ASME Code Case N546,which Provides Alternative Qualification Requirements That Allow Personnel Most Familiar with Walkdown of Plant Sys,Like License Operators to Perform VT-2 Examinations ML18107A5231999-09-0808 September 1999 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1999. Rept Summarizes Liquid & Gaseous Releases & Solid Waste Shipments from Salem Generating Station for Period of Jan-June 1999 ML20211N2531999-09-0808 September 1999 Responds to Request to Exceed 60,000 Mwd/Mtu Lead Rod Burnup in Small Number of Fuel Rods in North Anna Unit 2.Informs That NRC Offers No Objection to Requested Use of Rods in Reconstituted Fuel Assembly.Se Supporting Request Encl ML20211L9151999-09-0101 September 1999 Forwards Response to NRC Request for Comments Re Closure of Review of Response to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML20211J2211999-08-31031 August 1999 Approves Request to Remove Augmented ISI (Aii) Program for RCS Bypass Lines from North Anna Licensing Basis.Se Re Request to Apply LBB to Eliminate Augmented Insp Program on RCS Bypass Lines Encl ML20211H4131999-08-27027 August 1999 Informs That Util Revised Encl Bases for TS 2.2.1, Reactor Trip Sys Instrumentation Setpoints, Discussing Steam Flow/ Feed Flow Mismatch Portion of Steam Flow/Feed Flow Mismatch & Low SG Water Level Reactor Trip Setpoint ML18107A4981999-08-26026 August 1999 Forwards Response to NOV That Resulted from Predecisional Enforcement Conference Conducted on 990624.Corrective Actions:Communications to Supervisors Reinforced Employee Right & Duty to Raise Nuclear Safety Issues ML18107A5061999-08-26026 August 1999 Provides First Feedback from Observation of NRC Insp Under Pilot Nuclear Power Plant Insp Program.Attached Are Completed Insp Feedback Forms for Procedure 71111, Attachment 21 & Procedure 71151 ML18107A5181999-08-26026 August 1999 Forwards Ninety Day Rept for ISI Activities Conducted at Sgs,Unit 2 During Ninth Extended Outage & Tenth Refueling Outage.List of Encl,Provided ML18107A5051999-08-24024 August 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Request IAW Requirements of 10CFR50.91(b) & Has No Comments Re Amend to FOL Change Request S99-02 to Modify TS Re Penetration Valves ML18107A4921999-08-23023 August 1999 Provides Suppl Info Re 971024 Amend Request to Modify TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Util Withdraws from Request All Proposed Changes Associated with Filter Testing,Per Issuance of GL 99-02 ML20138B3241999-08-23023 August 1999 Forwards Draft Response to Question 1 Re NAPS USI A-46 ML20211D9041999-08-20020 August 1999 Forwards Revised Pages to Third Ten Year ISI Program & Relief Requests, Replacing Pages in 990408 Submittal ML18107A4911999-08-20020 August 1999 Forwards Revised Plant Attribute Libraries for Salem & Hope Creek Generating Stations,Iaw 10CFR50,App E,Vi, Emergency Response Data Sys, 3.b.Changes Are Identified by Rev Bars ML18107A4861999-08-19019 August 1999 Forwards NPDES Discharge Monitoring Rept, for Salem Generating Station for Month of Jul 1999.Rept Required by & Prepared Specifically for EPA & Nj Dept of Environ Protection ML18107A4831999-08-18018 August 1999 Submits Licensee Comments on NRC 990730 Ltr Which Provided Notification of Close Out of TAC Numbers MA0567 & MA0568 Re GL 92-01,Rev 1,Suppl 1 ML20211B3871999-08-17017 August 1999 Requests Permission to Routinely Discharge from SW Reservior to Waste Heat Treatment Facility Under Existing Vpdes Permit Through Outfalls 108 & 103.Discharges Are Scheduled to Commence on 990907,due to High Priority Placed on Project ML20210T0671999-08-13013 August 1999 Informs of Completion of Review of Proposed Revs of Schedule for Withdrawal of Rv Surveillance Capsules Submitted by VEPCO on 981217.Approves Proposed Revs.Forwards Safety Evaluation ML18107A4801999-08-13013 August 1999 Requests That Pse&Gs Contact in NUREG-0383, Directory of Compliance for Radioactive Matl Packages, Be Changed ML20210Q9841999-08-12012 August 1999 Forwards Rev 1 to Vepc COLR for North Anna Unit 2,Cycle 13 Pattern Ud, Per TS 6.9.1.7.d.COLR Was Revised to Include Temp Coastdown Operation at End of Cycle 13 ML20210Q7661999-08-0606 August 1999 Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas ML18152B4081999-08-0606 August 1999 Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves. ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML20210N2921999-08-0505 August 1999 Discusses Which Submitted Proposed TSs Bases Change for Containment Leakage. Licensee Changes to Bases May Be Subj to Future Insps or Audits ML18107A4751999-08-0505 August 1999 Forwards Fitness for Duty Performance Data Rept for Six Month Period Ending 990630 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217K4391999-10-18018 October 1999 Provides Response to RAI to Support USI A-46 Program Submittal for North Anna Power Station,Units 1 & 2.Rev 10 to BNL Rept 52361,encl ML18107A5601999-10-18018 October 1999 Submits 30-day Fuel Clad Temp Rept,Iaw 10CFR50.46.Rept Describes Changes to Calculated Peak Clad Temp (PCT) for Salem Generating Station Large & Small Break LOCA Analyses ML20217H3301999-10-14014 October 1999 Forwards Rev 0 to COLR for North Anna 2 Cycle 14 Pattern Su. No New Commitments Intended by Ltr ML18152B3541999-10-12012 October 1999 Requests Use of Code Case N-532 & N-619,per Provisions of 10CFR50.55a(a)(3).Detailed Info Supporting Request,Encl. Attachment 2 Includes Technical White Paper That Provides Further Technical Info ML18107A5531999-10-0808 October 1999 Forwards Summary Rept of Plant Startup & Power Ascension Testing for Sgs,Unit 2 Cycle 11,per Requirements of TS 6.9.1.1 ML18107A5561999-10-0707 October 1999 Requests Relief Associated with Containment Examinations at Hope Creek & Salem Generating Stations.Attachment 1 Includes Proposed Alternatives & Supporting Justification for Relief Requests ML18107A5591999-10-0505 October 1999 Informs That Nj Dept of Environ Protection Has No Comments on License Change Request S99-07 for Sgs,Units 1 & 2 ML18107A5521999-10-0505 October 1999 Encourages NRC to Support Abb Combustion Engineering Nuclear Power Request for Priority Review of Generic TR Re Crossflow Ultrasonic Flow Measurement Sys ML18107A5501999-10-0505 October 1999 Provides Current Status of Pse&G Actions Re GL 98-01, Y2K Readiness of Computer Sys at Npps, for Salem Nuclear Generating Station,Units 1 & 2 & Hope Creek Nuclear Generating Station ML20216K1681999-10-0101 October 1999 Forwards Vols I-VIII of Rev 35 to UFSAR for Naps.Rev Also Includes Update to Chapter 17 of Ufsar,Which Contains Operational QA Program.Changes to Program Description Do Not Reduce Commitments Contained Therein ML18152B3391999-09-27027 September 1999 Forwards Revised Relief Request IWE-3,which Now Includes Addl Visual Exam Requirement After post-repair or Mod Pressure Testing Is Completed,Per Telcon with NRC ML20212G5091999-09-22022 September 1999 Forwards in Triplicate,Applications for Renewal of Licenses for Listed Individuals.Encls Withheld,Per 10CFR2.790(a)(6) ML18107A5341999-09-22022 September 1999 Provides Data Re Operator Licensing Exam for Salem & Hope Creek Station,In Response to NRC Form 536 (7-1999) ML18152B6671999-09-17017 September 1999 Forwards Two NRC Forms 536,containing Info on Proposed Site Specific Operator Licensing Exam Schedules & Estimated Number of Applicants Planning to Take Exams And/Or Gfes,In Response to NRC Administrative Ltr 99-03 ML18152B3331999-09-17017 September 1999 Forwards Revised 180-day Response to NRC GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves. ML18152B3341999-09-16016 September 1999 Requests Relief from Specific Requirements of Subsection IWE of 1992 Edition with 1992 Addenda of ASME Section XI Re Containment Liner Examination Requirements,For North Anna & Surry Power Stations ML20212B3631999-09-14014 September 1999 Forwards Rev 13 to Salem - Hope Creek Security Plan,Iaw 10CFR50.54(p).Summary of Proposed Changes to Plan,Encl. Encl Withheld ML18107A5321999-09-13013 September 1999 Forwards Revised 10CFR50.92 Evaluation to Clarify Util Response to Question Number 1 Re Amend to Modify TS 3/4 8.1, AC Power Sources. ML18107A5351999-09-13013 September 1999 Informs That NRC Has Reviewed Pse&G Request Proposing to Modify TS Which Allow EDG to Be Operated for 24 Surveillance Test During Any Mode,Iaw 10CFR50.91(b) & Has No Comments ML18107A5331999-09-13013 September 1999 Provides Notification That PSEG Intends to Utilize ASME Code Case N-481 During Second ISI Interval at Sgs Units 1 & 2 ML18107A5231999-09-0808 September 1999 Forwards Salem Generating Station Semiannual Radioactive Effluent Release Rept for Jan-June 1999. Rept Summarizes Liquid & Gaseous Releases & Solid Waste Shipments from Salem Generating Station for Period of Jan-June 1999 ML18107A5221999-09-0808 September 1999 Requests Approval to Use ASME Code Case N546,which Provides Alternative Qualification Requirements That Allow Personnel Most Familiar with Walkdown of Plant Sys,Like License Operators to Perform VT-2 Examinations ML18152B4471999-09-0101 September 1999 Requests That NRC Remove Listed Labels from Distribution ML20211L9151999-09-0101 September 1999 Forwards Response to NRC Request for Comments Re Closure of Review of Response to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity ML20211H4131999-08-27027 August 1999 Informs That Util Revised Encl Bases for TS 2.2.1, Reactor Trip Sys Instrumentation Setpoints, Discussing Steam Flow/ Feed Flow Mismatch Portion of Steam Flow/Feed Flow Mismatch & Low SG Water Level Reactor Trip Setpoint ML18107A4981999-08-26026 August 1999 Forwards Response to NOV That Resulted from Predecisional Enforcement Conference Conducted on 990624.Corrective Actions:Communications to Supervisors Reinforced Employee Right & Duty to Raise Nuclear Safety Issues ML18107A5181999-08-26026 August 1999 Forwards Ninety Day Rept for ISI Activities Conducted at Sgs,Unit 2 During Ninth Extended Outage & Tenth Refueling Outage.List of Encl,Provided ML18107A5061999-08-26026 August 1999 Provides First Feedback from Observation of NRC Insp Under Pilot Nuclear Power Plant Insp Program.Attached Are Completed Insp Feedback Forms for Procedure 71111, Attachment 21 & Procedure 71151 ML18107A5051999-08-24024 August 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Request IAW Requirements of 10CFR50.91(b) & Has No Comments Re Amend to FOL Change Request S99-02 to Modify TS Re Penetration Valves ML18107A4921999-08-23023 August 1999 Provides Suppl Info Re 971024 Amend Request to Modify TS 3/4 7.7, Auxiliary Bldg Exhaust Air Filtration Sys. Util Withdraws from Request All Proposed Changes Associated with Filter Testing,Per Issuance of GL 99-02 ML20138B3241999-08-23023 August 1999 Forwards Draft Response to Question 1 Re NAPS USI A-46 ML20211D9041999-08-20020 August 1999 Forwards Revised Pages to Third Ten Year ISI Program & Relief Requests, Replacing Pages in 990408 Submittal ML18107A4911999-08-20020 August 1999 Forwards Revised Plant Attribute Libraries for Salem & Hope Creek Generating Stations,Iaw 10CFR50,App E,Vi, Emergency Response Data Sys, 3.b.Changes Are Identified by Rev Bars ML18107A4831999-08-18018 August 1999 Submits Licensee Comments on NRC 990730 Ltr Which Provided Notification of Close Out of TAC Numbers MA0567 & MA0568 Re GL 92-01,Rev 1,Suppl 1 ML18107A4801999-08-13013 August 1999 Requests That Pse&Gs Contact in NUREG-0383, Directory of Compliance for Radioactive Matl Packages, Be Changed ML20210Q9841999-08-12012 August 1999 Forwards Rev 1 to Vepc COLR for North Anna Unit 2,Cycle 13 Pattern Ud, Per TS 6.9.1.7.d.COLR Was Revised to Include Temp Coastdown Operation at End of Cycle 13 ML20210Q7661999-08-0606 August 1999 Requests Exception to 10CFR50.4 Requirement to Provide Total of Twelve Paper Copies When Submitting Surry & North Anna UFSAR Updates.Seek Approval to Submit Only Signed Original & One CD-ROM Version,Per Conversation with J Skoczlas ML18152B4081999-08-0606 August 1999 Forwards Response to NRC 990520 & 0525 RAIs Re North Anna & Surry Responses to GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves. ML18152B4091999-08-0505 August 1999 Forwards Vepc semi-annual fitness-for-duty Program Performance Data Rept for 990101-990630,IAW 10CFR26.71(d) ML18107A4751999-08-0505 August 1999 Forwards Fitness for Duty Performance Data Rept for Six Month Period Ending 990630 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 ML20210J8861999-08-0202 August 1999 Provides Clarification to Commitment Made in Identifying Extent by Which Existing Plant Design Complied with RG 1.97,specifically Re Variable, Radiation Exposure Rate ML20210F6121999-07-28028 July 1999 Forwards Supplemental Info on Proposed Irradiation of Fuel Rods Beyond Current Lead Rod Burnup Limit,Documenting Info Provided During 990624 Meeting & Suppl Original Submittal ML18152B3971999-07-26026 July 1999 Provides Estimates of Licensing Actions Expected to Be Submitted in Fys 2000 & 2001,in Response to NRC AL 99-02 ML20209E7621999-07-0909 July 1999 Provides Addl Info to Justify Use of Less than One Gpm Detectable Leakage Rate to Establish Required Margin for Crack Stability in LBB Analysis,Per 980623 Application on Reactor Coolant Loop Bypass Lines 05000338/LER-1999-005, Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl1999-07-0808 July 1999 Forwards LER 99-005-00,IAW 10CFR50.73.Commitment Made by Util Encl ML18107A4431999-07-0606 July 1999 Informs That Nj Dept of Environ Protection Bureau of Nuclear Engineering Has Reviewed Pse&G License Change Request S98-17 Re Permissible Enrichment Values for New Fuel Storage & Has No Comments ML18152B4401999-07-0101 July 1999 Informs NRC That on 990511,Dominion Resources,Inc,Executed Amended & Restated Agreement & Plan of Merger with Consolidated Natural Gas Co ML18107A4181999-06-30030 June 1999 Submits Response to NRC Request for Info Re Y2K Readiness at Npps,Per GL 98-01,suppl 1.Disclosure Encl ML18107A4131999-06-25025 June 1999 Provides Further Clarification of Licensing & Design Basis for 125 Vdc Battery Margins for Sgs & HCGS for Meeting Station SBO & Loca/Loop Loading Requirements,Per Util 990426 Ltr & Discussion with NRC 1999-09-08
[Table view] Category:VENDOR/MANUFACTURER TO NRC
MONTHYEARML20043H9431990-04-27027 April 1990 Requests Withholding of Proprietary WCAP-12351, North Anna Unit 1 Evaluation for Tube Vibration Induced Fatigue, from Public Disclosure Per 10CFR2.790 ML18094B4141990-03-27027 March 1990 Outlines Plan & Basis for Plan to Update Steam Generator Tube Fatigue Evaluations Performed by Westinghouse in Response to NRC Bulletin 88-002 ML18094B2541990-01-0808 January 1990 Requests Info Re Procedures to Expedite NRC Qualification of Control Sys Products for Nuclear Power Plant Backfitting to Enable Vendor to Market Products in Us.Negotiations W/Pse&G Underway Re Purchase of Feedwater Control Sys ML20246D7521989-08-0404 August 1989 Requests That Proprietary Topical Rept WCAP-12349, North Anna Unit 1 Steam Generator Update Tube Bundle Structural Integrity Presentation, Be Withheld (Ref 10CFR2.790) ML20244B7031989-05-23023 May 1989 Requests That Rev 1 to North Anna Unit 1 890225 Steam Generator Leak Event Rept Be Withheld from Public Disclosure (Ref 10CFR2.790) ML20012D7581989-05-0303 May 1989 Requests That Proprietary WCAP-12265, North Anna Unit 2 Evaluation for Tube Vibration Induced Fatigue, Be Withheld from Public Disclosure (Ref 10CFR2.790) ML20245K7281989-04-19019 April 1989 Requests That Proprietary Steam Generator Leak Event Rept, Be Withheld from Public Disclosure,Per 10CFR2.790(b)(4) ML20155D1571988-08-12012 August 1988 Requests That Proprietary WCAP-11929, Safety Evaluation: Zirconium Base Advanced Cladding Matls Usage in North Anna Unit 1 Demonstration Fuel Assemblies, Be Withheld,Per 10CFR2.790(b)(4) NRC-87-3280, Forwards Type I & II Ltrs Re Steam Generator Tube Rupture Event,In Response to Request for Info on Generic Implications of Tube Rupture for Use in Preparing Presentation to NRC on 871109.Part 21 Related1987-10-28028 October 1987 Forwards Type I & II Ltrs Re Steam Generator Tube Rupture Event,In Response to Request for Info on Generic Implications of Tube Rupture for Use in Preparing Presentation to NRC on 871109.Part 21 Related ML20235K4141987-09-23023 September 1987 Requests Proprietary WCAP-11601, North Anna Unit 1 Steam Generator Tube Rupture & Remedial Actions Technical Evaluation, Be Withheld from Public Disclosure (Ref 10CFR2.790) NRC-87-3266, Forwards Proprietary WCAP-11601 & Nonproprietary WCAP-11602, North Anna Unit 1 Steam Generator Tube Rupture & Remedial Actions Technical Evaluation & Application for Withholding of Proprietary Info (Ref 10CFR2.790)1987-09-23023 September 1987 Forwards Proprietary WCAP-11601 & Nonproprietary WCAP-11602, North Anna Unit 1 Steam Generator Tube Rupture & Remedial Actions Technical Evaluation & Application for Withholding of Proprietary Info (Ref 10CFR2.790) NRC-87-3261, Forwards Proprietary North Anna 1 Steam Generator Tube Rupture & Remedial Actions Technical Evaluation. Rept Withheld (Ref 10CFR2.790)1987-09-13013 September 1987 Forwards Proprietary North Anna 1 Steam Generator Tube Rupture & Remedial Actions Technical Evaluation. Rept Withheld (Ref 10CFR2.790) ML18093A2401987-07-0808 July 1987 Forwards Proprietary Viewgraphs from 870709 Presentation to NRC Re Elimination of RTD Bypass Line,Per NRC Concerns During 870615 Telcon ML20207R9291987-03-11011 March 1987 Requests Withholding of Proprietary Info Re Plant Demonstration Fuel Assemblies from Public Disclosure (Ref 10CFR2.790) NRC-87-3208, Forwards Proprietary Info Re Plant Demonstration Fuel Assemblies.Encl Withheld (Ref 10CFR2.790)1987-03-11011 March 1987 Forwards Proprietary Info Re Plant Demonstration Fuel Assemblies.Encl Withheld (Ref 10CFR2.790) ML20205S4511987-03-10010 March 1987 Requests Withholding of Proprietary Viewgraphs from 870305 Meeting Re Steam Generator Tube Integrity from Public Disclosure Per 10CFR2.790.Affidavit Encl ML20210T0781987-02-0505 February 1987 Requests That Encl Proprietary Viewgraphs of Facility Demonstration Assembly Irradiation Program Be Withheld (Ref 10CFR2.790) NRC-87-3201, Forwards Nonproprietary & Proprietary Viewgraphs of Facility Demonstration Assembly Irradiation Program,Per NRC Request for Info Re Use of Advanced Fuel Rod Cladding.Proprietary Viewgraphs Withheld (Ref 10CFR2.790)1987-02-0505 February 1987 Forwards Nonproprietary & Proprietary Viewgraphs of Facility Demonstration Assembly Irradiation Program,Per NRC Request for Info Re Use of Advanced Fuel Rod Cladding.Proprietary Viewgraphs Withheld (Ref 10CFR2.790) NRC-86-3160, Submits Updated List of Plants Affected by Changes in ECCS Evaluation Model Described in Rev 1 to Addendum 3 to WCAP-9561-P1986-09-0808 September 1986 Submits Updated List of Plants Affected by Changes in ECCS Evaluation Model Described in Rev 1 to Addendum 3 to WCAP-9561-P ML20213D7921986-08-20020 August 1986 Requests Proprietary WCAP-11163, Technical Bases for Eliminating Large Primary Loop Pipe Rupture as Structural Design Basis for North Anna Units 1 & 2, Be Withheld (Ref 10CFR2.790) ML18092B1471986-05-0909 May 1986 Responds to NRC 860324 Ltr Re Deviations on Training of QC Inspectors Noted in Insp Repts 50-272/86-05 & 50-311/86-05. Corrective Actions:Personnel Trained Per IE Bulletin 79-19 ML18142A0481984-08-0707 August 1984 Requests Proprietary Comparison to Alternate Code Calculations Be Withheld (Ref 10CFR2.790).Affidavit Encl ML18089A5621984-03-28028 March 1984 Requests Withholding Proprietary Info from Public Disclosure Under Previously Submitted Encl 770406 Application for Withholding AW-77-18 & Affidavit Approved on 771028 ML18087A8041983-03-24024 March 1983 Responds to 830323 Request for Info Re 830113-18 Westinghouse Servicing of Breakers.Uv Trip Attachment Cleaned & Lubricated W/Calforex 78-A ML18087A7871983-03-22022 March 1983 Submits Updated Info Re Investigation of Reactor Trip Switchgear Malfunctions.Technical Bulletin Recommending Independent Testing of Undervoltage & Shunt Trip Attachments for Manual Reactor Trip Expected by 830325 ML18086B1011981-11-25025 November 1981 Authorizes Utilization of Encl 761201 Affidavit for Withholding Info from Public Disclosure in Support of Util Document Entitled, Reactor Actuation Sys Setpoint Methodology. ML18139B4851981-07-28028 July 1981 Requests That Vendor Proprietary Info Forwarded in Util 810724 Ltr Re High Pump Burnup Radiological Consequences Be Withheld (Ref 10CFR2.790).Authorizes Use of Original Affidavit AW-76-51 Dtd 761018 ML20040C2661980-12-23023 December 1980 Application for Withholding Proprietary Summary Rept: Westinghouse Reactor Vessel Level Instrumentation Sys for Monitoring Inadequate Core Cooling (Microprocessor Sys). ML19320A9631980-06-25025 June 1980 Lists Topical Repts for Which Responses to Outstanding Questions Will Be Provided to Allow SER to Proceed & to Avoid Delays in Approval of Full Power Operation.Includes WCAP-9226,-9230 & -9236 ML18085A9151980-05-22022 May 1980 Requests That Util Proprietary Info Re Environ Qualification of safety-related Equipment Be Withheld (Ref 10CFR2.790) ML18082A4921980-05-12012 May 1980 Forwards Schedule for Evaluation of Westinghouse Steam Generator Row One U Bends.Requests Delay of NRC Issuance of Generic Ltrs to near-term OL Plants Requiring Plugging of Row One Tubes,In Confirmation of 800415 Meeting ML19312D1861980-03-14014 March 1980 Forwards Turbine Disc Integrity Task Force Concensus Response to Generic Questions Contained in NRC 800225 Ltr Re Turbine Disc Integrity.Portions Withheld (Ref 10CFR2.790) ML19296C4721980-02-20020 February 1980 Forwards Corrected Copy of Matl Properties of Facility Disc 4 & Tables of A/Acr.Encls Withheld (Ref 10CFR2.790) ML19305B9591979-11-0707 November 1979 Discusses Undetectable Failure in Engineered Safety Features Actuation Sys.Failure of P-4 Permissive Circuit in Both Redundant Protection Trains Could Result in Failure of Sys to Automatically Initiate Protective Function.Details Encl ML20125B9391979-11-0505 November 1979 Notifies That Westinghouse Briefed Utils Re Problems W/ Stress Corrosion Cracking in Westinghouse Low Pressure Rotors & Problems w/1,800 Rpm Low Pressure Turbines, Reportable Per 10CFR50.55(e) or 10CFR21.W/lists of Plants ML19268C0321979-10-0505 October 1979 Responds to IE Bulletin 79-03 Followup Rept.Data Developed on Generic Basis & Considered Applicable to All Snupps Units.Fabrication & Installation of Youngstown Pipe Spools Will Proceed W/All Matl Analyzed Prior to Use ML19209A3941979-09-10010 September 1979 Responds to IE Bulletin 79-15, Deep Draft Pump Deficiencies, Re safety-related Application,Mfg & Overall Dimensions ML18078A6711979-01-12012 January 1979 Forwards Westinghouse Rept Evaluation of the Reactor Coolant Sys Considering Subcompartment Pressurization Following a LOCA for Unit. ML18078A6871979-01-12012 January 1979 Forwards Fuel Grid Impact Loads for Salem Unit No 2 (Proprietary),Synopsis of WCAP-9283(nonpropietary).W/encl Applications for Withholding AW-79-04 & AW-77-27 ML18078A6921979-01-12012 January 1979 Forwards Fuel Grid Impact Loads for Salem Unit No 2 (Nonproprietary) ML18078A4801978-12-0101 December 1978 Forwards Proprietary & non-proprietary Reptdynamic Analysis of the Reactor Coolant Sys for Loss of Coolant Accidents:Salem Nuc Generating Stations I & II, Affidavit for Withholding & Appl for Withholding ML20150D3441978-11-28028 November 1978 Advises NRC That Due to Extension of Date for FSAR Submission,B&W Will Defer Submission of Revs to Topical Repts BAW-10026 & 10026P Reactor Vessel Model Flow Tests for 145 Fuel Assembly Cores ML18085A9171976-08-27027 August 1976 Requests That Proprietary Info Re Equipment Qualification Programs & Thermal Environ Qualification Curve Be Withheld (Ref 10CFR2.790).Original Affidavit AW-76-39 Dtd 760903 Encl 1990-04-27
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Water Reactor
Westinghouse .
Electric Corporation Divisions s.355 .
Mnscurgh Pemsytvania 18.230
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' November 7, 1979 NS-TMA-2150 Mr. Victor Stello, Jr. ' -
Director -
Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555
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Dear Mr. Stello:
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Subject:
Undetectable Failure in Engineered Safety Features Actuation System As a result of our continuing reviews of systems important to safety, !
I Westinghouse has identified an undetectable failure which potentially could J exist in a circuit associated with Engineered Safeguards and which is required for reactor protection. i
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The specific circuit is described in the attachment. The design function of l the circuit-is a permissive to provide the operator, depending on plant con-ditions, the capability to manually reset and block Safety Injection.
A failure a'nalysis, which assumed a failure of the affected circuit in both of the redundant protection trains (per IEEE-379), showed that the system's
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ability to automatically initiate the protective function could be lost under certain conditions.
Despite the low probability of the events necessary to set up the conditions,
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the WRD Safety Review Committee concluded on Novemoer 6,1979, that the poten-tial loss of the protective function is reportable to the NRC under Title 10CFR Part 21 for operating plants and Title 10CFR50.55(e) for plants under construction.
Detailed information, affected plants and recor= ended corrective action is contained in the attachment. This information has already been communicated to the utility owners of the affected plants. .
Please refer any questions to Mr. D. H. Rawlins, the Manager of Safety Standards in the Westinghouse Nuclear Technology Division.
Very truly yours,
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X/4&nWJ/ .
> T. M. Anderson, Panager
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Nucle Nf'ty 0 partment FWM/TMA/bek Attachment ,
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Undetectable Failure in Engineered Safety Features Actuation System (ESFAS)
Design (refer to accompanying typical functional logic diagram)
The P-4 permissive is used to input the status (open or closed) of the
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Reactor Trip breakers to the Engineered Safety Features Actuation System (ESFAS). This P-4 permissive provides an interlock in the ESFAS to enable or defeat the capability to manually reset and block Safety Injection (SI).
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In operation, the initiation of SI instantly trips the reactor and simultaneously starts an electric timer. After a preset time interval, deter-ined by plant specific system analyses, the timer effectively returns system control to the operators for manual reset and block of SI in order to either begin ECCS switchov'er from the injection phase to
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the recirculation phase or terminate SI. The system permits manual reset and block of SI only if the P-4 permissive indicates that the trip breakers are open (i.e., the reactor is tripped).
During normal plant power operation, the P-4 permissive prevents manual actions whicti could electrically block SI.
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Imolementation
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The P-4 permissive is derived from a switch contact operated via a mechanical linkage within the reactor trip breaker. When the breakers move (open or. closed), the switch contact changes position. The contacts are hardwired to the ESFAS input logic which registers the trip breaker position to allow or prevent operator action as described above.
'Testino .
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During normal plant operation, ESFAS logic is required to be periodically l tested. On newer plants with the Solid : a Prot--tion System, this l l
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testing is performed via automatic self test circuits which verify system operability. On older plants with a relay logic protection system, this
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testing is performed manually.
1 In addition, the rea,ctor trip breakers are also periodically tested.'
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Potential Concern Currently, the tests described above do not provide for checking the operation of the P-4 contacts or the interconnecting wiring. Therefore, a potential failure of the P-4 contacts or in the wiring would be
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undetectable.
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IEEE 379 requires that in the case of undetectable failures either (1) ,
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provide revised test schemes to identify failures or redesign to
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eliminate them, or (2) in system failure analyses demonstrate that the safety function can be assured assuming both the undetectable failures ,
have occurred and a random single failure has also occurred.
The failure modes of the P-4 contacts are (1) contacts fail to close when the reactor trip breakers open, or (2) contacts fail to open when 9
the breakers are closed. Failure mode (1) could prevent the normal made of resetting and blocking SI and alter the sequence of switchover opera- '
tions from injection to recirculation phase. The consequences of failure l
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mode (2) are such that following a previous initiation of SI and manual i reset and block, the block of SI could remain following the reset of the reactor trip breakers and when the plant was returned to power.
- No credit can be taken for illuminated Control Board windows (lamp
! bulbs) which would alert the operators to the hazard since they are '
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not safety grade and are not implemented as such. l
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Affected Domestic Plants ,
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Operating Plants SSPS '
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D. C. Cook Units 1 and 2 Farley Unit 1 Beaver Valley Unit 1 Trojan
' Salem Unit 1 North Anna Unit 1
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Relay Logic Zion Units 1 and 2 .
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Prairie Island Units I and 2
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Kewaunee
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Indian Point Unit 3
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Non-Operating Plants
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SSPS Farley Unit 2 .
Byron Units 1 and 2
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Braidwood Units 1 and 2 Virgild.Succer ShearonHarrisdniks1,2,3and4
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McGuire Units 1 and 2 Catawba Units 1 and 2
- Beaver Valley Unit 2 Yogtle Units 1 and 2 Jamesport Units 1 and 2 O
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. SSPS Seabrook Units 1 and 1 M111 stone Unit 3 .
. Marble Hill Units 1 and 2 Diablo Canyon Units 1 and 2 .
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Salem Unit 2 SNUPPS Units
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Comanche Peak Units 1 and 2
. South Texas Project Units 1 and 2 Sequoyah Units 1 and 2 '
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North Anna Unit 2
- Watts Bar Units 1 and 2 ,
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Haven Units 1 and 2
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All other domestic plants are unaffected.
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Recommended Corrective Actions
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. A. Plants Using Reactor Tripped Signal in Safety Injection Reset
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Circuit of Engineered Safeguards Relay Racks
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Iion Units 1 and 2 Kewaunee -
Prairie Island Units 1 and Z Indian Point Unit 3
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In the Engineered Safeguards Relay Racks for the above plants, a reactor tripped signal (Reactor Trip Breaker RTA and Bypass Breaker BYA open for Train A and Reactor Trip Breaker RTB and Bypass Breaker
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BYB open for Train B) energizes Relay RTA in Train A and Relay RTB in Train B. These relays are located in the rear compartment of
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on-lina testing of the Safeguards Relay Racks. In addition to this l
l testing. it is necessary to verify that the relays are operated by .
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r the auxiliary switch contacts of the Reactor Trip Switchgear. ,
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T. During normal plant operation, immediately verify that relays RTA and RTB are deenergized.
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- 2. After each reactor trip operation, verify that relays RTA and RTB are energized. .
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- 3. After closing the reactor trip , breakers on plant startup, .
verify that relays RTA and RTB become deenergized.
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I'f verification shows a relay is not in the correct position, j 4,
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- check the interconnecting wires to the Reactor Trip Switchgear l
and the breaker auxiliary switch and cell switch contacts.
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- 5. Verification of the correct relay position can be made by visual observation of the relays. (For Indian Point Unit 3, verification is made by observing 11e test lamp " Reactor Trip
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Relay Rack.)
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NOTE 1:~ During on-line testing of the reactor trip breakers, relays I
RTA and RTB do not change position due to the closing of the bypass
- breaker for the test. Following on-line testing of the reactor
- trip breakers, observe that relays RTA and RTB remain energized.
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NOTE 2: The interconnecting wiring from the Engineered Safeguiwds Relay Racks to the Recctor Trip Switchgear for relays RTA and RTB can. be verified during normal plant operation. At the switchgear controi teminal bTocks, use a 0-150 volts de range voltmeter or multimeter to measure the voltage across the two tenninals con- *
- necting the switch contacts to the coil circuit of Relay RTA in the Train A Engineered Safeguards Relay Rack. A nominal 125 volts (dependent upon battery system voltage) reading should be indicated on the voltmeter. A zero reading indicates an open or short cir-
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cuit in the interconnecting wiring from the relay racks or closed switch contacts, requiring corrective action. Repeat the voltmeter measurement across the two terminals connecting the switch contacts
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to Relay RTB coil circuit in the Train B Engineered Safeguards
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Relay Racks. .
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Revise appropriate procedures to require the verification tests noted above following automatic or manual reactor trip. Repeat the tests following reclosure of the reactor trip breakers and prior to rod withdrawal.
. S. Byron /Braidwood/ Marble Hill
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- Assure the following test sequence is adopted for each train of SSPS, with the plant at, shutdown and the SSPS in Ncrmal Operation:
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T Place a impson Mode \ '260 multimeter in the 50 VDC range.
- 2. At the reactor trip switchgear, place the (+) lead on the
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tenninal leading to the SSPS, TB506-4.
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! 3. Place the (-) lead on the terminal leading to the SSPS,
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TB506-5. -
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- 4. The multimeter should read 0 VDC (nominal) with the reactor
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trip breaker tripped open.
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5 This indicates either the reactor trip breaker P-4 contact is properly c,losed, the blocking diode on printed circuit card
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A519* is failed open or interconnecting wiring is open. The
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diode and wiring will be confirmed in the following steps.
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- 6. With the multimeter still connected as in steps (2) and (3),
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close the reactor trip breaker.
7 The multimeter should read 48 VDC (ncminal).
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- 8. This indicates the reactor trip breaker P-4 contact is properly
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open, and confirms the blocking diode on printed circuit card
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A519* as well as the interconnecting wiring. End of test.
91 Should step (7) not yield a 48 VDC (nominal) reading, either the P-4 contact is not open, the blocking diode on printed cir,cuit card A519* is open, or interconnecting wiring is open.
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terminal leading to the SSPS, TB508-7.
- 12. P' lace the (-) lead on the terminal leading to the SSPS, TB508-8.
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CAUTION / ,
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DO NOT CLOSE BOTH BYPASS BREAKERS A & B SIMULTANEOUSLY.
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DOING SO WILL RESULT IN ALL BREAKERS INSTANTLY TRIPPIN"
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- 15. With the multimeter still connected as in steps (11) and (12),
close the bypass breaker.
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- 16. The multimeter should read 48 VDC (nominal).
17.~ This indicates the bypass breaker P-4 contact is properly open, and confirms blocking diode on printed circuit card .
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A519* and the interconnecting wiring. End of test.
- 18. Should step (16) not yield a 48 VDC (nominal) reading, either the P-4 contact is not open, the blocking diode on printed
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circuit card A519* is open, or interconnecting wiring is open.
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- 19. Initiate corrective action.
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The appropriate procedures should reflect a requirement to perform the above tests following automatic reactor trip or any condition requiring opening of tae reactor trip breakers. Repeat the tests ;
- following reclosure of the reactor trip breaks and prior to rod l withdrawal.
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C. Farley Unit 1 D. C. Cook Units 1 and 2. Beaver Valley Unit 1 Tro,1an, Salem Unit 1 North Anna Unit 1
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Inmediately perform the following for each train of SSPS:
- 1. Place a Siinpson Model 260 multimeter in the 50 VDC range.
- 2. At the reactor trip switchgear, place the (+) lead on the terminal leading to the SSPS, TB506-4.
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- 3. Plac,e the (-) lead on the terminal leading to the SSPS, TB506-5.
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- 4. ' The multimeter should read 48 VDC (nominal).
S'. This indicates that P-4 contact (s) is (are) properly open, and
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confirms the blocking diode on printed circuit card A518* as
- well as the interconnecting wiring. End of test.
- 6. Should step ,(4) not yield a 48 VDC (nominal) reading, either P-4
, contact (s) is (are) not open, blocking diode on printed circuit t
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caN A518* is open or interconnecting wiring is open.
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7._ Initiate corrective action.
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Implement the test sequence in part 0 for future periodic. testing when the plant is shutdown. Revise appropriate procedures to require verification by test of the P-4 contact status following autcmatic reactor trip or any condition requiring opening of the reactor trip breakers. Repeat the test following reclosure of the reactor trip breakers and prior to rod withdrawal.
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D. All Other Non-Operating Plants With An SSPS Which Are Not Identified in Part B or C
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Incorporate- the following test sequence for each train of SSPS, when the plant is at shutdown- and the SSPS' in normal operation: ,
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- 1. Place a Simpson Model 260 multimeter in the 50 VDC range.
- 2. At the reactor trip switchgear, place the (+) lead on the terminal leading to the SSPS, TB506-4.'
3 Place the (-) lead on the terminal leading to the SSPS,'TB506-5.
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- 4. The multimeter should read 0 VDC (nominal). ,
5.~ This.indicatestheP-4 contact (s)is(are)properlyclosed, the blocking diode on printed circuit card A518* is failed open or interconnecting wiring is open. The diode and wiring will be confirmed in the following steps.
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- 6. With the multimeter still connected as in steps (2) and (3),
close the reactor trip breakers.~ .
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- 7. The multimeter should read 48 VDC (nominal).
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- 8. This.indicatestheP-4 contact (s)is(are)properlyopen,and confirms the blocking diode on printed circuit card A518* as well as the interconnecting wirir.g. End of test.
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$hould step (7) not yield a 48 VDC (nominal) reading, either
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the P-4 contact (s) is (are) not open, the blocking diode on
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. printed circuit card A518* is not open, or interconnectin'g wiring- is open.
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10'. Initiate corrective action. / .
Revise appropriate procedures to require verification, by the above tests, of the P-4 contact status following automatic reactor trip or any condition requiring opening of the ieactor trip' breakers.
Repeat the tests following reclosure of the reactor trip breakers and prior to iod withdrawal.
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