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MONTHYEARIR 05000413/19870311987-10-16016 October 1987 Insp Repts 50-413/87-31 & 50-414/87-31 on 870914-18. Violation Noted.Major Areas Inspected:Radiation Protection, Including Organization & Mgt Control,Training & Qualification & External & Internal Exposure Control Project stage: Request ML20236A3371987-10-16016 October 1987 Proposed Changes to Nuclear Svc Water (Rn) Sys Tech Specs to Show That Sys Contains Components Shared Between Units & to Allow Placing Sys in ESF Alignment When Number of Operable ESF Channels Less than Required Project stage: Request ML20236A2831987-10-16016 October 1987 Application for Amends to Licenses NPF-35 & NPF-52,changing Nuclear Svc Water (Rn) Sys Tech Specs.Encls Include Tech Spec Changes,Justification & Safety Analysis & Revised FSAR Pages.Fee Paid Project stage: Request ML20196D0971988-01-22022 January 1988 Forwards Request for Addl Info Re Proposed Changes to Tech Specs & FSAR Concerning Nuclear Svc Water Sys.Response Requested within 30 Days Project stage: RAI ML20149L8621988-02-18018 February 1988 Forwards Response to 880122 Request for Addl Info Re 871016 Proposed Changes to Tech Specs & Fsar.Approval of Tech Spec Changes & Proposed Change in Design Bases for Nuclear Svc Water Sys Requested Project stage: Request ML20154C2351988-05-12012 May 1988 Special Rept:On 880412,19 & 25,diesel Generator 1a Failed to Start.Caused by Faulty Pneumatic Logic Components.Blowdown Frequency for All Generator Starting Sys Air Compressors Increased from Once to Twice Per Shift Project stage: Other 1988-01-22
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217F8231999-10-13013 October 1999 Informs That on 990930,NRC Completed mid-cycle PPR of Catawba Nuclear Station.Based on Review,Nrc Did Not Identify Any New Areas That Warranted More than Core Insp Program Over Next Five Months.Historical Listing of Issues,Encl ML20217H0041999-10-13013 October 1999 Forwards MOR for Sept 1999 & Revised MOR for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 ML20217F1301999-10-0707 October 1999 Forwards Rev 1 to Request for Relief 99-03 from Requirements of ASME B&PV Code,In Order to Seek Relief from Performing Individual Valve Testing for Certain Valves in DG Starting (Vg) Sys ML20212J3011999-10-0101 October 1999 Forwards Exemption from Certain Requirements of 10CFR54.17(c) Re Schedule for Submitting Application for Operating License Renewal.Se Also Encl ML20217K2651999-10-0101 October 1999 Forwards Retake Exams Repts 50-413/99-302 & 50-414/99-302 on 990921-23.Two of Three ROs & One SRO Who Received Administrative Section of Exam Passed Retake Exam, Representing 75 Percent Pass Rate 05000414/LER-1999-004, Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments1999-09-27027 September 1999 Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments 05000413/LER-1999-015, Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept1999-09-27027 September 1999 Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept ML20217A7911999-09-24024 September 1999 Forwards Insp Repts 50-413/99-05 & 50-414/99-05 on 990718- 0828 at Catawba Facility.Nine NCVs Identified Involving Inadequate Corrective Actions Associated with Degraded Svc Water Supply Piping to Auxiliary Feedwater Sys ML20212E6471999-09-24024 September 1999 Discusses GL 98-01 Issued by NRC on 980511 & DPC Responses for Catawba NPP & 990615.Informs That NRC Reviewed Response for Catawba & Concluded That All Requested Info Provided.Considers GL 98-01 to Be Closed for Catawba ML20212F0941999-09-21021 September 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals for Cns,Units 1 & 2 ML20212M2001999-09-20020 September 1999 Confirms 990913 Telcon Between M Purser & R Carroll Re Management Meeting to Be Conducted on 991026 in Atlanta,Ga to Discuss Operator Licensing Issues 05000414/LER-1999-005, Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments1999-09-20020 September 1999 Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments ML20212D5321999-09-15015 September 1999 Informs That Duke Energy Corp Agrees to Restrict Max Fuel Rod Average Burnup to 60,000 Mwd/Mtu,In Order to Support NRC Final Approval & Issuance of Requested Amend ML20212B4641999-09-14014 September 1999 Forwards Monthly Operating Repts for Aug 1999 & Revised Monthly Operating Rept for Catawba Nuclear Station,Units 1 & 2 ML20212A4131999-09-14014 September 1999 Informs That TR DPC-NE-2009P Submitted in 990817 Affidavit, Marked Proprietary,Will Be Withheld from Public Disclosure, Pursuant to 10CFR2.709(b) & Section 103(b) of Atomic Energy Act of 1954,as Amended ML20212M1931999-09-13013 September 1999 Refers to 990909 Meeting Conducted at Region II Office Re Presentation of Licensee self-assessment of Catawba Nuclear Station Performance.List of Attendees & Licensee Presentation Handout Encl ML20212A3751999-09-10010 September 1999 Informs That Postponing Implementation of New Conditions Improved by RG 1.147,rev 12,acceptable Since Evaluation on Relief Based on Implementation Code Case for Duration of Insp Interval ML20212A5191999-09-0808 September 1999 Requests NRC Approval for Relief from Requirements of ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,App VI,VI-2430(c) & 2440(b).Approval of 99-GO-002 Is Requested by 000301 05000413/LER-1999-014, Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment1999-09-0101 September 1999 Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment 05000414/LER-1999-003, Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev1999-08-31031 August 1999 Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev 0 of LER ML20211H1741999-08-30030 August 1999 Forwards Comments on Catawba Nuclear Station Units 1 & 2 & McGuire Nuclear Station,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid.Ltr Dtd 990107,rept ATI-98-012-T005 & Partial marked-up Rept WCAP-14995 Encl ML20211M4451999-08-30030 August 1999 Forwards Summary of Util Conclusions Re Outstanding Compliance Issue Re Staff Interpretation of TS SR 3.0.1,per Insp Repts 50-369/99-03 & 50-370/99-03,as Discussed with NRC During 990618 Meeting 05000413/LER-1999-013, Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER1999-08-25025 August 1999 Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER ML20211M8191999-08-25025 August 1999 Confirms 990825 Telcon Between G Gilbert & R Carroll Re Mgt Meeting to Be Held on 990909 in Atlanta,Ga,To Allow Licensee to Present self-assessment of Catawba Nuclear Station Performance ML20211A9641999-08-20020 August 1999 Forwards SE Authorizing Licensee 990118 Request for Approval of Proposed Relief from Volumetric Exam Requirements of ASME B&PV Code,Section XI for Plant,Units 2 ML20211C1191999-08-18018 August 1999 Forwards ISI Rept Unit 1 Catawba 1999 RFO 11, Providing Results of ISI Effort Associated with End of Cycle 11 ML20211B9471999-08-18018 August 1999 Forwards Request for Relief 99-02,associated with Limited Exam Results for Welds Which Were Inspected During Unit 1 End of Cycle 11 RFO ML20211C3651999-08-17017 August 1999 Forwards Rev 25 to Catawba Nuclear Station Units 1 & 2 Pump & Valve Inservice Testing Program, Which Includes Reformatting of Manual & Addl Changes as Noted in Attached Summary of Changes ML20211F2971999-08-17017 August 1999 Forwards non-proprietary & Proprietary Updated Pages for DPC-NE-2009,submitted 980722.Pages Modify Fuel Design & thermal-hydraulic Analysis Sections of DPC-NE-2009. Proprietary Page 2-4 Withheld,Per 10CFR2.790 05000413/LER-1999-011, Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment1999-08-16016 August 1999 Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment ML20211B1121999-08-16016 August 1999 Forwards Topical Rept DPC-NE-2012, Dynamic Rod Worth Measurement Using Casmo/Simulate, Describing Results of Six Drwm Benchmark Cycles at Catawba & McGuire & Discusses Qualification to Use Drwm at Catawba & McGuire ML20210V0321999-08-13013 August 1999 Forwards Insp Repts 50-413/99-04 & 50-414/99-04 on 990606- 0717.Six Violations of NRC Requirements Identified & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy ML20210S2751999-08-12012 August 1999 Forwards Monthly Operating Repts for July 1999 for Catawba Nuclear Station,Units 1 & 2.Revised Rept for June 1999,encl ML20210Q3751999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Authorized Representative of Facility Must Submit Ltr as Listed,Thirty Days Before Exam Date,In Order to Register Individuals for Exam ML20210N9521999-08-0404 August 1999 Forwards Changes to Catawba Nuclear Station Selected Licensee Commitments Manual.Documents Constitutes Chapter 16 of Ufsar.With List of Effective Pages IR 05000413/19980131999-08-0202 August 1999 Discusses Integrated Insp Repts 50-413/98-13,50-414/98-13, 50-413/98-16,50-414/98-16 & NRC Special Repts 50/413/99-11 & 50-414/99-11 Conducted Between Aug 1998 & May 1999.Six Violations Occurred,Based on OI Investigation & Insp ML20210M6411999-07-29029 July 1999 Forwards Request for Relief 99-03 from Requirements of ASME Boiler & Pressure Vessel Code,In Order to Seek Relief from Performing Individual Valve Testing for Certain Valves in DG Starting Air (Vg) Sys 05000413/LER-1999-010, Forwards LER 99-010-01 Which Replaces LER 99-002.Rept Number Has Been Changed in Order to Conform to Numbering Convention Specified in NUREG-1022,since Primary Event Involved Both Units1999-07-22022 July 1999 Forwards LER 99-010-01 Which Replaces LER 99-002.Rept Number Has Been Changed in Order to Conform to Numbering Convention Specified in NUREG-1022,since Primary Event Involved Both Units IR 05000413/19990101999-07-22022 July 1999 Discusses Insp Rept 50-413/99-10 & 50-414/99-10 on 990314- 0424 & Forwards Notice of Violation Re Failure to Comply with TS 3.7.13,when Misalignment of Two Electrical Breakers Rendered SSS Inoperable from 981216-29 ML20217G5241999-07-20020 July 1999 Forwards Exam Repts 50-413/99-301 & 50-414/99-301 on 990524- 27,0603,07-10 & 16.Of Fourteen SRO & RO Applicants Who Received Written Exams & Operating Tests,Eight Applicants Passed & Six Failed Exam 05000413/LER-1999-009, Forwards LER 99-009-00,re Inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits.Root Cause & Corrective Actions Associated with Event Are Being Finalized & Will Be Provided in Supplement to Rept1999-07-19019 July 1999 Forwards LER 99-009-00,re Inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits.Root Cause & Corrective Actions Associated with Event Are Being Finalized & Will Be Provided in Supplement to Rept 05000414/LER-1999-001, Forwards LER 99-001-01 Re Unanalyzed Condition Associated with Relay Failure in Auxiliary Feedwater Sys,Due to Inadequate Single Failure Analysis.Rev Is Being Submitted to Include Results of Failure Analysis Which Was Performed1999-07-15015 July 1999 Forwards LER 99-001-01 Re Unanalyzed Condition Associated with Relay Failure in Auxiliary Feedwater Sys,Due to Inadequate Single Failure Analysis.Rev Is Being Submitted to Include Results of Failure Analysis Which Was Performed ML20209H4431999-07-14014 July 1999 Forwards Monthly Operating Repts for June 1999 for Catawba Nuclear Station,Units 1 & 2.Revised Rept for May 1999 on Unit Shutdowns Also Encl ML20210A5771999-07-14014 July 1999 Forwards Revsied Catawba Nuclear Station Selected Licensee Commitments Manual, Per 10CFR50.71(e),changing Sections 16.7-5,16.8-5,16.9-1,16.9-3,16.9-5 & 16.11-7.Manual Constitute Chapter 16 of UFSAR ML20216D3941999-07-14014 July 1999 Forwards Revs to Catawba Nuclear Station Selected Licensee Commitments Manual NUREG-1431, Forwards SER Agreeing with Util General Interpretation of TS LCO 3.0.6,but Finds No Technical Basis or Guidance That Snubbers Could Be Treated as Exception to General Interpretation1999-07-0909 July 1999 Forwards SER Agreeing with Util General Interpretation of TS LCO 3.0.6,but Finds No Technical Basis or Guidance That Snubbers Could Be Treated as Exception to General Interpretation ML20196L0371999-07-0808 July 1999 Approves Requested Schedule Change of Current two-year Requalification Examinations to non-outage dates.Two-year Cycle Will Start on 991001 & Will End on 020930 05000413/LER-1999-008, Forwards LER 99-008-00,re Operation Prohibited by TS 3.5.2. Rev to LER Will Be Submitted by 990812 Which Will Include All Required Info About Ventilation Sys Pressure Boundry Breach1999-07-0808 July 1999 Forwards LER 99-008-00,re Operation Prohibited by TS 3.5.2. Rev to LER Will Be Submitted by 990812 Which Will Include All Required Info About Ventilation Sys Pressure Boundry Breach ML20196J9001999-07-0606 July 1999 Informs That 990520 Submittal of Rept DPC-NE-3004-PA,Rev 1, Mass & Energy Release & Containment Response Methodology, Marked Proprietary Will Be Withheld Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954 IR 05000413/19990031999-07-0101 July 1999 Discusses Insp Repts 50-413/99-03 & 50-414/99-03 Completed on 990605 & Transmitted by Ltr .Results of Delibrations for Violation Re Discovery of Potentially More Limiting Single Failure Affecting SGTS Analysis Provided 1999-09-08
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217H0041999-10-13013 October 1999 Forwards MOR for Sept 1999 & Revised MOR for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 ML20217F1301999-10-0707 October 1999 Forwards Rev 1 to Request for Relief 99-03 from Requirements of ASME B&PV Code,In Order to Seek Relief from Performing Individual Valve Testing for Certain Valves in DG Starting (Vg) Sys 05000414/LER-1999-004, Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments1999-09-27027 September 1999 Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments 05000413/LER-1999-015, Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept1999-09-27027 September 1999 Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept 05000414/LER-1999-005, Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments1999-09-20020 September 1999 Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments ML20212D5321999-09-15015 September 1999 Informs That Duke Energy Corp Agrees to Restrict Max Fuel Rod Average Burnup to 60,000 Mwd/Mtu,In Order to Support NRC Final Approval & Issuance of Requested Amend ML20212B4641999-09-14014 September 1999 Forwards Monthly Operating Repts for Aug 1999 & Revised Monthly Operating Rept for Catawba Nuclear Station,Units 1 & 2 ML20212A5191999-09-0808 September 1999 Requests NRC Approval for Relief from Requirements of ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,App VI,VI-2430(c) & 2440(b).Approval of 99-GO-002 Is Requested by 000301 05000413/LER-1999-014, Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment1999-09-0101 September 1999 Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment 05000414/LER-1999-003, Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev1999-08-31031 August 1999 Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev 0 of LER ML20211H1741999-08-30030 August 1999 Forwards Comments on Catawba Nuclear Station Units 1 & 2 & McGuire Nuclear Station,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid.Ltr Dtd 990107,rept ATI-98-012-T005 & Partial marked-up Rept WCAP-14995 Encl ML20211M4451999-08-30030 August 1999 Forwards Summary of Util Conclusions Re Outstanding Compliance Issue Re Staff Interpretation of TS SR 3.0.1,per Insp Repts 50-369/99-03 & 50-370/99-03,as Discussed with NRC During 990618 Meeting 05000413/LER-1999-013, Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER1999-08-25025 August 1999 Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER ML20211B9471999-08-18018 August 1999 Forwards Request for Relief 99-02,associated with Limited Exam Results for Welds Which Were Inspected During Unit 1 End of Cycle 11 RFO ML20211C1191999-08-18018 August 1999 Forwards ISI Rept Unit 1 Catawba 1999 RFO 11, Providing Results of ISI Effort Associated with End of Cycle 11 ML20211C3651999-08-17017 August 1999 Forwards Rev 25 to Catawba Nuclear Station Units 1 & 2 Pump & Valve Inservice Testing Program, Which Includes Reformatting of Manual & Addl Changes as Noted in Attached Summary of Changes ML20211F2971999-08-17017 August 1999 Forwards non-proprietary & Proprietary Updated Pages for DPC-NE-2009,submitted 980722.Pages Modify Fuel Design & thermal-hydraulic Analysis Sections of DPC-NE-2009. Proprietary Page 2-4 Withheld,Per 10CFR2.790 05000413/LER-1999-011, Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment1999-08-16016 August 1999 Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment ML20211B1121999-08-16016 August 1999 Forwards Topical Rept DPC-NE-2012, Dynamic Rod Worth Measurement Using Casmo/Simulate, Describing Results of Six Drwm Benchmark Cycles at Catawba & McGuire & Discusses Qualification to Use Drwm at Catawba & McGuire ML20210S2751999-08-12012 August 1999 Forwards Monthly Operating Repts for July 1999 for Catawba Nuclear Station,Units 1 & 2.Revised Rept for June 1999,encl ML20210N9521999-08-0404 August 1999 Forwards Changes to Catawba Nuclear Station Selected Licensee Commitments Manual.Documents Constitutes Chapter 16 of Ufsar.With List of Effective Pages ML20210M6411999-07-29029 July 1999 Forwards Request for Relief 99-03 from Requirements of ASME Boiler & Pressure Vessel Code,In Order to Seek Relief from Performing Individual Valve Testing for Certain Valves in DG Starting Air (Vg) Sys 05000413/LER-1999-010, Forwards LER 99-010-01 Which Replaces LER 99-002.Rept Number Has Been Changed in Order to Conform to Numbering Convention Specified in NUREG-1022,since Primary Event Involved Both Units1999-07-22022 July 1999 Forwards LER 99-010-01 Which Replaces LER 99-002.Rept Number Has Been Changed in Order to Conform to Numbering Convention Specified in NUREG-1022,since Primary Event Involved Both Units 05000413/LER-1999-009, Forwards LER 99-009-00,re Inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits.Root Cause & Corrective Actions Associated with Event Are Being Finalized & Will Be Provided in Supplement to Rept1999-07-19019 July 1999 Forwards LER 99-009-00,re Inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits.Root Cause & Corrective Actions Associated with Event Are Being Finalized & Will Be Provided in Supplement to Rept 05000414/LER-1999-001, Forwards LER 99-001-01 Re Unanalyzed Condition Associated with Relay Failure in Auxiliary Feedwater Sys,Due to Inadequate Single Failure Analysis.Rev Is Being Submitted to Include Results of Failure Analysis Which Was Performed1999-07-15015 July 1999 Forwards LER 99-001-01 Re Unanalyzed Condition Associated with Relay Failure in Auxiliary Feedwater Sys,Due to Inadequate Single Failure Analysis.Rev Is Being Submitted to Include Results of Failure Analysis Which Was Performed ML20216D3941999-07-14014 July 1999 Forwards Revs to Catawba Nuclear Station Selected Licensee Commitments Manual ML20209H4431999-07-14014 July 1999 Forwards Monthly Operating Repts for June 1999 for Catawba Nuclear Station,Units 1 & 2.Revised Rept for May 1999 on Unit Shutdowns Also Encl ML20210A5771999-07-14014 July 1999 Forwards Revsied Catawba Nuclear Station Selected Licensee Commitments Manual, Per 10CFR50.71(e),changing Sections 16.7-5,16.8-5,16.9-1,16.9-3,16.9-5 & 16.11-7.Manual Constitute Chapter 16 of UFSAR 05000413/LER-1999-008, Forwards LER 99-008-00,re Operation Prohibited by TS 3.5.2. Rev to LER Will Be Submitted by 990812 Which Will Include All Required Info About Ventilation Sys Pressure Boundry Breach1999-07-0808 July 1999 Forwards LER 99-008-00,re Operation Prohibited by TS 3.5.2. Rev to LER Will Be Submitted by 990812 Which Will Include All Required Info About Ventilation Sys Pressure Boundry Breach ML20196G7461999-06-22022 June 1999 Requests Exemption from Requirements of 10CFR54.17(c) That Application for Renewed Operating License Not Be Submitted to NRC Earlier than 20 Yrs Before Expiration of Operating License Currently in Effect ML20196E9541999-06-18018 June 1999 Forwards SG Tube Insp Conducted During Unit 1 End of Cycle 11 Refueling Outage.Attachments 1,2,3 & 4 Identify Tubes with Imperfections in SGs A,B,C & D,Respectively ML20195K4571999-06-14014 June 1999 Forwards MORs for May 1999 & Revised MORs for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20195J1691999-06-10010 June 1999 Forwards Written Documentation of Background & Technical Info Supporting Catawba Unit 1,notice of Enforcement Discretion Request Re TS 3.5.2 (ECCS-Operating),TS 3.7.12 (Auxiliary Bldg Filtered Ventilation Exhaust Sys) ML20217G5771999-06-0909 June 1999 Forwards Post Exam Comments & Supporting Reference Matls for Written Exams Administered at Catawba Nuclear Station on 990603 05000414/LER-1999-002, Forwards Abstract of LER 99-002-00 Re Forced Shutdown of Plant as Result of Flow Restriction Caused by Corrosion of Afs Assured Suction Source Piping Due to Inadequate Testing. Final LER Will Be Submitted No Later than 9907081999-06-0303 June 1999 Forwards Abstract of LER 99-002-00 Re Forced Shutdown of Plant as Result of Flow Restriction Caused by Corrosion of Afs Assured Suction Source Piping Due to Inadequate Testing. Final LER Will Be Submitted No Later than 990708 ML20207F2381999-06-0101 June 1999 Forwards Copy of Catawba Nuclear Station Units 1 & 2 1998 10CFR50.59 Rept, for NRC Files ML20195J1131999-05-26026 May 1999 Requests Approval to Change Cycle Dates for Two Year Requalification Training Program Required by 10CFR55.59,to Improve Scheduling of Requalification Exams to non-outage Periods 05000413/LER-1999-007, Forwards LER 99-007-00,re Operation Prohibited by TS 3.4.7. Commitments Identified in LER Are Listed in Planned Corrective Actions Section1999-05-26026 May 1999 Forwards LER 99-007-00,re Operation Prohibited by TS 3.4.7. Commitments Identified in LER Are Listed in Planned Corrective Actions Section ML20195B4751999-05-24024 May 1999 Forwards Rev 7 to UFSAR Chapter 2 & Chapter 3 from 1998 UFSAR for Catawba Nuclear Station.List of Instructions on Insertion Encl ML20196L1851999-05-20020 May 1999 Forwards Proprietary & non-proprietary Version of Rev 1 to TR DPC-NE-3004, Mass & Energy Release & Containment Response Methodology, Consisting of Finer Nodalization of Ice Condenser Region.Proprietary Info Withheld ML20196L1791999-05-20020 May 1999 Communicates Util Licensing Position Re Inoperable Snubbers. Licensee Has Determined That Structure of ITS Has Resulted in Certain Confusion Re Treatment of Inoperable Snubbers 05000413/LER-1997-009, Forwards LER 97-009-02, Unanalyzed Postulated Single Failure Affecting SG Tube Rupture Analysis, Suppl Revises Planned C/A Described in Suppl 1 to Ler.Current Status of C/As & Addl C/As Planned,Provided in Rept1999-05-17017 May 1999 Forwards LER 97-009-02, Unanalyzed Postulated Single Failure Affecting SG Tube Rupture Analysis, Suppl Revises Planned C/A Described in Suppl 1 to Ler.Current Status of C/As & Addl C/As Planned,Provided in Rept ML20206T4481999-05-13013 May 1999 Forwards Rev 3 to Topical Rept DPC-NE-3002-A, UFSAR Chapter 15 Sys Transient Analysis Methodology, IAW Guidance Contained in NUREG-0390 ML20206R1721999-05-13013 May 1999 Forwards Monthly Repts for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 & Revised Monthly Operating Repts for Mar 1999 ML20206T0281999-05-12012 May 1999 Forwards Changes to Catawba Nuclear Station Selected Licensee Commitments Manual. Document Constitutes Chapter 16 of UFSAR 05000413/LER-1999-006, Forwards LER 99-006-00,re CR Ventilation Sys Inoperability. Root Cause & Corrective Actions for Occurence Are Being Finalized & Will Be Reported in Supplement Rept on 9906071999-05-10010 May 1999 Forwards LER 99-006-00,re CR Ventilation Sys Inoperability. Root Cause & Corrective Actions for Occurence Are Being Finalized & Will Be Reported in Supplement Rept on 990607 ML20206N8201999-05-10010 May 1999 Forwards Revs 15 & 16 to Catawba Unit 1 Cycle 12 COLR, Per TS 5.6.5.Rev 15 Updates Limits for New Catawba 1 Cycle 12 Reload Core & Rev 16 Revises Values Re Min Boron Concentrations for Rwst,Cla & SFP ML20206J4431999-05-0303 May 1999 Forwards Changes to Catawba Nuclear Station Selected Licensee Commitments Manual, Per 10CFR50.71(e).Document Constitutes Chapter 16 of UFSAR ML20206D2141999-04-29029 April 1999 Forwards 1998 Annual Radioactive Effluent Release Rept for Catawba Nuclear Station,Units 1 & 2, Per Plant TS 5.6.3. Rept Contains Listed Documents ML20206E4101999-04-26026 April 1999 Forwards Four Copies of Rev 9 Todpc Nuclear Security & Contingency Plan,Per 10CFR50.54(p)(2).Changes Do Not Decrease Safeguards Effectiveness of Plan.Encl Withheld,Per 10CFR73.21 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L5491990-09-14014 September 1990 Forwards Proprietary Response to Question Re Scope of Review of Topical Rept, Safety Analysis Physics Parameter & Multidimensional Reactor Transients Methodology, Per & 900723 Meeting.Response Withheld ML20059L5521990-09-14014 September 1990 Forwards Response to 18 Questions Re Topical Rept DPC-NE-2004,per NRC 900802 Request for Addl Info.Encl Withheld (Ref 10CFR2.790) ML20059K2021990-09-12012 September 1990 Submits Supplemental Response to Generic Ltr 89-14, Svc Water Sys Problems Affecting Safety-Related Equipment. Intake Structure Insp Program Developed.Procedures for Insp Implemented & Intake Structures Sampled & Analyzed ML20064A8041990-09-0505 September 1990 Notifies NRC of Mod to 890301 Response to Violations Noted in Insp Repts 50-413/86-18-01 & 50-414/86-18-01 Re Valves. All Valve Locking Mechanisms Would Be Installed by End of Unit 2 Refueling Outage (Approx Aug 1990) ML20064A5741990-09-0404 September 1990 Discusses Re Info to Support Util Position Relative to Resolving Issue of Main Steam Line Breaks Inside Ice Condenser Containments & Requests That Info Be Withheld (Ref 10CFR2.790) ML20059G3011990-09-0404 September 1990 Forwards Response to NRC 900327 Request for Addl Info Re BAW-10174, Mark-BW Reload LOCA Analysis for Catawba & Mcguire ML20059G8321990-08-30030 August 1990 Withdraws 880726 Proposed Tech Spec Change,Clarifying Tech Spec 3/4.7.6 Re Emergency Power Requirements for Control Room Ventilation Sys ML20059D2011990-08-27027 August 1990 Forwards Piedmont Municipal Power Agency , Authorizing Use of Annual Rept for NRC Docket Requirements ML20059D2441990-08-24024 August 1990 Forwards Special Rept PIR-1-C90-0261 on 900725 Re Cathodic Protection Sys Failure to Pass Acceptance Criteria of 60-day Surveillance.Std Work Request Generated to Check Voltage Potential at Test Station TS-36 on Weekly Basis ML20056B4981990-08-22022 August 1990 Responds to NRC Request for Addl Info Re General Relief Request for Pump Vibration Submitted 900315.Relief Request Changed to Insure Data Taken Over Range That Encompasses All Main Potential Noise Contributors ML20056B5011990-08-22022 August 1990 Responds to Violation Noted in Insp Repts 50-413/90-17 & 50-414/90-17.Corrective Actions:Review Will Be Conducted to Determine Category of Infrequently Run Procedures Needing Addl Verification Controls ML16259A2391990-08-22022 August 1990 Forwards Public Version of Rev 27 to Company Crisis Mgt Implementing Procedure CMIP-2, News Group Plan. W/ Dh Grimsley 900906 Release Memo ML20056B4971990-08-20020 August 1990 Clarifies Info Submitted in 871207 & s Re Steam Generator Tube Rupture Analysis Demonstration Runs. Demonstration Runs Met plant-specific Requirements in Section D to NRC SER on WCAP-10698 ML20059C1201990-08-20020 August 1990 Forwards Rept Summarizing Util Findings Re Three False Negative Blind Performance Urine Drug Screens Which Occurred During Jan & Feb 1990.Recommends That NRC Consider Generic Communication to Clearly State Reporting Requirement ML20059B6581990-08-17017 August 1990 Responds to Violation Noted in Insp Repts 50-413/90-15 & 50-414/90-15.Corrective Actions:Present Methods of Testing Operability of CO2 Fire Protection Sys Will Be Evaluated by 910201 to Determine If Addl Testing Necessary ML20059C1591990-08-17017 August 1990 Suppls by Providing Addl Info to Support Util Position Re Anl Confirmatory Analysis of Main Steamline Breaks in Ice Condenser Plants.Encl Withheld ML20063Q0951990-08-15015 August 1990 Forwards Monthly Operating Rept for Jul 1990 for Catawba Nuclear Station Units 1 & 2 & Revised Rept for June 1990 ML20059C1231990-08-15015 August 1990 Advises That Util Submitting Special Rept Re Valid Failure of Diesel Generator 2B Would Be Delayed Until 880229 Had Incorrect Ltr Date.Date of Ltr Should Have Been 880204 Instead of 880104.Corrected Ltr Encl ML20063Q2671990-08-14014 August 1990 Forwards Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 36 to CMIP-1,Rev 32 to CMIP-4,Rev 36 to CMIP-5,Rev 41 to CMIP-6,Rev 40 to CMIP-7,Rev 27 to CMIP-8 & Rev 35 to CMIP-9.W/DH Grimsley 900821 Release Memo ML20059C2211990-08-13013 August 1990 Forwards Revised Chapter 16, Selected Licensee Commitments Manual, to Plant Updated Fsar,Per 10CFR50.4 & 50.71.Manual Contains Commitments Which Require Control But Not Appropriate in Tech Specs ML20063Q0261990-08-10010 August 1990 Forwards Rev 0 to Catawba Unit 2 Cycle 4 Core Operating Limits Rept, Per Tech Spec 6.9.1.9 ML20063Q0671990-08-10010 August 1990 Submits Revised Response to Violations Noted in Insp Rept 50-413/90-09.Procedure to Verify Test Inputs Modified to Verify Dummy Input Signal to Channel RTD Circuit ML20058N0181990-08-0808 August 1990 Forwards Response to Request for Addl Info Re BAW-10174, Mark-BW Reload LOCA Analysis for Catawba & Mcguire ML20081E1601990-08-0101 August 1990 Advises of Completion of 900330 Commitment Re Standing Work Request for Insp of Air Flow Monitors & Dampers,Per Violations Noted in Insp Rept 50-413/90-03 & 50-414/90-03 ML20058P3261990-08-0101 August 1990 Forwards Public Version of Rev 26 to Station Directive 3.8.4, Onsite Emergency Organization ML20081E0951990-07-27027 July 1990 Forwards Decommissioning Financial Assurance Certification Rept for Duke Power Co,co-owner of Catawba Nuclear Station Units 1 & 2 ML20055H9741990-07-26026 July 1990 Forwards end-of-cycle 3 Steam Generator Insp Rept.Nineteen Tubes Removed from Svc by Plugging W/Rolled Mechanical Plug ML20055H5231990-07-24024 July 1990 Discusses co-licensee Relationship & Obligations Re Decommissioning Financial Assurance for Facilities ML20055H4571990-07-19019 July 1990 Responds to NRC Re Violations Noted in Insp Repts 50-413/90-11 & 50-414/90-11.Corrective actions:I-beams/ Hoists Rolled to Ends of Ice Condenser & Securely Located on Rails to Prevent Any Movement ML20055H1741990-07-18018 July 1990 Withdraws 880527 & 0725 Amends Clarifying Requirements for Containment Pressure Control Sys ML20055J3441990-07-17017 July 1990 Advises That Commitment Re Procedure IP/O/A/3190/01,per Violation in Insp Repts 50-413/90-06 & 50-414/90-06, Completed on 900619 ML20055H4131990-07-16016 July 1990 Forwards Public Version of Epips,Including RP/0/A/5000/07 & HP/0/B/1009/04 ML20055F8991990-07-13013 July 1990 Forwards Monthly Repts for June 1990 for Catawba Nuclear Station Units 1 & 2 & Operating Status Rept for May 1990 ML20055G2311990-07-13013 July 1990 Withdraws 880311 Proposed Amend to Tech Spec Table 3.3-3, Item 8.f Re Number of Instrumentation Channels Associated W/ Main Feedwater Pumps.Util Determined That Change Unnecessary ML20055F8461990-07-12012 July 1990 Requests 14-day Extension Until 900802 to Submit LER 414/90-010 to Investigate Power Supply Realignment ML20058P1231990-07-0707 July 1990 Advises That Commitment to Revise Maint Mgt Procedure 1.12 to Include Functional Verification Requirements & to Develop Retest Manual to Address Retest Requirements for Any Maint Performed on Components Completed on 900614 ML20055F4131990-07-0505 July 1990 Forwards Inservice Insp Rept Unit 1 Catawba 1990 Refueling Outage 4, Per 10CFR50.55(a)(q) & Tech Spec 4.0.5.Insp Performed Per Section XI of ASME Boiler & Pressure Vessel Code & Applicable Addenda ML20055D4291990-06-29029 June 1990 Supplemental Response to Violations Noted in Insp Repts 50-413/89-13 & 50-414/89-13,per .Personnel Responsible for Maintaining Crisis Mgt Ctr Drawing Trained. Util Will Continue to Evaluate Changes Made to Program ML20055E2191990-06-29029 June 1990 Submits Revised Commitment Dates Re Implementation of Dept Guidance on post-maint Testing,Per Commitment Made in 891002 Response to Violations in Insp Repts 50-413/89-19 & 50-414/89-19.Completion Date Changed to 900701 ML20044B0621990-06-26026 June 1990 Forwards Public Version of Revised EPIP HP/0/B/1009/05, Personnel/Vehicle Monitoring for Emergency Conditions. W/Dh Grimsley 900716 Release Memo ML20043H6921990-06-18018 June 1990 Advises of Revised Completion Date for VA Ductwork Cleaning to 901231,per Insp Repts 50-413/90-03 & 50-414/90-03. Vendor Personnel Assigned to Task Unavailable to Complete Cleaning Until Late 1990 Due to Outage Support Needs ML20043G1691990-06-15015 June 1990 Forwards Monthly Operating Repts for May 1990 for Catawba Nuclear Station,Units 1 & 2 & Corrected Monthly Operating Repts for Apr 1990 Re Personnel Exposure ML20055C8041990-06-15015 June 1990 Responds to NRC Re Violations Noted in Insp Repts 50-413/90-10 & 50-414/90-10.Corrective Actions:Instrument Root Valves Unisolated & Analog Channel Operational Tests for Low Temp Overpressure Protection Completed ML20043G4331990-06-13013 June 1990 Withdraws 900423 Proposed Amend to Tech Spec 4.6.1.8 Re Lab Test of Carbon Samples from Annulus Ventilation Sys ML20043G3771990-06-13013 June 1990 Withdraws 900423 Proposed Amend to Tech Spec 4.7.7 Which Required That Lab Test of Carbon Samples from Auxiliary Bldg Filtered Exhaust Sys Be Tested for Methyl Iodide Penetration of 0.71% ML20043G2511990-06-12012 June 1990 Withdraws 900419 Suppl to 871221 Application for Amends to Licenses NPF-35 & NPF-32 Re Tech Specs 4.7.6 Re Control Room Area Ventilation Surveillance Requirements ML20043G1741990-06-0707 June 1990 Responds to Request for Addl Info Re BAW-10174, Mark-BW Reload LOCA Analysis for Catawba & Mcguire. Correct RCS Operating Pressure Would Be 2,250 Psia as Identified in Table 3-1 ML20043G3451990-06-0707 June 1990 Forwards Proprietary Response to Request for Addl Info Re Topical Rept BAW-10174, Mark-BW Reload Safety Analysis for Catawba & Mcguire. Response Withheld ML20043G0721990-06-0707 June 1990 Responds to NRC 900510 Ltr Re Violations Noted in Insp Repts 50-413/90-09 & 50-414/90-09.Corrective Actions:Vc/Yc Train a Returned to Svc W/Supply Power from 2ETA.Terminal Box 1TB0X0346 Inspected & Insured Operable ML20043F6111990-06-0606 June 1990 Advises That Response to Request for Addl Info Re Operator Response Times During Simulated Steam Generator Tube Rupture at Facility,Will Be Delayed Until 900630 1990-09-05
[Table view] |
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-s Duxu Powen Gonnm 15.0. IlOX 3318D Cll AHLOTTI!. N.C. 2112 4 2 IIAL 11. TUCKER trtmenown wwa revennent (704) OrMS34 WM4 RAS PacDETTlos February 18, 1988 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555
Subject:
Catawba Nuclear Station Doc':et Nos. 50-413 and 50-414
Dear Sir:
Dr. K. N. Jabbour's letter of January 22, 1988 transmitted a Request for Additional Information concerning proposed changes to the Technical Specifications and Final Safety Analysis Report (FSAR) description for Catawba's Nuclear Service Water (RN) System which were submitted on October 16, 1987.
Attached are responses to the questions along with marked-up FSAR pages related to the question responses.
As a part of the October 16, 1987 submittal, revised FSAR pages were submitted which reflected changes to the RN system including a proposed change in the design bases. This change would delete the previous assumption of a simultaneous LOCA and seismic event. As a result the Staff concluded that the proposed Technical Specification change would involve an increase in the probability or consequences of previously evaluated accidents.
First of all, the Staff is mixing the proposed change in the design bases of the RN system with the proposed Technical Specification change, which is a separate matter. The changes to the Technical Specification added clarifying statemelits to more accurately reflect the shared nature of Catawba's RN system. These changes would not change the way the RN system is currently operated.
Secondly, Duke does not agree that changing the design bases of the RN system to no longer consider LOCA and seismic as simultaneous events would increase the probability or consequences of any previously evaluated accident since this change will not result in any changes in the design or operation of the system.
Duke has already deleted the swapover from Lake Wylie to the standby nuclear service water pond on a LOCA (Sp) signal and plans to add additional pit level ins t rumentation. Both changes are discussed in the Staff's SER dated September 30, 1987. No other changes are contemplated as a result of separating seismic and LOCA events.
The change in the design bases was requested in order to revise an over-commitment in the FSAR. A simultaneous LOCA and seismic event was not considered credible by the NRC in the recent GDC-4 rulamaking and the combination of these independent events is not regulred by the Standard Review Plan or applicable General Design Criteria. 00g 88022*f009 080210 \
PDR Ale &M 05000413 P PDR
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- 0. S. Nuclear Rsgulatory Connission February 18, 1980 ,
Page Two Therefore, it is again requested that the NRC Staff approve the Technical Specification changes and the proposed change in the design bases for the Nuclear Service Water System. Since this submittal supplements Duke's lettor of October
- 16, 1987, no additional Part 170 fees are included.
Very truly yours, b Vbo n Hal B. Tucker ROS/1403/sbn Attachment xc: Dr. J. Nelson Grace, Regional Administration U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 i
Mr. P. K. Van Doorn NRC Resident Inspector ,
Catawba Nuclear Station I
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w Duke power Company Catawba Nuclear Station Response to NRC RAI - January 22, 1988 r
(1) In Technical Specification Tables 3.3-3, and 4.3-2, Item 14.g, the Applicable Mode is identified as Modes 1, 2, 3, 4. This should ha revised to identify that it applies when either unit is in Modes 1, 2, 3, 4 because even for single unit operation, both pump pits must be operable. ,
Responss:
The manner in which the proposed revision to Table 3.3-3 and 4.3-2 item 14.g is presented is consistent with the specifications associated with other shared systems (denoted "Unit 1 and 2"). Specification 3.0.5 requires that in this case "the ACTION requirements will apply to both units simultaneously". The specification appears proper as proposed.
(2) The proposed rovision to the Basos B3/4.7.4 identifies that one RN pump has sufficient capacity to maintain a unit indefinitaly in COLD SHUTDOWN (commencing 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following a trip from full power) while supplying the post-LOCA loads on the other unit. However, the proposed Specification 3/4.7.4 for the RN system does not consider the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> time period. For example, Specification 3.7.4 discusses "both units in MODE 1, 2, 3 or 4".
It should discuss either both units in MODE 1, 2, 3 or 4 or one unit in MODE 1, 2, 3 or 4 plus the other unit in MODE 5 or 6 for less than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Revise your propored specification to reflect this 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> period.
Response
Two Nuclear Service Water loops are required by specification 3.7.4 to assure the performance of the safety function in the event of a single i failure coincident with a LOCA (or other accident). As with many specifications, the ACTION allows continued operation without single failure :
l protection for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, with a subsequent requirement to be in COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The proposed revision to the Bases B3/4.7.4 is consistent with these ACTION requirements.
(3) The proposed surveillance requirements for the standby nuclear service water pond (SNSWp) under technical Specification 3/4.7.5 added a requirement to measure the RN temperature in the discharge path of an operating RN pump i during the months of July, August and September while the RN system is aligned to Lake Wylie. While this may be a necessary operation, it does not appear to affect the operability of the SNSWp as imolied by the location of the requirement. Revise the proposed specifications to identify why this measurement is necessary and place it in the proper location.
-Response:
l Ar described in the proposed revision to Bases B3/4.7.5, Nuclear Service l Water temperatures is an input to the containment pressure analysis. The l SNSWP is the assured source of RN and as such is appropriately controlled by I LCO 3/4.7.5. Operator actions following a LOCA require a knowledge of Lake Wylie temperature in order to determine if a manual realignment to the SNSWp i la needed. The proposed addition of specification 4.7.5.d assures that this i ten.perature is monitored and recorded daily. This assures that SNSWp will I
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desponse to NRC RAI page 2 be properly utilized in the event of an accident and thus is associated with SNSWP OPERABILITY. The safety analysis is unaffected by Lake Wylie temperature, therefore, a specification limiting temperature is not appropriate.
(4) In the proposed FSAR amendment the RN flows to the containment spray (CS) heat exchanger and component cooling water (CCW) heat exchanger have been decreased from 4500 gpm to 3800 gpm and 6500 gpm to 5200 gpm, respectively.
Consequently the design heat transfer capability has been correspondingly decreased. provide the following related information:
(a) Explain the reason for this decrease in RN flows and discuss why no other RN cooled components are affected, i.e., is there a corresponding flow reduction to other components cooled by the RN system?
(b) Why is there no corresponding change in the post - LOCA containment pressure / temperature profiles? If there is a change then the FSAR should be revised accordingly.
Response
Westinghouse has reanalyzed the accident heat loads which has resulted in a reduced required flow rate to heat exchangers cooled by nuclear service water (RN). Credit for these reduced flows has been taken in order to support one RN pump operation. This analysis reduced flow only to the major i essential header components (Component Cooling and Containment Spray Heat Exchangers), since the nonessential header is isolated in an accident. The I Westinghouse analysis is referenced in the one RN pump calculations which '
were submitted by H. B. Tucker's letter of January 4, 1988. '
The Westinghouse analysis (see Attachment 12 of January 4, 1988 submittal),
i using approved WCAP-10325, revised the mass and energy releases for the LOCA analysis. These changes would have resulted in lower post-LOCA !
pressure / temperature profiles. Instead, assumptions for heat exchanger flow l
and heat transfer were revised to take advantage of the increased margin. l The post-LOCA peak pressure and temperature were held essentially constant I and are bounded by the current Technical Specification and EQ envelope. The post-LOCA temperature / pressure profiles did change and will be included in an update to the FSAR.
i (5) Because the revised design results in a situation where following an accident, the RN system might very well continue to draw from Lake Wylie instead of the SNSWp, the maximum temperature of Lake Wylie should also fall within the Technical Specification. Revise your Technical specifications i accordingly (Refer to Question 3 above).
Response
See response to question 3.
(6) Do crocsover valves between SW trains still receive close signals on a safety injection signal (SIS) or containment isolation signal? It was the staff's understanding that only the switchover from Lake Wylie to the SNSWp l would be eliminated following an SIS. Specifically identify those valves
Neaponse to NRC RAI Page 3 whose operation on an SIS or containment isolation signal will be different following the proposed changes. The staff's concern is that all possible scenarios are considered, especially if the proposed changes involve more than the switchover between Lake Wylie and the SNSWP. In your response specifically identify whether you still have automatic isolation between trains and discuss when such isolation would occur.
Response
Crossover valves between service water trains still close on a containment isolation (Sp) signal on the affected unit. Important RN valves that change position or used to change position upon ESF signale are:
Group I -- Isolate RN from Lake Wylie IRNIA 1RN843B 1RN2B 1RN847A 1RN5A 2RN847A 1RN6B 1RN8498 1RN52A 2RNB49B Group II -- Isolate RN from the SNSWP 1RN3A 2RN846A l 1RN4B 1RN848B 1RN846A 2RN848B Group III -- RN Supply Header Crossover Isolation Valves IRN47A 1RN48B 2RN47A 2RN48B Group IV -- RN Return Header Crossover Isolation Valves 1RN53B 1RN54A 1
Valves f rom Group I used to close upon Ss or upon emergency low RN pump pit level signal. Now, they automatically close only on an emergency low RN pump pit level signal.
Valves in Group II used to open upon Ss or upon emergency low RN pump pit l I level signal. Now, they automatically open only on an emergency low RN pump pit level signal, )
j Valves in Group III used to have the following interlocks: ;
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(a) 1RN47A and 1RN48B close upon Sp from Unit 1.
I (b) 2RN47A and 2RN48B close upon Sp from Unit 2.
1 (c) 1RN47A close upor. Sp from Unit 2 with emergency low level signal from i RN pump pit B.
RhsponzotoNRCRAI Page 4 (d) 1RN48B close upon Sp from Unit 2 with emergency low level signal from RN pump pit A.
(e) 2RN47A close upon Sp from Unit 1 with emergency low level signal from RN pump pit B.
(f) 2RN48B close upon Sp from Unit I with emergency low level signal from RN pump pit A.
Now, valves in Group III have the following interlocks:
(a) 1RN47A and 1RN48B close upon Sp from Unit 1.
(b) 2RN47A and 2RN48B close upon Sp from Unit 2.
(c) 1RN47A and 2RN47A close upon emergency low level in pump pit B.
(d) 1RN48B and 2RN48B close upon emergency low level in pump pit A.
Valves in Group IV used to close upon Sp or upon emergency low RN pump pit level. Now, they automatically close only on an emergency low RN pump pit level signal.
The changes in Group III and IV valves were intended to correct a weakness detected in the previous logic and to minimize the consequences of postulated active valve failures.
Regarding Group III, under the old logic, if there was a loss of Lake Wylie end a single failure removed an RN loop from service, the RN system would not autoratically separate into independent loops unless there was an Sp signal on one of the units. Under the new logic, loss of Lake Wylie would be detected by the RN pump pit level instrumentation, and trains would separate upon emergency low level without regard to ESF signals. ESF logic is not degraded by this change, and the RN System loops will separate to insure continued operation in a non-ESF failure scenario.
The comments on Group III valves also apply to Group IV valves, but Fore comments are needed on Group IV valves. There are two main crossovers on the RN cupply header, one for Unit 1 and one for Unit 2. There is only one return header crossover, isolated by 1RN53B and 1RN54A. Each loop has an independent return path to the SNSWP, but the single RN returit path to Lake Wylie originates from RN loop A. Since RN will be aligned to Lake Wylio during all modes of operation, including ESF events in which a RN pump pit emergency low level signal is not given, the Sp interlocks had to ve removed from the return header crossover isolation valves to provide a discharge path from loop B to Lake Wylie.
In addition to the above, there is no change in the Sp interlocks to 1RN58B and 1RN63A, which isolate main RN discharge to the SNSWP. These vales open upon an Sp from either unit to assure a discharge path for the RN System in case 1RN53B, 1RN57A or 1RN843B transfer closed during an ESF event.
i Response to NRC RAI ,
Page 5 '
l The result of the forgoing changes is to make RN loop separation totally dependent upon RN pump pit level. The partial loop separation on Sp on the affected unit serves only to isolate nonessential flow rates during ESF ,
events.
(7) If automatic crosacvor isolation or isolation of nonessen+.ial loads does not occur until loss of bake Wylie, then you should reevaluate various scenarios other than LOCA during different modes of operation to ensure that the proposed Technical Specifications are acceptable. For example, evaluate a diversion of RN flow through a faulted nonossential portion of the system ,
under different possible accident scenarios. i
Response
The nonessential headers can be isolated by redundant isolation valves in series. The nonessential header isolation valves are in addition to the crossover isolation valves between loops. The branch connections for the l nonessential headers are located between the crossover isolation valves on each unit. If a diesel generator is declared inoperable, station procedures call for the RN crossover valve between the nonessential header and the loop !
with the inoperable D/G to be closed. This action would help the operator l recover if there was an accident and a subsequent single RN failure. '
Deletion of the Sp switchover on the RN header crossover valves does not j change the systems ability to respond to other than LOCA accidents.
(8) The safety injection pumps' and ccW pumps' heat exchanger inlet valves were previously identified as interlocked to open when their respective pump started. The proposed amendment deletes this from the FSAR. Explain why this change was made and justify deletion of this interlock.
Response
To minimize fouling and degradation problems in small diameter cooling water pipe, several small heat loads (pump motor coolers and oil coolers) formally cooled by RN nre now cooled by KC (CCW). The safety injection pump motor coolers and component cooling water pump motor coolers are among the pump auxiliaries now cooled by KC. When these pump motor coulers were cooled by RN, condensation on the tubes during cold weather was a concern. To minimize the condensation, the inlet isolation valve was fitted with an operator interlocked to open when the pump was started, close when the pump was stopped. When the pump motor cooler was moved to KC, condensation was not a big concern, and the inlet isolation valve was not fitted with an operator. Valves are left open so flow can be supplied to the safety injection and component cooling pump motor coolers during any mode.
The change by which small heat loads were moved to KC was done during the construction phase of Catawba. This FSAR change was made to reflect the transfer of these loads to KC.
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Respons9 to NRC RAI Page 6 F
(9) On page 9.2-1 of the FSAR, it was previously stated that "should Lake Wylie be lost due to a seismic event... the SNSWP contains sufficient water to bring the station safely to a cold shutdown condition following a single loss of coolant accident". Your proposed FSAR amendment deletes "following a single loss of coolant accident". Even though your proposed change is to delete a simultaneous LOCA and seismic event as a design basis, the SNSWP still must be capable of handling a LOCA upon loss of Lake Wylie. You should revise the aubject FSAR statement to state that the SNSWP contains sufficient water to bring the station safely to a cold shutdown condition under all normal, transient and accident conditions. The automatic switchover on low pump pit level should assure this function.
Response
The reason for deleting the reference of a LOCA from Section 9.2.1.2.1 was due to the separation of the seismic and LOCA events. It was not intended
- to portray the RN system as not being capable of handling a LOCA with a later (not simultaneous) loss of Lake Wylle. The SNSWP does contain ;
sufficient water to bring the station safely to a cold shutdown condition for all cases (normal, trancient and accident). The automatic switchover 3 function assures this should Lake Wylie be lost for any reason, not just LOCA. The FSAR statement will be revised as requested. (see Attachment)
(10) On page 8.2-9 of the FSAR, you deleted the statement that "the operation of any two pumps on either or both supply lines is sufficient to supply all cooling water requirements for the two unit plant for post-accident operation". Your revision does not include "post-accident operation". Does this deletion / revision mean that two pumps cannot handle all accident situations, or are you implying that one pump is sufficient under all l accident conditions? The reason for this change should be made clear, j
Also, if you are saying one pump is sufficient, then supporting analysis I
should be provided such that the staff can make its own independent evaluation.
Response
Reference should be to page 9.2-2 (sixth paragraph). Two pumps are l sufficient to supply all cooling water requirements for unit startup, j cooldown, refueling and post-accident operation of two units. However one pump has sufficient capacity to supply all cooling water requirements during normal power operation of both units or during post accident conditions if the unaffected unit is already in cold shutdown. One RN pump calculations were submitted by H. B. Tucker's letter of January 4, 1988. The FSAR will be revised to clarify this. (see Attachment)
(11) In a similar vuln, on revised FSAR page 9.2-5 it is stated that "bearing luce oil injection flow is maintained to all RN pumps at all times, even though only one pump is required to meet all the normal and accident flow requirements of both units". Previously, this was considered to be I applicable only under normal conditions. You should clarify what the design bases for the RN system is, one pump or two pumps. From the Technical
, Specification bases it appears that one pump in sufficient for accident i
situations only after one unit has been shutcovn for greater than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
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1 R'osponso to NRC RAI Page 7
Response
As discussed in FSAR Section 9.2.1 and the proponed Bases to Specification 3/4.7.4, two RN pumpa are needed to supply post .DCA loads on one Unit and ;
shutdown and cooldown loads on the other unit. However, with one unit in cold shutdown, only one RN pump and its associated emergency diesel generator are needed. The cited FSAR paragraph will be revised to eliminate the current confusion over design bases. (see Attachment) !
(12) On page 9.2-7 of your proposed FSAR revision, you have deleted the fact that ;
the RN system is designed to handle a LOCA in one unit with a simultaneous shutdown of the other unit plus the loss of Lake Wyl*.e. This is unacceptable. The staff requires that this remain a design basis for the RN system and the ultimate heat sink, the standby nuclear service water pond.
Although simultaneous LOCA and seismic loads do not have to be considered, reliance on Lake Wylie which is not designed to seismic Category I requirements is not acceptable under LOCA conditions i.e., General Design Criterion 2 and 10 CFR Part 100.
Response
During the extensive review of Catawba's RN System it was recognized that the original design basis, which assumed a simultaneous LOCA and seismic ;
evelt, was an unnecessary over commitment. This assumption led to the auttratic swapover from the normal cource of cooling water (Lake Wylie) to the standby nuclear service water pond (SNSWP) on a LOCA signal. It was recog71:ed that Lake Wylie was a highly reliable source of cooling water, in a numler of respects preferable to the SNSWP.
Therefore there was no practical reason to automatically swap from Lake Wylie to the SNSWP. This swap represented an unnecessary challenge to the ,
RN System. In order to improve reliability of the RN System, the automatic
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swapover on a LOCA (Sp) signal was deleted with NRC staff concurrence (S. 7.. I Varga } etter of September 30, 1987). One additional modification is planneo, the addition of RN pump sit level monitoring instrumentation from a 1 out of 2, to a 2 out of 3 system. No other hardware or operational chnnges are contemplated as a result of the change in the design basis. The SNSWP will still be the "ltimate heat sink for the station and the RN system !
will automatically realign to the SNSWP on a low level in Lake Wylie or by !
operator action. The net effect has been an improvement in the reliability of the RN System.
In reviewing the applicable regulatory documents, it was concluded that the FSAR commitment to a simultaneous LOCA and seismic event was unnecessary:
NUREG-0800 - Section 9.2.1 of the Standard Review Plan does not include simultaneous LOCA and seismic events as criteria for an acceptable station service water system.
NUREG-0954 - Section 9.2.1 of the Catawba SER makes no mention of and gives no credit for a capability of mitigating seismic and LOCA events simultaneously.
Responce to NR*] RAI page 8 i
I GDC-4 Ruleraking - As acknowledged by the NRC in this recent rulemaking proceedir.g seismic and LOCA loads do not have to be considered J concurrently, i.e., that LOCA and seismic are independent events.
Other FSAR's - A review of the FSAR's of other operating reactorn with similarly configured station service water systems indicates that '
] simultaneous LOCA and seismic events was not consistently assumed by other utilities. ,
ope-ation of the RN System in the current configuration, i.e., the SNSWp pros m ng an assured backup to Lake Wylie, is consistent with the operation of other systems found acceptable by the NRC. For example, the condensate-quality water supply for the Auxiliary Feedwater System is.
l non-safety and non-seismic. These sources are the preferred sources for any -
event requiring auxiliary feedwater initiation. The swapover to the assured safety grada source (RN System) is not perfomed on a LOCA signal (SP). The swapover is made only if the non-saft ty sources oecome unavailable. There is no presumpticn of a simultaneous LOCA and seismic event.
(13) Additionally, on the revised Safety Evaluation Section of the FSAR (page .
9.2-7) you state that upon compie*e channel separation, both units are essured of 6 t ng a source of water and at least one pump. This is not as i clear as in ' M v iginal FSAR where it is stated that each unit will have at l least one it. J 4. nt capacity pump. Revise this proposed change to I identify whethwr each unit is assured of having at least one pump or not.
If you intend to rely on a single pump for both units then the appropriate I analysis should be provided, i
j Response:
The Safety Evaluation correctly states that the normal configuration will provide each unit with at least one 100% capacity pump, one essential headar and an assured source of water. In the case of having a diesel generator out-of-service for an extended period of time, not a normal configuration, (and its associated unit in cold shutdown) the consequences of a simultaneous LOCA on the operating unit, loss of offsite power and single l failure anywhere on the system could result in having only one RN pump 1 operable to provide cooling water to the LOCA unit and to maintain the other I unit in cold shutdown. This has been demonstrated by the one RN pu::p l analysis which was submitted on January 4, 1988.
l l (14) In the original FSAR, Section 9.2.1.3, you stated that any one diesel
, generator can be down for maintenance and the RN system can still shut the plant down safely assuming a LOCA, seismic event, blackout, and single failure. In your proposed amendment you have eliminated the seismic event.
- Identify the bases for this elimination as you apparently have made no
- design changes that contradict this design basis. At any rate, the loss of Lake Wylie should be considered as part of the design basis in conjunction with a LOCA (refer to Question 11 above).
l Response:
See response to question 12 above.
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CNS
- 9. 2 WATER SYSTEMS 9.2.1 NUCLEAR SERVICE WATER SYSTEM 9.2.1.1 Design Bases The Nuclear Service Water System (RN) provides essential auxiliary support functions to Engineered Safety Features of the station. The system is designed l
to supply cooling water to various heat loads in both the safety and non-safety portions of each unit. Provisions are made to ensure a continuous flow of cooling water to those systems and components necessary for plant safety during normal operation and under accident conditions. Sufficient redundancy of piping and components is provided to ensure that cooling is maintained to essential loads at all times. See Table 3.2.2-2 for a listing of RN System component design codes, locations, missile protection and seismic consideration.
9.2.1.2 System Description The Nuclear Service Water System is shown diagramatically on Figures 9.2.1-1.
through 9.2.1-12. The piping and components shown on Figures 9.2.1-1 through 4 are shared between units, while the piping and component; shown on Figures 9.2.1-5 through 12 are duplicated for each unit unless otherwise stated in the following text. Functionally the system consists of four sections which, when i
j put together in series, serve to assure a supply of river water to various
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< station heat loads and return the heated effluent back to its proper heat sink.
In order of flow, these are:
- a. Source and intake section
- b. RN Pumphouse section
- c. Station heat exchanger section
- d. Main discharge section 9.2.1.2.1 Source and Intake Section 1
Two bodies of water serve as the ultimate heat sink for the components cooled by the RN System. Lake Wylie is the normal source of nuclear service water. A single transport line conveys water from a Class 1 seismically designed intake i structure at the bottom of the lake to both the A and B pits of the Nuclear i I
Service Water Pumphouse serving the RN pumps in operation. Isolation of each line is assured by two valves in series and fitted with electric motor opera-tors powered from separate power supplies.
l Should Lake Wylie be lost due to a seismic event in excess of the design of Wylie Dam; the Standby Nuclear Service Water Pond (SNSWP), formed by the Class 1 seismically designed SNSWP Dam, contains sufficient water to bring the station safely ,to a cold shutdown conditionj The SNSWP has an intake l l
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{ structure redundantdesigned to Cl_ ass lines to transport 1 seismic water requirements, independently with to each pit twoRN in the Class 1 seismic, Pump-I house. Each line is secured by a single motor operated valve. Automatically pon loss of Lake Wylie (as detected by RN pit level instrumentation), [
l v~ha + 4Tc 4 >-Etc 4 .
9.2-1 l _ _ _ _ - - _
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CNS Lake Wylie double isolation valves are closed and the SNSWP valves are opened l to both pit A and pit B. ,
The Nuclear Service Water lines cross over the condenser cooling water lines.
These CCW lines are low pressure lines and could only affect the NSW lines by undermining the surrounding soil due to a possible loss of cooling water.
Detection of this loss and system shutdown would occur prior to any detrimental effects to the NSW lines; further, the NSW lines are self-supporting over a considerable distance should any undermining occur.
Ultimate heat sink adequacy is discussed and analyzed in Section 9.2.5.
9.2.1.2.2 RN Pumphouse Section f' The RN Pumphouse is a Class 1 seismically designed structure that contains two separate pits from which two independent and redundant channels of RN pumps take suction. Each pit can be supplied from both the normal source and also the assured source of water. Either pit is capable of passing the flow needed for a simultaneous unit LOCA and unit cooldown. Flow spreaders in front of#all l the intake pipe entrances prevent vortices and flow irregularities while removable lattice screens protect the RH pumps from solid objects.
Pumps 1A and 2A take suction from pit A and discharge through RN strainers 1A I
and 2A.respectively. The ontlet piping of the 1A and 2A RN strainers then join back together to form the channel A Supply line to channel A components in both units, ,
RN pumps 1B and 28 are physically separated from RN pumps 1A and 2A by a l concrete wall, and take suction from pit 8, discharging through RN strainers 18 J and 2B respectively. The outlet piping of strainers 18 and 2B join together to form the channel B supply line to channel B components in both units. See Table 9.2.1-1 for a listing of RN System component design parametets. l Outside the Auxiliary Building wall, the channel A supply line splits, with 1A supply header entering on the Unit 1 side, isolated by an EMO valve powered by !
the 1A normal and assured power supplies, and the 2A supply header entering the l building on the Unit 2 side, isolated by an EHO valve powered by the 2A normal and assured power supplies.
Likewise, the channel B supply line splits with the 1B supply header entering on the Unit 1 side of the Auxiliary Building and the 2B supply header entering on the Unit 2 side, each isolated by EMO valves powered by corresponding normai ;
and assured power supplies. l The 1 and return headers are arranged and fitted with isol such that ical crack in either header can b - e and will not
_g jeopardize the sa e ctions of thi m or flood out other safety V i related equipment. The ope any two pumps on either or both supply
$ 8' lines is sufficient -
pply all 'n water requirements for the two unit plant for artup, cooldown, and re u However additional pumps a ly started for unit startup and cooldown fueling.
(are 9.2-2
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l Revision to Section 9.2.1.2.2 (page 9.2-2)
Revise the sixth paragraph as follows:
The supply and return headers ala i.rranged and fitted with isolation valves such that a critical crack in either header can be isolated and will not jeopardize the safety functions of this syatem or flood out other safoty related equipment. The opericion of any two pumps on either or both supply lines is sufficient to supply a'l cooling water requirements for unit startup, cooldown, refueling and post-accident operation of two units. However, ono pump has sufficient capacity to supply all cooling water requirements during normal power operation of both units or during post accident conditions if the unaffected unit is already in cold shutdown. All pumps (two per unit) are started during_the hypothetical combined accident and loss of normal power. In an accident, the safety injection signal automatically starts both RN pumps on each unit, thus providing complete rrdundancy.
Add a paragraph between the sixth and seventh paragraphs as follows:
If a diesel generator (or an RN pump) is out-of-service for an extended period of time (then, its associated unit is in cold shutdown), one RN pump is sufficient to provide adequate cooling water requirements for the operating unit and maintain the other unit in cold shutdown in the event of a hypothetical combined accident and loss of normal power.
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t CNS System flow demands outside of the RN pumphouse. Nominal Nuclear Service Water Flow System flow demands inside of the RN pumphouse are listed separately in Table 9.2.1-5.
Essential components receiving Nuclear Ser' lice Water flow are described below:
The RN pump motors are of the totally enclosed, water cooled type which have internal water-to-air heat exchangers. Cooling water is provided to the RN pump motor coolers only when the motor is in operation. This prevents the formation of condensate in the motor internals by the passage of cold water through an idle motor. The control valves for the RN pump motor coolers are manually set. ,
The RN pump motor upper bearing oii coolers are supplied cooling flow only when their respective RN pumps are in operation to prevent harmful condensation from forming in the oil. The RN pump motor coolers and RN pump motor upper bearing oil cooler on each pump are located downstream. A motor op e ated isolation valve is interlocked to open when the pump motor starts and close when the pump motor stops. ,.
Bearing lobe injection flow is maintained to all RN pumps at all times c even- -
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h=;;h erb := m 4 e 99Mt mt :P the wre! 2M sccMat fW l res_i c. s te vi L e u,o m. This water is supplied through redundant self-cleaning strainers. One strainer is supplied per train. A crossover allows a single operating RN pump to supply its own bearing lube injection flow plus that of the redundant channel RN pumps. Upon Engineered Safety Features actuation, all four pumps start and the crossover valves close, allowing each channel to supply the bearing lube requirements of its corresponding channel RN pumps.
. The nuclear service water strainers backflush automatically on a time cycle unless overridden by a pre-set high pressure drop. Internal water pressure is the motive force for dislodging strained particles as a backflush drive motor turns a backwash arm past the various strainer assemblies. The discharge is released to atmospheric pressure and dumps into a trash basket.outside the RN Entrained trash is collected and the water is returned to the
! Pumphouse.
Standby Nuclear Service Water Pond, which overflows to Lake Wylie.
Diesel generator engine starting air compressor af tercooler is supplied con-stantly as the compressor operates periodically to maintain starting air tank pressure. Flow is set by a manual throttling valve. Cooling water is supplied to the diesel generator engine jacket water cooler only when the diesel is in ;
operation. This is accomplished by an electric motor operated valve inter-locked to open when the diesel starts, close when the diesel stops. Flow is ;
assured to all diesel generators no matter which RN pumps are in operation by I the normal valve positions identified on Figure 9.2.1-2.
Those heat exchangers in which a tube leak could allow radioactive fluid tc enter the cooling water are cooled indirectly through the closed loop Component Cooling System (KC). Heat is then transferred to the RN System via the compo-nent cooling heat exchanger. The heat load provided by the RN normal loads will probably provide RN pump minimum flow requirements, but should this not be 9.2-5 1 i