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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217F8911999-10-13013 October 1999 Forwards Copy of FEMA Region IV Final Rept for 990623-24, Grand Gulf Nuclear Station Exercise.Rept Indicates No Deficiencies or Areas Requiring Corrective Action Identified During Exercise ML20216J8891999-10-0404 October 1999 Forwards Details of Existing Procedural Guidance & Planned Administrative Controls.Util Respectfully Requests NRC Review & Approval of Changes by 991020.Date Will Permit to Implement Changes & Realize Full Benefit During Refueling ML20217B0361999-10-0404 October 1999 Refers to Investigation Conducted by NRC OI Re Activities at Grand Gulf Nuclear Station.Investigation Conducted to deter- Mine Whether Security Supervisor Deliberately Falsified Unescorted Access Authorizations.Allegation Unsubstantiated ML20212J8151999-09-29029 September 1999 Forwards Insp Rept 50-416/99-12 on 990725-0904.One Violation Noted & Being Treated as Noncited Violation.Licensee Conduct of Activities at Grand Gulf Facility Characterized by Safety Conscious Operations,Sound Engineering & Maint Practices ML20216J6811999-09-28028 September 1999 Ack Receipt of ,Transmitting Rev 31 to Physical Security Plan for GGNS Under Provisions of 10CFR50.54(p). NRC Approval Not Required,Based on Determination That Changes Do Not Decrease Effectiveness & Limited Review ML20212J7361999-09-28028 September 1999 Forwards Insp Rept 50-416/99-11 on 990830-0903.No Violations Noted.Purpose of Insp to Review Solid Radioactive Waste Management & Radioactive Matl Transportation Programs ML20212J5321999-09-27027 September 1999 Forwards Insp Rept 50-416/99-14 on 990830-0903.No Violations Noted.Inspectors Determined That Radioactive Waste Effluent Releases Properly Controlled,Monitored & Quantified ML20216J7101999-09-26026 September 1999 Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator License Examinations ML20216J8141999-09-26026 September 1999 Forwards Proprietary Renewal Applications for Licensed Operators for Wk Gordon & SA Elliott at Grand Gulf Nuclear Station.Proprietary Info Withheld ML20212F5521999-09-23023 September 1999 Forwards SER Accepting Util Analytical Approach for Ampacity Derating Determinations at Grand Gulf Nuclear Station,Unit 1 & That No Outstanding Ampacity Derating Issues as Identified in GL 92-08 Noted ML20212D9211999-09-16016 September 1999 Informs That NRC Staff Completed Midcycle PPR of GGNS on 990818 & Identified No Areas in Which Licensee Performance Warranted Insp Beyond Core Insp Program.Details of Insp Plan Through March 2000 Encl ML20212A9331999-09-13013 September 1999 Forwards Partially Withheld Insp Rept 50-416/99-15 on 990816-20 (Ref 10CFR73.21).One Violation of NRC Requirements Occurred & Being Treated as Ncv,Consistent with App C of Enforcement Policy ML20211P7631999-09-10010 September 1999 Discusses Staff Issuance of SECY-99-204, Kaowool & FP-60 Fire Barriers at Plant.Proposed Meeting to Discuss Subj Issues Will Take Place in Oct or Nov 1999 ML20212A6951999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20212A8341999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20211Q3471999-09-0909 September 1999 Forwards Federal Emergency Mgt Agency Final Rept for 990623 Plant Emergency Preparedness Exercise.No Deficiencies Noted & One Area Requiring Corrective Action Identified ML20211Q3091999-09-0909 September 1999 Forwards Safety Evaluation Accepting BWROG Rept, Prediction of Onset of Fission Gas Release from Fuel in Generic BWR, Dtd July 1996 ML20211Q4861999-09-0808 September 1999 Informs That Util Has Discovered Dose Calculation Utilized non-conservative Geometry Factor for Parameter.Calculation Error Being Evaluated in Accordance with Corrective Action Program ML20211Q0091999-09-0808 September 1999 Forwards Request for Addl Info Re Individual Plant Exam of External Events for Grand Gulf Nuclear Station,Unit 1. Response Requested by 000615 ML20211N4301999-09-0808 September 1999 Discusses Proposed Meeting to Discuss Kaowool Fire Barriers. Staff Requesting That Affected Licensees Take Issue on Voluntary Initative & Propose Approach for Resolving Issues ML20211P4171999-09-0707 September 1999 Ack Receipt of ,Which Transmitted Addendum to Rev 30 to Physical Security Plan for Ggns,Per 10CFR50.54(p).NRC Approval Is Not Required,Since Util Determined That Changes Do Not Decrease Effectiveness of Plan ML20211P4121999-09-0707 September 1999 Requests NRC Staff Review & Approval of Integrated Nuclear Security Plan (Insp) & Integrated Security Training & Qualification Plan (Ist&Q), for Use by All Entergy Operations,Inc.Encl Withheld,Per 10CFR2.790(d) ML20211K6061999-08-31031 August 1999 Informs That Plant Has No Candidates to Take 991006 Generic Fundamentals Exam ML20211K5641999-08-31031 August 1999 Forwards Rev 39 to Grand Gulf Nuclear Station Emergency Plan Non-Safety Related, IAW 10CFR50,App E,Section V. Changes Do Not Decrease Effectiveness of Plan & Continues to Meets Stds of 10CFR50.47(b) & Requirements of App E ML20211J2321999-08-26026 August 1999 Advises That Info Contained in to Support NRC Review of GE Rept, Prediction of Onset of Fission Gas Release from Fuel in Generic BWR, Will Be Withheld from Public Disclosure ML20211J3761999-08-25025 August 1999 Corrected Ltr Informing That Info Provided (on Computer Disk & in Ltr to Ineel ) Marked as Proprietary Will Be Withheld from Public Disclosure Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954,as Amended.Corrected 990827 ML20211F4881999-08-25025 August 1999 Advises That Info Submitted by 990716 Application & Affidavit Containing Diskette & to Ineel Mareked Proprietary Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) of AEA of 1954 ML20211F7751999-08-24024 August 1999 Forwards Insp Rept 50-416/99-10 on 990809-13.No Violations Noted.Insp Covered Licensed Operator Requalification Program & Observations of Requalification Activities ML20211C4381999-08-20020 August 1999 Forwards Rev 31 to Physical Security Plan for Protection of Grand Gulf Nuclear Station,Iaw 10CFR50.54(p).Util Has Determined That Rev Does Not Decrease Effectiveness of Plan. Encl Withheld,Per 10CFR73.21 ML20211C5101999-08-19019 August 1999 Forwards Certified Copies of Liability Insurance Policy Endorsements Issued in First Half of 1999 for Each Entergy Operations,Inc Nuclear Unit,Per 10CFR140.15 ML20211B3761999-08-16016 August 1999 Submits Voluntary Response to NRC AL 99-02, Operating Reactor Licensing Actions Estimates, for Fys 2000 & 2001, ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl ML20211A9481999-08-12012 August 1999 Informs of Completion of Analysis of Heat Transfer in Cooler During Fan Coast Down & Concludes That Potential Exists for Steam Foundation,Under Conditions Where Dcw Sys Flow Is Lost Prior to Full Isolation Valve Closure ML20210P8411999-08-0909 August 1999 Forwards Insp Rept 50-416/99-09 on 990613-0724.No Violations Noted.Activities at Facility Generally Characterized by safety-conscious Operations,Sound Engineering & Maint Practices & Careful Radiological Work Controls ML20210N6401999-08-0303 August 1999 Informs That Eighteeen Identified Penetrations Will Be Restored to Conformance with Licensing Requirements Prior to Restart from RFO10,scheduled for Fall 1999,per GL 96-06. Example of Piping Analysis Being Performed,Encl ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams ML20210K1951999-07-30030 July 1999 Forwards Insp Rept 50-416/99-03 on 990405-08 & 0510-11.No Violations Identified ML20211K7491999-07-30030 July 1999 Forwards Ltr Rept Documenting Work Completed Under JCN-W6095,analyses Performed at Ineel to Calculate Minimum Time to Fuel Pin Failure in Boiling Water Reactors (BWR) ML20210K6661999-07-29029 July 1999 Forwards Fitness for Duty Program Performance six-month Rept for Period Covering Jan-June 1999,per 10CFR26.71 ML20210F3591999-07-26026 July 1999 Forwards Proprietary Version & Redacted Version of Wyle Test Rept M-J5.08-Q1-45161-0-8.0-1-0,re Pressure Locking & Thermal Binding Test Program.Proprietary Version Withheld ML20210E3251999-07-23023 July 1999 Forwards Insp Rept 50-416/99-07 on 990622-25.No Violations Noted.Emergency Plan & Procedures During Biennial Emergency Preparedness Exercise Was Conducted ML20210D2401999-07-21021 July 1999 Informs of Resignation of Operator WE Griffith,License OP-20806-1,from Entergy Operations,Inc ML20209J0311999-07-16016 July 1999 Forwards Proprietary Info Supporting Review of Generic Alternate Source Term Request.Proprietary Info Withheld Per 10CFR2.790 ML20209G4791999-07-15015 July 1999 Forwards Proposed Emergency Plan Change as Addendum to Changes Previously Submitted Via GNRO-98/00028 & GNRO-99/00007,for NRC Review & Approval ML20210B1031999-07-15015 July 1999 Forwards Insp Rept 50-416/99-08 on 990502-0612.Determined That Three Severity Level IV Violations Occurred & Being Treated as Noncited Violations ML20210H3211999-07-14014 July 1999 Forwards Proprietary Info Supporting Review of 970506 Submittal of BWROG Rept, Prediction of Onset of Fission Gas Release from Fuel in Generic Bwr. Proprietary Info Withheld Per 10CFR2.790 ML20209D7511999-07-0909 July 1999 Responds to RAI on GL 92-01,rev 1,suppl 1, Rv Structural Integrity. as Result of NRC Review of Util Responses,Info Revised in Rvid & Rvid Version 2 Will Be Released ML20209D7671999-07-0101 July 1999 Submits Response to Violations Noted in Insp Rept 50-416/99-02 on 990222-26 & 0308-12.Corrective Actions: Contractor Performance Has Been re-evaluated in Regards to UFSAR Reviews ML20196K4901999-07-0101 July 1999 Discusses Relief Requests PRR-E12-01,PRR-E21-01,PRR-E22-01, PRR-P75-01,PRR-P81-01,VRR-B21-01,VRR-B21-02,VRR-E38-01 & VRR-E51-01 Submitted by EOI on 971126 & 990218.SE Accepting Alternatives Proposed by Util Encl ML20196J5711999-06-30030 June 1999 Advises That Versions of Submitted Info in 990506 Application & Affidavit, Re Proposed Amend to Revise Ts,Marked Proprietary Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) of AEA 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20216J8891999-10-0404 October 1999 Forwards Details of Existing Procedural Guidance & Planned Administrative Controls.Util Respectfully Requests NRC Review & Approval of Changes by 991020.Date Will Permit to Implement Changes & Realize Full Benefit During Refueling ML20216J7101999-09-26026 September 1999 Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator License Examinations ML20216J8141999-09-26026 September 1999 Forwards Proprietary Renewal Applications for Licensed Operators for Wk Gordon & SA Elliott at Grand Gulf Nuclear Station.Proprietary Info Withheld ML20211Q4861999-09-0808 September 1999 Informs That Util Has Discovered Dose Calculation Utilized non-conservative Geometry Factor for Parameter.Calculation Error Being Evaluated in Accordance with Corrective Action Program ML20211P4121999-09-0707 September 1999 Requests NRC Staff Review & Approval of Integrated Nuclear Security Plan (Insp) & Integrated Security Training & Qualification Plan (Ist&Q), for Use by All Entergy Operations,Inc.Encl Withheld,Per 10CFR2.790(d) ML20211K6061999-08-31031 August 1999 Informs That Plant Has No Candidates to Take 991006 Generic Fundamentals Exam ML20211K5641999-08-31031 August 1999 Forwards Rev 39 to Grand Gulf Nuclear Station Emergency Plan Non-Safety Related, IAW 10CFR50,App E,Section V. Changes Do Not Decrease Effectiveness of Plan & Continues to Meets Stds of 10CFR50.47(b) & Requirements of App E ML20211C4381999-08-20020 August 1999 Forwards Rev 31 to Physical Security Plan for Protection of Grand Gulf Nuclear Station,Iaw 10CFR50.54(p).Util Has Determined That Rev Does Not Decrease Effectiveness of Plan. Encl Withheld,Per 10CFR73.21 ML20211C5101999-08-19019 August 1999 Forwards Certified Copies of Liability Insurance Policy Endorsements Issued in First Half of 1999 for Each Entergy Operations,Inc Nuclear Unit,Per 10CFR140.15 ML20211B3761999-08-16016 August 1999 Submits Voluntary Response to NRC AL 99-02, Operating Reactor Licensing Actions Estimates, for Fys 2000 & 2001, ML20211A9481999-08-12012 August 1999 Informs of Completion of Analysis of Heat Transfer in Cooler During Fan Coast Down & Concludes That Potential Exists for Steam Foundation,Under Conditions Where Dcw Sys Flow Is Lost Prior to Full Isolation Valve Closure ML20210N6401999-08-0303 August 1999 Informs That Eighteeen Identified Penetrations Will Be Restored to Conformance with Licensing Requirements Prior to Restart from RFO10,scheduled for Fall 1999,per GL 96-06. Example of Piping Analysis Being Performed,Encl ML20211K7491999-07-30030 July 1999 Forwards Ltr Rept Documenting Work Completed Under JCN-W6095,analyses Performed at Ineel to Calculate Minimum Time to Fuel Pin Failure in Boiling Water Reactors (BWR) ML20210K6661999-07-29029 July 1999 Forwards Fitness for Duty Program Performance six-month Rept for Period Covering Jan-June 1999,per 10CFR26.71 ML20210F3591999-07-26026 July 1999 Forwards Proprietary Version & Redacted Version of Wyle Test Rept M-J5.08-Q1-45161-0-8.0-1-0,re Pressure Locking & Thermal Binding Test Program.Proprietary Version Withheld ML20210D2401999-07-21021 July 1999 Informs of Resignation of Operator WE Griffith,License OP-20806-1,from Entergy Operations,Inc ML20209J0311999-07-16016 July 1999 Forwards Proprietary Info Supporting Review of Generic Alternate Source Term Request.Proprietary Info Withheld Per 10CFR2.790 ML20209G4791999-07-15015 July 1999 Forwards Proposed Emergency Plan Change as Addendum to Changes Previously Submitted Via GNRO-98/00028 & GNRO-99/00007,for NRC Review & Approval ML20210H3211999-07-14014 July 1999 Forwards Proprietary Info Supporting Review of 970506 Submittal of BWROG Rept, Prediction of Onset of Fission Gas Release from Fuel in Generic Bwr. Proprietary Info Withheld Per 10CFR2.790 ML20209D7671999-07-0101 July 1999 Submits Response to Violations Noted in Insp Rept 50-416/99-02 on 990222-26 & 0308-12.Corrective Actions: Contractor Performance Has Been re-evaluated in Regards to UFSAR Reviews ML20209B6081999-06-30030 June 1999 Submits Response to NRC GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Nuclear Power Plants. Disclosure Encl ML20195J6351999-06-16016 June 1999 Forwards Addendum to Rev 30 of GGNS Physical Security Plan IAW 10CFR50.54(p).Addendum Is Submitted to Announce Relocation/Reconfiguration of Plant Central & Secondary Alarm Station Facilities.Rev Withheld,Per 10CFR73.21 ML20195G0281999-06-0909 June 1999 Submits Summary on Resolution of GL 96-06 Re Eighteen Penetrations Previously Identified as Being Potentially Susceptible to Overpressurization ML20207F5041999-06-0202 June 1999 Forwards Updated Medical Rept IAW License Condition 3 for DA Killingsworth License OP-20942-1.Without Encls ML20206P2981999-05-13013 May 1999 Forwards Responses to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, Cancelling 990402 Submittal ML20206N1921999-05-10010 May 1999 Provides Revised Attachment 2 for Alternative Request IWE-02,originally Submitted 990429 Re Bolt Torque or Tension Testing of Class Mc pressure-retaining Bolting as Specified in Item 8.20 of Article IWE-2500,Table IWE-2500-1 ML20206J0941999-05-0404 May 1999 Forwards Proprietary & Redacted ME-98-001-00,both Entitled, Pressure Locking & Thermal Binding Test Program on Two Gate Valves with Limitorque Actuators. Rept ME-98-002-00 Re Flexible Wedge Gate Valves,Encl.Proprietary Rept Withheld ML20206E7811999-04-29029 April 1999 Proposes Alternatives to Requirements of ASME B&PV Code Section XI,1992 Edition,1992 Addenda,As Listed.Approval of Alternative Request on or Before 990915,requested ML20206D8171999-04-29029 April 1999 Informs NRC of Results of Plant Improvement Considerations Identified in GGNS Ipe,As Requested in NRC . Licensee Found Efforts Have Minimized Extent of Radiological Release in Unlikely Event That Severe Accident Occurred ML20206D7281999-04-28028 April 1999 Forwards South Mississippi Electric Power Association 1998 Annual Rept, Per 10CFR50.71(b).Licensee Will Submit 1998 Annual Repts for System Energy Resources,Inc,Entergy Mississippi,Inc & EOI as Part of Entergy Corp Annual Rept ML20206C9551999-04-22022 April 1999 Forwards 1999 Biennial Emergency Preparedness Exercise Scenario. Without Encl ML20205M1311999-04-0202 April 1999 Forwards Response to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves. Info Was Discussed During Conference Call with NRC on 990126.Wyle Position Paper Encl.Subj Paper Withheld ML20205H5861999-04-0101 April 1999 Requests Relief from ASME B&PV Code,Section XI for Period of Time That Temporary non-code Repair Was in Effect,Per 10CFR50.55a(g)(5)(iii) ML20205F1781999-03-31031 March 1999 Forwards Consolidated Entergy Submittal to Document Primary & Excess Property Damage Insurance Coverage for Nuclear Sites of Entergy Operations,Inc,Per 10CFR50.54(w)(3) ML20196K7101999-03-26026 March 1999 Submits Reporting & Recordkeeping for Decommissioning Planning,Per 10CFR50.75(f)(1) ML20205A6511999-03-25025 March 1999 Responds to NRC Re Violations Noted in Insp Rept 50-416/99-01 on 990201-05.Corrective Actions:Program Will Be Implemented to Ensure Accessible Areas with Radiation Levels Greater than 1000 Mrem/H ML20204E7391999-03-15015 March 1999 Forwards Objectives for June 1999 Emergency Preparedness Exercise for Plant.Without Encl ML20207H9291999-03-0404 March 1999 Submits Update to Original Certification of Grand Gulf Nuclear Station Simulation Facility IAW Requirements of 10CFR55.45(b)(5) ML20207E3081999-03-0303 March 1999 Informs That GGNS Severe Accident Mgt Implementation Was Completed on 981223.Effort Was Worthwhile & Station Ability to Respond & Mitigate Events That May Lead to Core Melt Has Been Enhanced ML20207E3221999-03-0303 March 1999 Notifies of Change in Status of Mj Ellis,License SOP-43846. Conditional License Requested to Accommodate Medical Condition.Revised NRC Form 396 with Supporting Medical Evidence Attached.Without Encls ML20207A8161999-02-24024 February 1999 Forwards 1998 Annual Operating Rept for Ggns,Unit 1. Listed Attachments Are Encl ML20207A9901999-02-24024 February 1999 Informs That Util Has No Candidates from GGNS to Nominate for Participation in Planned Gfes,Scheduled for 990407 ML20203A1551999-02-0101 February 1999 Forwards Grand Gulf Nuclear Station Fitness for Duty Program Performance six-month Rept for Reporting Period 980701-981231 ML20202G0791999-01-26026 January 1999 Informs That He Mcknight Has Been Permanently Reassigned from Position Requiring License to Perform Assigned Duties. License Is No Longer Needed,Effective 981231 ML20199K4151999-01-20020 January 1999 Forwards Proposed Addendum to Emergency Plan Changes Previously Submitted Via GNRO-98/00028 for NRC Review & Approval as Required by 10CFR50.54(q) & 50.4 ML20199K6771999-01-14014 January 1999 Provides Notification of Planned ERDS Software Change Scheduled to Take Place on 990215 ML20199D8811999-01-11011 January 1999 Submits Response to SE JOG Program on Periodic Verification of motor-operated Valves,In Response to GL 96-05 ML20199D9521999-01-0808 January 1999 Informs That CE Cresap,License SOP-4220-4,has Been Permanently Reassigned from Position Requiring License & No Longer Has Need for License,Per 10CFR50.74 ML20199A6081999-01-0606 January 1999 Submits List of Plant Info Brochures Disseminated Annually to Public & List of Updated State &/Or Local Emergency Plan Info,Per NRC Administrative Ltr 94-07, Distribution of Site-Specific & State Emergency Planning Info ML20202B7531998-12-21021 December 1998 Submits Ltr Confirming Discussion with J Tapia,Documenting Extension for Response to NOV 50-416/98-13.Util Response Will Be Submitted by 990212 1999-09-08
[Table view] Category:UTILITY TO NRC
MONTHYEARAECM-90-0169, Forwards Operator Licensing Natl Exam Schedule for FY91 Through FY94,per Generic Ltr 90-071990-09-17017 September 1990 Forwards Operator Licensing Natl Exam Schedule for FY91 Through FY94,per Generic Ltr 90-07 AECM-90-0172, Forwards Endorsement 67 to Nelia Policy NF-257 & Endorsement 46 to Maelu Policy MF-1061990-09-17017 September 1990 Forwards Endorsement 67 to Nelia Policy NF-257 & Endorsement 46 to Maelu Policy MF-106 AECM-90-0174, Forwards List of Submittals Pending NRR Review Re Grand Gulf Licensing Activities1990-09-14014 September 1990 Forwards List of Submittals Pending NRR Review Re Grand Gulf Licensing Activities AECM-90-0165, Forwards Addl Info on NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount1990-09-12012 September 1990 Forwards Addl Info on NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount AECM-90-0158, Forwards Quarterly Status Rept for Reg Guide 1.97 Re Neutron Monitoring Sys for Period Ending 900630.Rept Includes Major Actions Completed to Date for Unit ex-core Sys.Estimated Milestone Schedule for Activities Also Encl1990-09-0808 September 1990 Forwards Quarterly Status Rept for Reg Guide 1.97 Re Neutron Monitoring Sys for Period Ending 900630.Rept Includes Major Actions Completed to Date for Unit ex-core Sys.Estimated Milestone Schedule for Activities Also Encl AECM-90-0163, Forwards Endorsement 61 to Nelia Policy NF-257,Endorsement 40 to Maelu Policy MF-106,Endorsement 62 to Nelia Policy NF-257,Endorsement 41 to Maelu Policy MF-106 & Endorsement 63 to Nelia Policy NF-2571990-09-0606 September 1990 Forwards Endorsement 61 to Nelia Policy NF-257,Endorsement 40 to Maelu Policy MF-106,Endorsement 62 to Nelia Policy NF-257,Endorsement 41 to Maelu Policy MF-106 & Endorsement 63 to Nelia Policy NF-257 AECM-90-0161, Forwards Quarterly Status Rept Re Degraded Core Accident Hydrogen Control Program, for Apr-June 19901990-08-30030 August 1990 Forwards Quarterly Status Rept Re Degraded Core Accident Hydrogen Control Program, for Apr-June 1990 AECM-90-0149, Forwards Semiannual Radioactive Effluent Release Rept for Jan-June 1990 & Rev 3 to Process Control Program1990-08-30030 August 1990 Forwards Semiannual Radioactive Effluent Release Rept for Jan-June 1990 & Rev 3 to Process Control Program AECM-90-0162, Forwards fitness-for-duty 6-month Rept for Period Ending June 1990,per 10CFR26.Success of Program Evident in Statistical Data Indicating Extremely Low Incident Rate1990-08-29029 August 1990 Forwards fitness-for-duty 6-month Rept for Period Ending June 1990,per 10CFR26.Success of Program Evident in Statistical Data Indicating Extremely Low Incident Rate ML20028G8591990-08-27027 August 1990 Forwards Updated Svc List to Be Used for Licensee Correspondence.Requests That Author Be Primary Addressee for All Correspondence Re Plant AECM-90-0144, Forwards Security Boundary Upgrade Bimonthly Status Rept for Period Ending 900731,per 900330 Commitment.Rept Covering Period 900801-0930 Will Be Submitted in Oct 19901990-08-22022 August 1990 Forwards Security Boundary Upgrade Bimonthly Status Rept for Period Ending 900731,per 900330 Commitment.Rept Covering Period 900801-0930 Will Be Submitted in Oct 1990 ML20056B3511990-08-20020 August 1990 Suppls Info Re 900806 Application for Amend to License NPF-29,changing Tech Specs on Alternate DHR Sys,Per NRC Comments.Proposed Tech Spec 3/4.5.2 Encl AECM-90-0147, Informs That Annual Emergency Preparedness Exercise for Facility Scheduled for Wk of 9108261990-08-14014 August 1990 Informs That Annual Emergency Preparedness Exercise for Facility Scheduled for Wk of 910826 AECM-90-0142, Forwards Supplemental Info Re 900705 Application for Amend to License NPF-29,revising Tech Specs Due to Addition of Alternate DHR Sys1990-08-0909 August 1990 Forwards Supplemental Info Re 900705 Application for Amend to License NPF-29,revising Tech Specs Due to Addition of Alternate DHR Sys AECM-90-0143, Notifies That Cd Bland No Longer Employed by Util,Effective 9007191990-08-0202 August 1990 Notifies That Cd Bland No Longer Employed by Util,Effective 900719 AECM-90-0139, Forwards Endorsement 68 to Nelia Policy NF-257,Endorsement 47 to Maelu Policy MF-106 & Revised Endorsement 35 to Maelu Policy MF-1061990-08-0202 August 1990 Forwards Endorsement 68 to Nelia Policy NF-257,Endorsement 47 to Maelu Policy MF-106 & Revised Endorsement 35 to Maelu Policy MF-106 ML20055J0551990-07-27027 July 1990 Forwards Summary of Environ Protection Program Re Const of Unit for 6-months Ending 900630,per Exhibit 2-A in Subsection 3.E.1 of CPPR-119 AECM-90-0136, Forwards Executed Amend 4 to Indemnity Agreement B-72,per NRC 891214 Request1990-07-27027 July 1990 Forwards Executed Amend 4 to Indemnity Agreement B-72,per NRC 891214 Request AECM-90-0130, Forwards Corrected Pages to Rev 17 to Physical Security Plan.Pages Withheld (Ref 10CFR73.21)1990-07-17017 July 1990 Forwards Corrected Pages to Rev 17 to Physical Security Plan.Pages Withheld (Ref 10CFR73.21) ML20044A9251990-07-0909 July 1990 Forwards Rev 1 to Relief Request I-00018 Correcting Valve Number & Description of One Component.Review & Approval Requested Prior to 901001 ML20044A7861990-06-29029 June 1990 Responds to NRC 900601 Ltr Re Violations Noted in Insp Rept 50-416/90-08.Corrective Actions:Operations Superintendent Counseled Individuals Re Inoperable Reactor Water Level Transmitter & Met W/All Shift Senior Reactor Operators AECM-90-0121, Withdraws 880831 & 890324 Proposed Amends,Deleting Certain Test,Vent & Drain Valves from Tech Spec Table 3.6.4-11990-06-27027 June 1990 Withdraws 880831 & 890324 Proposed Amends,Deleting Certain Test,Vent & Drain Valves from Tech Spec Table 3.6.4-1 AECM-90-0115, Forwards List of Followup Actions as Result of NRC Requalification Reexam of Three Licensed Operators on 900531.Lessons Learned Guideline Will Be Prepared Re Ability of Training Personnel to Evaluate Simulator Crew1990-06-26026 June 1990 Forwards List of Followup Actions as Result of NRC Requalification Reexam of Three Licensed Operators on 900531.Lessons Learned Guideline Will Be Prepared Re Ability of Training Personnel to Evaluate Simulator Crew ML20044A2931990-06-22022 June 1990 Responds to NRC Request for Addl Info Re Boraflex Gap Analysis.If Vibratory Ground Motion Exceeding OBE Occurs,Per 10CFR100,App a & as Previously Committed,Plant Will Be Shut Down.Listed Addl Surveillance Will Be Performed ML20043G6231990-06-14014 June 1990 Forwards Evidence That Cash Flow Would Be Available for Payment of Deferred Premium Obligation for Facility.Sys Energy Resources,Inc Responsible for Generating 90% of Required Cash Flow ML20043G3341990-06-11011 June 1990 Forwards Rev 9 to GGNS-TOP-1A, Operational QA Manual, for Evaluation ML20043G5861990-06-0808 June 1990 Forwards Bimonthly Status Repts Re Security Boundary Upgrade Project for Period Ending 900531 ML20043F5121990-06-0808 June 1990 Forwards List of Directors & Officers of Entergy Operations, Inc.Operation of All Plants Transferred to Entergy on 900606 ML20043E8011990-06-0707 June 1990 Forwards Nonproprietary ANF-90-060(NP), Criticality Safety Analysis for Grand Gulf Fuel Storage Racks W/ANF-1.4 Fuel Assemblies. ML20043E7831990-06-0707 June 1990 Forwards Updated Svc List to Be Used Re Plant Correspondence.Requests WT Cottle Be Primary Addressee for All Correspondence Concerning Plant ML20043E8161990-06-0606 June 1990 Informs That Sys Energy Resources,Inc Received Necessary Regulatory Approvals to Transfer Performance Activities for Facility to Entergy Operations & All Conditions in Amend 9 to CP CPPR-119 Implemented,Effective on 900606 ML20043F2061990-06-0606 June 1990 Forwards 1989 Annual Financial Repts for Sys Energy Resources,Inc & South Mississippi Electric Power Assoc ML20043E8111990-06-0606 June 1990 Informs That Sys Energy Resources,Inc Received Necessary Regulatory Approvals to Transfer Operating Responsibility for Facility to Entergy Operations & All Conditions in Amend 65 to License NPF-29 Implemented,Effective on 900606 ML20043C8611990-05-31031 May 1990 Forwards Preliminary Drafts of Plant Specific Tech Specs in Order to Facilitate NRC Validation of BWR Owners Group Improved Tech Specs,Per NRC Request.Understands That Util & NRC Will Meet During Wk of 900716 to Discuss NRC Review ML20043B6811990-05-24024 May 1990 Forwards Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Jan-Mar 1990 ML20043B6021990-05-23023 May 1990 Confirms NRC Understanding That Safety Evaluation Will Be Written for Use of New Tech Spec 3.0.4 Flexibility Regardless of Plant Condition at Time Flexibility Required ML20043B2471990-05-18018 May 1990 Forwards Final Response to Generic Ltr 89-04, Guidance on Developing Acceptable Inservice Testing Programs & Rev 4 to Pump & Valve Inservice Testing Program. ML20043A9651990-05-17017 May 1990 Forwards Draft Tech Specs for Power Distribution Limits,Rcs, ECCS & Plant Sys as Part of Util Involvement W/Bwr Owners Group as BWR-6 Lead Plant ML20042G6931990-05-0909 May 1990 Forwards Rev 4 to Fire Hazards Analysis. Design Changes Include Installation of Alternate DHR Sys & Access Hatch in Pipe Chase ML20042G8681990-05-0909 May 1990 Forwards Response to Recommendations Re Areas of Concern Noted in NRC SER Dtd 900316 & 900316 Request for Addl Info Re Design Criteria for Cable Tray Supports in Turbine Bldg ML20042G6731990-05-0909 May 1990 Notifies of Cancellation of Emergency Plan Procedure 10-S-01-13, Onsite Radiological Monitoring. Info Incorporated Into Procedure 10-S-01-14,Rev 13, Radiological Monitoring. ML20042F4891990-05-0404 May 1990 Requests Extension of 90 Days to Provide Addl Time for Securities & Exchange Commission Review Re Implementation of Amend 65 to License NPF-29 ML20042F4441990-05-0404 May 1990 Forwards Response to Generic Ltr 89-19 Re USI A-47, Safety Implications of Control Sys in LWR Nuclear Power Plants. Plant Has Adequate Automatic Reactor Vessel Overfill Protection,Procedures & Tech Specs ML20042F1791990-04-30030 April 1990 Responds to NRC 900402 Ltr Re Violations Noted in Insp Rept 50-416/90-03.Corrective Actions:Valves Closed,Effectively Isolating Flow of Contaminated Water Into Makeup Water Sys & Demineralized Water Sys Flushed & Cleaned of Contamination ML20042F1811990-04-30030 April 1990 Responds to Generic Ltr 89-15, Emergency Response Data Sys. Util Volunteers to Participate in Emergency Response Data Sys ML20042F3711990-04-30030 April 1990 Forwards Certificate of Insurance for Nuclear Property Insurance Submitted by Nuclear Mutual Ltd for Policy Period 900401-910401 & Certificate of Insurance Evidencing Increased Excess Property Insurance,Per 900330 Ltr ML20042F1751990-04-30030 April 1990 Advises That Util Will Not Be Able to Provide Complete Supplemental Summary Rept on Dcrdr by 900430,as Indicated in Util 891221 Ltr.Supplemental Rept Will Be Submitted by 900930 ML20012F3311990-04-0202 April 1990 Forwards GE Affidavit Requesting That All Drawings Presently Denoted as Proprietary in Rev 4 to Updated FSAR Re Offgas Sys Should Remain Proprietary (Ref 10CFR2.790) ML20012E2961990-03-26026 March 1990 Forwards Updated Svc List for NRC Correspondence to Util. Facility Fee Bills Sent to Wrong Primary Addressee ML20011F2171990-02-23023 February 1990 Responds to NRC 900131 Ltr Re Violations Noted in Insp Rept 50-416/89-30.Corrective Actions:Quality Deficiency Rept Initiated to Document & Resolve Incident & Incident Rept & Reportable Events Procedure Enhanced 1990-09-08
[Table view] |
Text
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O O
MISSISSIPPI POWER & LIGHT COMPANY
] Helping Build Mississippi P. O. BOX 1640 J A C K S O N, MIS SIS SIP PI 39215-1640 EdhhilHidd5 March 29, 1985 NUCLEAR LICENSING & SAFETY DEPARTMENT U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C. 20555 Attention: Mr. Harold R. Denton, Director
Dear Mr. Denton:
SUBJECT:
Grand Gulf Nuclear Station Units 1 and 2 Docket Nos. 50-416 and 50-417 License No. NPF-29 File: 0260/0755/6450 Additional Information on NUREG-0737 Item II.D.1 AECM-85/0099 Enclosed please find Mississippi Power & Light's response to your request for aiditional information dated December 12, 1984. This request for information concerned a report submitted to the NRC by the BWR Owners Group in response to NUREG-0737 Item II.D.1 entitled " Analysis of Generic BWR Safety Relief Valve Operability" (NEDE-24988-P) . Attachment 1 to this letter describes the basis for application of the BWR Owners Group test results to Grand Gulf Nuclear Station (GGNS) by responding to NRC questions one through six. Attachment 2 provides actual Wyle Laboratory test results for the GGNS valve which was tested.
This response completes Mississippi Power & Light's efforts on this issue.
Should you have any further questions regarding this item please advise.
Yours truly, 3M N L. F. Dale Director ARR/SHH:rw Attachments cc: (See Next Page) 8504040298 850329 dkO gDR ADOCK 05000416 \
PDR l
Member Middle South Utilities System J0P14AECM85032102 - 1 L.
2
-0 AECM-85/0099 Page 2 cc: Mr. J. B. Richard (w/a)
Mr. O. D. Kingsley, Jr. (w/a)
Mr. R. B. McGehee (w/a)
Mr. N.-S. Reynolds (w/a)
Mr. G. B. Taylor (w/o)
Mr. James M. Taylor, Director (w/a)
Office of Inspection & Enforcement U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Dr. J. Nelson Grace, Regional Administrator (w/a)
U. S. Nuclear Regulatory Commission Region II 101 Marietta St., N. W., Suite 2900 Atlanta, Georgia 30323 l
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i J0P14AECM85032102 - 2 i:
.m Attechm:nt I to AECM-85/0099 NRC OUESTION 1-The BWR/GE test program utilized a " rams head" discharge pipe configuration.
Most plants. utilize a " tee" quencher configuration at the end of the discharge line. Describe the discharge pipe configuration used at your plant and
- compare the anticipated loads in this configuration to the measured loads in the test program. Discuss the inpact of any differences in loads on valve operability.
RESPONSE TO QUESTION 1 The safety / relief valve (SRV) discharge piping configuration at Grand Gulf Muclear Station (GGNS) utilizes an "X" quencher at the discharge pipe exit.
The average length of the 20 SRV discharge lines is 94 feet and the submergence depth in the suppression pool is approximately 13.8 feet. The SRV test program utilized a rams head at the discharge pipe exit, a pipe length of 112 feet and a submergence depth of approximately 13 feet. Loads on valve internals during the test program are larger than loads on valve internals in the GGNS configuration for the following reasons:
- 1. No dynamic mechanical load originating at the "X" quencher is transmitted to the valve in the GGNS configuration because there is an anchor point between the valve and the "X" quencher.
- 2. The length of the SRV discharge line piping between the SRV and the first elbow in the test facility was about the same length as in the GGNS configuration (12 feet in the test facility, 12.1 feet in the the GGNS configuration). However, unlike the rigid test configuration, CGNS has ball joints in the piping spool between the SRV and the first anchor point which will rotate similarly to a hinge in response to any externally applied moment. Hence, the mechanical load on the GGNS SRV will be much lower than that on the test facility valve.
- 3. Dynamic hydraulic loads (backpressure) are experienced by the valve internals in the GGNS configuration. The backpressure loads may be either (i) transient backpressures occurring during valve actuation, or (ii) steady-state backpressures occurring during steady-state flow following valve actuation.
(a) The key parameters affecting the transient backpressures are the fluid pressure upstream of the valve, the valve opening time, the fluid inertia in the submerged safety /reifef valve discharge line (SRVDL) and the SRVDL air volume. Transient backpressures increase with higher upstream pressure, shorter valve opening times, greater line submergence and smaller SRVDL air volume. The transient backpressure in the test program was maximized by utilizing an orifice plate in the SRVDL to create a 35-40% backpressure condition on the valve internals, body bowl and discharge flange. This induced backpressure simulated the maximum backpressure anticipated in the GGNS SRVDLs. The maximum transient backpressure occurs with high pressure steam flow conditions. The transient backpressure for the alternate shutdown cooling mode of operation is always much less than the design for steam flow conditions because of the lower upstream pressure and the longer. valve opening time.
J0Pl4ATTC85032201 - 1
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AttachmInt I to AECM-85/0099 (b) -The steady-state backpressure in the test program was maximized by utilizing an orifice plate in the SRVDL above the water level and before the rams head. The orifice was sized to produce a backpressure equal to or greater than that calculated for any of the GGNS SRVDLs.
The differences in the line configurations between GGNS and the test program as discussed above result in loads on the valve internals for the test facility which bound the actual CGNS loads. An additional consideration in the selection of the rams head for the test facility was to allow more direct measurement of the thrust load in the final pipe segment. Utilization of a "X" quencher in the test program would have required quencher supports that would unnecessarily obscure accurate measurement of the pipe thrust loads.
For the reasons stated above, differences between the SRVDL configurations in GGNS and the test facility will not have any adverse effect on SRV operability at GGNS relative to the test facility.
NRC QUESTION 2 The test configuration utilized no spring hangers as pipe supports. Pirst specific configurations do use spring hangers in conjunction with snubber and rigid supports. Describe the safety / relief valve pipe supports used at your plant and compare the anticipated loads on valve internals for the plant pipe supports to the measured loads in the test program. Describe the impact of any differences in loads on valve operability.
RESPONSE TO OUESTION 2 The GGNS SRVDLs are supported by a combination of snubbers, rigid supports, anchors, and spring hangers. The locations of snubbers and rigid supports at GCNS are such that there are supports near each change of direction in the pipe routing. Additionally, only 15 of the 20 SRVDLs at GCNS have spring hangers (1 or 2), all of which are located in the drywell. The snubbers, rigid-supports, and the anchor between the SRV and the "X" quencher are designed to accommodate combinations of loads resulting from piping dead weight, thermal conditions, seismic and suppression pool hydrodynamic events, and a high pressure steam discharge transient. ,The spring hangers are designed for the operating weight loads and to accommodate the pipe movements due to thermal, seismic and dynamic events.
The dynamic load effects on the piping and supports of the test facility due to the water discharge event (the alternate shutdown cooling mode) were found to be significantly lower than corresponding loads resulting from the high pressure steam discharge event. As stated in NEDE-24988-P, this finding is considered generic to all BWR's since the test facility was designed to be prototypical of the features pertinent to this issue.
During the water discharge transient there will be significantly lower dynamic loads acting on the snubbers and rigid supports than during the steam discharge transient. This will more than offset the small increase in the dead load on these supports due to the weight of the water during the alternate shutdown cooling mode of operation. An analysis was performed on J0P14ATTC85032201 - 2
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AttachmInt 1 to AECM-85/0099 GGNS for the alternate shutdown cooling mode. The results of this analysis for a typical SRVDL show that the dynamic loads due to this mode are significantly lower than t'te steam discharge loads. Therefore, design adequacy of the SRV pipe supports is assured as these supports are designed for the. larger steam discharge transient loads.
This question addresses the design adequacy of the spring hangers with respect to the weight of the water during the liquid discharge transient. Due to the nature of the design of spring hangers there will be little increased load on the spring hangers because of the water weight. The results of an analysis for a typical SRVDL show that the increased loads are mostly taken by the nearest rigid vertical supports. Therefore, it is concluded that sufficient margin exists in the GGNS SRVDL support design to adequately offset the increased dead load on the spring hangers in an unpinned condition due to a water filled condition. Furthermore, the effect of the water dead weight load does not affect the ability of SRVs to open to establish the alternate shutdown cooling path since the loads occur in the SRVDL only after valve opening.
NRC QUESTION 3 Report NEDE-24988-P did not identify any valve functional deficiencies or anomalies encountered during the test program. Describe the impact on valve safety function of any valve functional deficiencies or anomalies encountered during the program.
RESPONSE TO QUESTION 3 No functional deficiencies or anomalies of the safety / relief or relief valves were experienced during the testing at Wyle Laboratories for compliance with the alternate shutdown cooling mode requirement. All of the valves subjected to test runs, valid and invalid, opened and closed without loss of pressure integrity or damage. Anomalies encountered during the test program were all due to failures of test facility instrumentation, equipment, data acquisition equipment, or deviation from the approved test procedure.
The test specification for each valve required six runs. Under the test procedure, an anomaly caused the test run to be ~ judged invalid. All anomalies were reported in the test report. The Wyle Laboratories test log sheet for the Dikkers'8X10 Dual Function Safety / Relief Valve tests is shown in Attachment 2.- This valve is used in the Grand Gulf Nuclear Station.
Each Wyle test report for the respective valves identifies each test run performed and documents whether or not the test run is valid or invalid and states the reason for considering the run invalid. No anomaly encountered during the required test program affects any valve safety or operability function.
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J0P14ATTC85032201 - 3
Attach =snt I to AECM-85/0099 All valid test runs are identified in Table 2.2-1 of NEDE-24988-P. The data presented in Table 4.2-1 for each valve was obtained from the Table 2.2-1 test runs and was based upon the selection criteria of:
.(a) Presenting the maximum representative loading information obtained from the steam run data, (b) Presenting the maximum representative water loading information obtained from the 15'F subcooled water test data, (c) Presenting the data on the only test run performed for the 50*F subcooled water test condition.
NRC OUESTION 4 The purpose of the test program was to determine valve performance under conditions' anticipated to be encountered in the plants. Describe the events and anticipated conditions at your plant for which the valves are required to operate and compare these plant conditio.s to the conditions in the test program. Describe the plant features assumed in the event evaluations used to scope the test program and compare them to plant features at your plant. For example, describe high level trips to prevent water from entering the steam
-lines under high pressure operating conditions as assumed in the test event and compare them to trips used at your plant.
RESPONSE TO QUESTION 4 The purpose of the SRV test program was to demonstrate that the SRV will open and reclose under all expected flow conditions. The expected valve operating conditions were determined through the use o' analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Revision 2. Single failures were applied to these analyses so that the dynamic forces on the safety and relief valves'would be maximized. Test pressures were the highest predicted by conventional safety analysis procedures. The BWR Owners Group, in their enclosure to the September 17, 1980 letter from D. B.
Waters to R. H. Vollmer, identified 13 events which may result in liquid or two-phase SRV inlet flow that would maximize the dynamic forces on the safety relief valve. These events were identified by evaluating the initial events described in Regulatory Guide 1.70, Revision 2, with and without the additional conservatism of a single active component failure or operator error postulated in the event sequence. It was concluded from this evaluation that the alternate shutdown cooling mode is the only expected event which will result in liquid at the valve inlet. Consequently, this was the event simulated in the SRV test program. This conclusion and the test results appifcable to GGNS are discussed below.
The SRV inlet fluid conditions tested in the BWR Owners Group SRV test program, as documented in NEDE-24988-P, are 15* to 50*F subcooled liquid at 20 psig to 250 psig. These fluid conditions envelope the conditions expected to occur at GGNS in the alternate shutdown cooling mode of operation.
i J0P14ATTC85032201 - 4
Attrchm:nt 1 to AECM-85/009)
The BWR Owners Group identified 13 events by evaluating the initiating events described in Regulatory Guide 1.70, Revision 2, with the additional conservatism of a single active component failure or operator error postulated in the events sequence. These events and the plant-specific features that mitigate these events, are summarized in Table 1 of this attachment. Of these 13 events, only 10 are applicable to GGNS because of its design and specific plant configuration. Three events, namely events #3, #6 and #11 are NOT applicable to GGNS because Grand Gulf does not have a High Pressure Core Injection (HPCI) system nor a Reactor Core Isolation Cooling (RCIC) head spray.
For the 10 remaining events the GGNS specific features, such as trip logic, power supplies, instrument line configuration, alarms and operator actions have been compared to the base cate analysis presented in the BWR Owners Group submittal of September 17, 1980. The comparison has demonstrated that the base case analysis is applicable to GGNS because the base case analysis includes plant features which are already present in the GGNS design. For these events.
Table 1 shows what GGNS specific features are included in the base case analyses presented in the BWR Owners Group submittal of September 17, 1980. It is seen from Table 1, that most plant features assumed in the event evaluation are also existing features in GGNS. All features included in this base case analysis are similar to plant features in the GGNS design or do not have a negative effect upon this comparison. Furthermore, the time available for operator action is expected to be longer at GGNS than in the base case analysis for each case where operator action is required.
Event #7, the alternate shutdown cooling mode of operation, is the only expected event which will result in liquid or two-phase fluid at the SRV inlet. Consequently, this event was simulated in the BWR SRV test' program.
At GGNS this event involves flow of subcooled water (approximately 35'F subcooled) at a pressure of approximately 135 psig. The test conditions clearly envelope these plant conditions.
As discussed above, the BWR Owners Group evaluated transients including single active failures that would maximize the dynamic forces on the safety.
relief valves. As a result of this evaluation, the alternate shutdown cooling mode is the only expected event involving liquid or two-phase flow.
Consequently this event was tested in the BWR SRV test program. The fluid conditions and flow conditions tested in the BWR Owners Group test program conservatively envelope the GGNS plant-specific fluid conditions expected for the alternate shutdown cooling mode of operation.
NRC OUESTION 5 The valves are likely to be extensively cycled in a controlled depressurization mode in a plant-specific application. Was this mode simulated in the test program? What is the effect of this valve cycling on valve performance and probability of the valve to fail open or to fail closed?
J0P14ATTC85032201 - 5
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-RESPONSE'TO QUESTION 5-u . .
!The BWR safety / relief valve operability test program was designed to simulate
>the alternate shutdown cooling mode, which is thelonly expected. liquid
-discharge event.for GGNS.- The sequence of events leading to the alternate l shutdown cooling mode is given below..
Following normal reactor shutdown, the reactor operator.depressurizes the reactor vessel by opening the turbine bypassL valves and removing heat through the main condenser. <If the main condenser-is unavailable, the operator could depressurize the reactor vessel by using the SRVs:to~ discharge steam to the s ! suppression pool.' .If SRV operation is required, the-operator cycles the valves in order.to assure that the cooldown rate is maintained within the technical
' specification limit of 100'F per hour.. When the vessel is depressurized, the operator-initiates normal shutdown cooling by use of the RHR system. If that.
system is unavailable because the valve on the RHR shutdown cooling suction
.line fails ~to open, the operator initiates the alternate shutdown cooling mode.
For alternate shutdown cooling, the operator opens one SRV and initiates either an RHR or core spray-pump utilizing the suppression pool as the suction source., The reactor vessel is filled such that water is allowed to flow into
-the main steam lines and out of the SRV and back to the suppression pool.
. Cooling of the system is provided by use of RHR heat exchanger. As a result,'an alternate cooling mode is maintained.
In' order to assure continuous long term heat. removal, the SRV is kept_open'and no-cycling of the valve is performed. In order,to control the reactor vessel cooldown rate .the operator is instructed to control the flow rate into the
' vessel. Consequently, no cycling of the SRV is required for the alternate.
shutdown: cooling mode, and no cycling of the SRV was performed for the generic
'BWR SRV operability test program.
~ The ability of the GGNS SRV to be extensively cycled for steam discharge conditions has been confirmed during steam' discharge qualification testing of the valve by the valve, vendor. Based on.the qualification testing of the SRV's, the cycling of thervalves<in a< controlled depressurization mode for-steam discharge conditions'ill w not adversely affect valve performance and the
. probability of the valve to fail open or closed is extremely low.
I
, NRC QUESTION 6
. Describe how the values'of valve C 's ,in report NEDE-24988-P will be used at
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' .your plant. .Show that the methodology used in the test program to determine the valve Cy is consistent with your application.
RESPONSE TO QUESTION 6 h
The flow coefficient, C , for the Dikkers safety relief valve utilized L at,GGNS was determined in the generic-SRV test program (NEDE-24988-P). The l average flow coefficient calculated from the test results for the Dikkers SRV is reported in Table 5.2-1 of NEDE-24988-P. This test value has been used by
-Mississippi Power and Light Company.to confirm that the liquid discharge flow I 'J0P14ATTC85032201 - 6 i: ..
I
r Attach: ant I to AECM-85/0099 capacity of the Dikkers SRVs will be suf ficient to remove core decay heat when injecting into the reactor pressure vessel (RPV) in the alternate shutdown cooling mode. An evaluation was performed to determine the number of SRVs required to discharge 7450 gallons per minute of water during the alternate shutdown cooling mode. This evaluation indicates that 3 SRVs are sufficient to perform this operation. Therefore, it is concluded from the C value determined in the SRV test that the GGNS SRVs are capable of performing the alternate shutdown cooling mode of operation.
If it were necessary for the operator to place GGNS in the alternate shutdown cooling mode,'he would be assured that adequate cort cooling was being provided by monitoring the following parameters: RHR or core spray flow rate, reactor vessel pressure and reactor vessel temperature.
The flow coefficient for the Dikkers SRV reported in NEDE-24988-P was determined from the SRV flow rate when the valve inlet was pressurized to approximately 250 psig. The valve flow rate was measured with the supply line flow venturi upstream of the steam chest. The C for the valve was calculated using the nominal measured pressure differential between the valve inlet (steam, chest) and 3 feet downstream of the valve and the corresponding measured flowrate. Furthermore, the test conditions and test configuration envelope the GGNS conditions for the alternate shutdown cooling mode, e.g. , pressure upstream of the valve, fluid temperature, friction losses and liquid flowrate.
Therefore, the reported Cy values are appropriate for application to GGNS.
J0P14ATTC85032201 - 7 i
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Attechm nt 1 to AECM-85/0099 Footnotes to Table 1
- 1. Not applicable because this initiating event can not occur. GGNS does not have HPCI.
- 2. Not applicable. A HPCI level 8 trip is not required to make the test results applicable because GGNS does not have HPCI.
3.- Not applicable because GGNS does not have RCIC head sprays.
- 4. Not applicable because CGNS does not have RCIC. Initiation on High Drywell Pressure. This does not affect the applicability of the test results to the GGNS SRVs, because lack of this feature can not cause liquid flow through the SRVs.
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TEST NO. P.EDIA LOAD LINE CONF ICURATION DATE RE. DARKS aj 4 Steam I 3/3/61 Test Acceptable.
101 Water 1 3/3/81 Test Acceptable. 1 102 1 Steam I 3/3/31 Test Acceptable.
103 Vater I 3/4/B1 Test Acceptable. ,
10L Steam I 3/4/81 Test Acceptable. h,-
105 i 3/4/81 Test Acceptable.
106 Vater ,
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