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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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MISSISSIPPI POWER & LIGHT COMPANY
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Helping Build Mississippi P. O. B O X 1840. J A C K S O N, MIS SIS SIP PI 3 9 2 0 5 l
NUCLEAR LICEN5f NG & 5AFETY DEPARTMENT
. August 13,1984 I
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Harold R. Denton, Director
Dear Mr. Denton:
SUBJECT:
Grand Gulf Nuclear Station Units I and 2 Docket Nos. 50-416 & 50-417 License No. NPF-13 File: 0260/L-860.0 Request for Exemption in Accordance with 10CFR50.12(a) - (10CFR50, Appendix J)
AECM-84/0415 As recently discussed with your staf f, Mississippi Power and Light Company (MP&L) has identified a need for an operating license condition and associated exemption from certain regulations. To the extent that current design does not comply fully with the latest NRC requirements applied to GGNS, MP&L requests a partial, schedular exemption from 10CFR50, Appendix J, as discussed herein.
Based on your staff's guidance and pursuant to 10CFR50.12(a), MP&L transmits its evaluation of the justification for a partial, schedular exemption to the regulations identified in Attachment I. This attachment provides the information required by 10CFR50.12(a), including a description of the issue addressed in the exemption and the basis upon which MP&L concludes that the exemption may be issued.
Attachment I provides the basis for the conclusion that there wili be no undue risk to the public during the first cycle of operation due to the granting of the res,uested exemption.
In support of evaluations required by 10CFR51.30, MP&L is also providing in Attachment 2, on assessment of the potential environmental impact associated with the exemption request.
r4081SO201 G40813 PDR ADOCK 05000 P
Member Middle South Utilities System p
AECM-84/0415 age 2 -
MISSISSIPPI POWER O L12MT COMPANY As discussed in this attachment, there is no significant increase in environmental impact associated with the exemption over the environmental impact associated "
with no exemption. As a result, MP&L believes that there is ample basis for the NRC staff to conclude that there is no significant environmental impact associated with granting the requested exemption.
Please advise if additional information is required.
Sincerely, m _
L. F. Dale Director, Nuclear Licensing & Safety LFD/sl Attachment cc: Mr. R. B. McGehee (w/o)
Mr. N. S. Reynolds (w/o)
Mr. G. B. Taylor (w/o)
Mr. Richard C. DeYoung (w/a)
Office of Inspection & Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 i Mr. J. P. O'Reilly (w/a)
Regional Administrator U.S. Nuclear Regulatory Commission Region 11 101 Marietta Street, N.W., Suite 2900 Atlanta, GA 30323
AECM-84/0415 Attachment i Pagei JUSTIFICATION FOR TIE REQUIRED EXEMPTION NRC regulations provide for specific exemptions in 10 CFR 50.12(a). The Commission has provided additional guidance regarding this regulation in an order in the Shoreham proceedingl , as modified by Commission action on July 25, 1984.2 In view of the standards in 10 CFR 50.12(a) and the Commission's guidance regarding the issuance of exemptions, we may synthesize the circumstances in which the requested exemptions are warranted as follows: (l) the activities to be conducted are authorized by law, (2) operation with the exemption does not endanger life or property because such would involve no undue risk to the health and safety of the public; (3) the common defense and security are not endangered, and (4) the exemption is in the public interest because, on balance, there is good cause for granting it (e.g., to avoid unnecessary delay and consequent financial hardship) and the public health and safety are adequately protected.
As demonstrated by the discussion herein, and in some instances supported by previous submittals to the Commission or previous safety evaluation reports, or both, referenced below, MP&L is entitled to the requested exemption.
- l. The Requested Exemption and the Activities Which Would Be Allowed Thereunder Are Authorized by Law MP&L is currently authorized to operate GGNS Unit I at low power (5% or less of full power) pursuant to License No. NPF-13, which was issued in accordance with the Atomic Energy Act as amended. GGNS Unit I has completed low power i Order, Long Island Lighting Company (Shoreham Nuclear Power Station, Unit I), CLl-84-8, May 6,1984.
2 Staff Requirements Memorandum MB40725A, July 27,1984.
. . . = . . . - ... , . - _.
AECM-84/0415 Attachment i Page 2 tests and, with the exception of the matters for which exemptions are sought, is essentially ready to perform the surveilliance tests prerequisite to, and to
. commence, power oscension.
if the criterio estabilshed in 50.12(a) are satisfied, os they are in this case, and if
- no other prohibition of law exists to prec!ude the activities which would be authorized by the requested exemption, and there is no such prohibition, then the i
. Commission is authorized by law to grant this exemption request.3
. II. The Requested Exemptions Will Not Endanger Life or Property
- II.A. Description of issue General Design Criterion (GDC) 55 of 10 CFR 50, Appendix A requires that each line that is part of the. reactor coolant pressure boundary and that penetrates primary containment shall be provided with containment isolation valves. The requirements are that twe isolation boundaries be provided which meet one of the GDC combinations consisting of locked closed or automatic isolation volves inside and outside primary containment, unless it con be demonstrated that the
[ containment isolation provisions for a specific class of lines are acceptable on f some other defined basis.
l
' Implementation of this criterion for the GGNS feedwater system recognizes the
~
- . paramount importance of maintaining reactor coolant make-up from all sources j of supply. Therefore, in accordance with the guidance provided in ANSI Standard N271-1976, each portion of the feedwater system that forms part of the reactor -
coolant pressure boundary and penetrates the primary containment has three Isolation volves. The isolation volve inside the containment is a simple check i
3 See: U.S. vs. Allegheny-Ludlum Steel Corp.,406 U.S. 742, 755 (1972).
i 4
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m AECM-84/0415 Attachment i Page 3 volve (QlB21-F010 A&B). The isolation valves outside the containment consist of a testable check valve (GlB21-F032 A&B) located neorest the containment, and a motor operated gate volve (OlB21-F065 A&B). The F032 operator is-Edesigned for testing the free-swing action of the disc. When the solenoids of the testable check volves (OlB21-F032 A&B) are de-energized, air is vented, and spring pressure will close the check valve disc provided normal feedwater pressure is not working against the valve.
During postulated transients and occidents, it is desirable to maintain feedwater system availability for reactor coolant make-up; and for this reason, the external volves do not automatically isolate upon a signal from the protection system.
However, these valves are capable of being remotely closed from the control room to provide leakage protection upon operator judgement that continued make-up from the feedwater source is unnecessary. Should a break occur in the feedwater line, the check volves prevent significant loss of inventory and offer immediate isolation. There is a recognized trade off between minimizing postulated leakage paths and retaining makeup capability. On balance, greater safety is assured by maintaining feedwater supply to the reactor. As Indicated in the " Reactor Safety Study Methodology Applications Program: Grand Gulf #1 BWR Power Plant", NUREG/CR-1659/4 of 4, the accident sequence with the highest core melt frequency is a transient initiated event such as loss of feedwater. Therefore, without further evaluation, it cleorly is not oppropriate to automatically isolate feedwater.
For GGNS, odditional design features were provided to prevent radiological leakoge from the feedwater system isolation valves by use of a water seal. The water seal is provided by the Feedwater Leakage Control System (FWLCS). l Following a LOCA when feedwater is no longer required or available, the operator Initiates the FWLCS by starting or verifying the RHR jockey pumps are running and repositioning certain motor operated valves to provide water from the Jockey pumps to fill the feedwater lines. As a result of this design feature
]
+ 1 AECM-84/0415 Attachment i Page 4 !
providing a pressurized water seal for 30 days, the acceptance criteria for leakage rate testing of the feedwater valves was established as a hydrostatic test of only the F065 A&B valves.
Subsequent to the NRC approval of the FWLCS and leakage testing requirements for feedwater isolation valves, further dynamic analysis of post-LOCA conditions indicated that a positive seal in the feedwater lines could not be assured for a short period of time immediately following the reactor blowdown. Assuming a LOCA and a loss of feedwater, for a short period of time following reactor blowdown, the remaining feedwater and sensible heat in the piping is sufficient to create steam, and pressurize the piping above the drywell pressure. Following FWLCS Initiation (conservatively assumed 20 minutes after the start of the accident), feedwater penetration repressurization could take up to 60 minutes following the event for some scenarios using conservative assumptions. Follow-ing repressurization by FWLCS, a 30 day dynamic water seal is maintained.
Because of the short period of time that the FWLCS cannot assure a water seal in the feedwater lines following a LOCA, MP&L committed to pneumatically test all of the feedwater containment isolation valves to 10 CFR 50, Appendix J requirements. However, strict interpretation of Appendix J requires that the leakage from the Type C tests for all valves be combined with the leakage from 4
all other Type B & C tests to meet the 0.6 La requirements, where La is the maximum primary reactor containment allowable leakage term. This interpreta-tion is extremely conservative in that the sun. af all of the valve leakages for a single containment penetration must be added together insteac' si : counting for only the leakage from the worst valve. In essence, such an interpretation could penalize a containment isolation design with additional isolation valves over the two barriers required. As a result, addition of the Type C leakage from all six valves and all other combined Type B & C leakage exceeds the 0.6 La requirement. Therefore, MP&L requests a schedular exemption from Appendix J Section Ill.C.3 acceptance criteria for Type B & C tests as applied to the feedwater isolation valves.
g.c AECM-84/0415 Attachment 1 Page 5 1
The requested exemption is schedular to the extent that MP&L will take ,
necessary action to come into literal compliance with subject regulation by the
^
startup following the first refueling outoge.
- 11. 8 Primary Containment, integrated and Local Leakage Testing Consistent with MP&L's commitments to conduct Type C testing of volves associated with the feedwater piping penetrations, pneumatic testing is being conducted on the subject valves. The valves associated with these penetrations ,
, are listed in FSAR Table 6.2-44 under Containment Penetration Nos. 9 and 10.
.For the key volves of interest, the approximate leak rates, based on recent '
testing, are presented below in standard cubic centimeters per minute (SCCM):
1 FW TRAIN A FW TRAIN B F010
F065 2,200 0
- Results not yet available
- Leakage from small diameter valves (3/4 inch) associated with these penetra-tions (per FSAR Table 6.2-44) was also measured during this testing and has been added to current leakage totals from Type 8 and C testing. Reference to Type B i and C total leakage other than B21-F010, 32, and 65 includes these small diameter valves. The current total value of all other Type B & C tests is
, approximately 21,000 SCCM. This value is reduced over that reported to the NRC previously in MP&L's December 20,1983 letter (AECM-83/0774) as a result
, of isolation volve maintenance and retesting, including additiono! penetrations requiring pneumatic testing.
For GGNS the maximum allowable leakoge term (La), as defined in Appendix J, is approximately 143,000 SCCM. Based on this value the acceptance criteria established in Appendix J are presented below (approximate values): ,
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AECM-84/0415 Attachment l Page 6 Maximum Allowable - La 143,000 SCCM Type A - ' O.75 La 108,000 SCCM Type B and C ' O.6 La 86,000 SCCM The latest containment integrated leak rate testing (ILRT) resulted in a leak rate of approximately 47,000 SCCM. This value includes total leakage from certain non-feedwater valves which were originally hydrostatically tested, but were later pneumatically tested.
Based upon available measured leak rate data, the following criteria have been established for leakage from penetration 9 and 10:
- 1. Leakage from the single valve in either penetration with the highest leak rate PLUS the lowest (check volve) leckage from the other penetration will be less than 0.7 La. Therefore, given a limiting single failure of the feedwater isolation valves and the additional leakage from the other penetration and the ILRT test, the total leakage will be established by test to be less than La.
- 2. Excluding the valve in each penetration exhibiting the highest leak rate, the total leakage from the remaining four valves PLUS the other Type B/C totals will be less than 0.6 La. Therefore, in each penetration there will be a minimum of two isolation barriers which exhibit accept-cble leakage characteristics even when tested pneumatic-ally.
- 3. In accordance with Section Ill.A.l(d) of Appendix J, feedwater penetration leakage should be added to the Type A testing results since these penetrations were isolated (B21-F065A and B closed) at the time of the containment ILRT. The penetration leakage based upon the lowest check volve leakage for both penetrations PLUS the previous Type A leak rate value will be less than 0./5 La.
While strict compliance with Appendix J requirements for Type C testing is not
= achieved, the extensive local and integrated containment leakage testing will establish (1) credible containment isolation barriers, given the worst case single
AECM-84/0415 Attachment l Pcge 7 failure and (2) favorable containment integrated leakage supported by accident analyses, ll.C Appropriate Criteria for Feedwater Line Leakage The intent of Appendix J containment leakage testing requirements assures that post-accident leakage will not exceed that assumed in radiological dose calcula-tions which demonstrate compliance with the limits of 10 CFR Part 100 and General Design Criterion 19 of 10 CFR 50, Appendix A. For this purpose it is necessary that the containment leakage not exceed Lo, or approximately 143,000 SCCM at GGNS through the containment walls and all penetrations. To assure a conservative approach, Appendix J establishes margin, with respect to Lo, for both Type A tests and Type B and C tests. Furthermore, the NRC Staff has interpreted Type C tests to include the leakage from "all volves," even those in a series path for containment penetration. A recent industry Standard, ANS 56.8-1981, Section 6.6.2, advocates a more realistic approach which utilizes the maximum leakage from a single barrier in a series path. Such a position assures that, even with a single failure of a valve to isolate, the leakage would not result in a condition that exceeds the basic safety limit of La. MP&L requests that this position be allowed on an interim basis for determining the acceptable leakage rate for the feedwater isolation valves.
For the feedwater isolation valves described above, appropriate limits on leakage will be met such that La is not exceeded, even given the most limiting single failure. This would imply that the Containment Integrated Leakage Rote Test
'(ILRT) which demon:,trates compliance with Type A test requirements plus the leakage from the feedwater isolation valves in each nf the two penetrations, given the limiting single failure, must not exceed La at GGNS Additionally, the available marg *n in the GGNS ILRT must be reduced to account for the feedwater check volve leakage since these lines were not initially included in the ILRT per Appendix J, Section Ill.A.l.(d). As described in Section ll.B obove, this criteria has the effect of limiting the combined maximum leakage of the two
AECM-84/0415 Attachment l Page 8 feedwater penetrations to about 96,000 SCCM. (This leakage is based upon subtracting the ILRT leak rate of approximately 47,000 SCCM from La of about 143,000 SCCM). This leakage rate would exceed the limits of Appendix J for Type B and C tests since the combined leakage would exceed 0.6 La criteria.
Limiting the combined maximum leakage (as discussed above) from the two feedwater penetrations to 96,000 SCCM, while exceeding 0.6 La criteria of Appendix J, still Incorporates substantial margin and conservatism to assure that
. the basic safety limits of 10 CFR Part 100 and GDC 19 are met. The leakage rate from all other Type B and C leak tests are conservatively based upon adding leckage from all valves, even those in series. After the initial 10 minute period during which no operator action is assumed, additional conservatism exists due to the dramatic effect on reducing leakage from the feedwater penetrations that would result from the operator closing both F065 va!ves. In this case, assuming the most limiting single failure of the feedwater check volves, the maximum leakage from both penetrations COMBINED with other Type B and C leakage is
~
well below 0.6 La.
II.D Justificotton for Proposed Exemption Ample justification for the requested exemption exists due to the low probability of an event which could lead to significant radiological leakage, the design of the feedwater isolation system including the FWLCS, and the conservative applica-tion of the leakage test results which still is less than the containment leakage assumed in the accident analysis.
The probability of plant conditions which would lead to the potential for significant leakage through the feedwater isolation valves is small. As indicated
, in RSSMAP, 90% of the overall core melt frequency results from dominant accident sequences which are predominately transient events. Only one of the dominant accident sequences is a small break LOCA, which would not result in long term containment pressures as challenging as those used in leakage rate
AECM-84/0415 Attachment i Page 9 testing nor is it likely to provide the conservative radiological source term used for accident analysis. As shown in NEDO-24708, realistic small break scenarios with or without operator action show no fuel failure. Additionally, with feedwater available following a LOCA, a potential leakage path would not be available ur postulated. The probability of a large break LOCA resulting in substantial fuel damage with a loss of fee.dwater and failure of the feedwater isolation valve in 2 manner which would result in the maximum allowable leakage is extremely small. The requested exemption does not increase the probability of an event which could lead to excessive leakage nor does it increase the consequences of such an event since the exemption would still maintair total containment leakage less than that assumed in the accident analysis.
The requested exemption would still maintain the containment isolation bounda-ries which are required by GDC 55 and are in accordance with ANSI Standard N271-1976. As indicated in Section ll.B, the leakage from the feedwater penetrations with the limiting single leak rate plus the current Type B&C totals is only 36% higher then the Appendix J allowable of .6 La. This exceedance -
would only exist for a short period of time clnce the F065 valves would be closed by operator action, as discussed later in this section. If application of the worst single valve leakage per penetration were allowed for all Type C tests, the tested leakage could be near the .6 La. In any event, with the use of the proposed criteria for the feedwater isolation valves, the tested leakage would not exceed La even if one summed the feedwater penetration leakages (limiting check valve failure in one penetration PLUS lowest check volve leakage in other penetration) with the Type A leakage.
Additionally, pneumatic testing of all feedwater isolation valves assumes no credit for the FWLCS which provides on effective long term (greater than 30 days) water seal offer the first hour of an accident. Therefore, there is only a small period of time where such leakage paths could exist using conservative assumptions. Furthermore, considerat!on of various aspects of the feedwater
A AECM-84/0415 Attachment !
Page 10 piping system's design and rellobility in maintaining a water seal between the main condenser and the feedwater penetrations provides additional assurance that containment integrity is maintained consistent with the accident analyses.
Realistically, even with the limiting check valve failure, one of the valves on each line with the best leakage chorocteristics would be available such that the accident leakage would be less than that which was used to meet the acceptance criteria. As shown in Section 11.B, the F065 A&B volves have extremely small tested leakages. The time period when a significant leakage path may potentially exist and exceed the Appendix J criteria is less than 10 minutes.
Operating procedures will be revised to instruct the operator to shut the F065A
& B volves following a LOCA if feedwater is not avaliable. As stated in Chapter 6 of the CGNS FS AR, such post-LOCA manual actions are assumed not to occur within 10 minutes even though the F065 valves con be remotely closed from the control room.
Alternatively, assuming the single failure removes the capability to close both F065 volves, then penetration leakage is controlled by the most leaktight check valve in each path. Criterion 3 of Section 11.8 would require that the containment's integrated leakage (ILRT) COMBINED with the lowest check valve
, leakage for both penetrations be less than 0.75 La.
In addition, as discussed earlier, appropriate corrective actions will be accomp-lished to achieve literal comp!'ance with the subject regulations by first refueling outage. The likelihood of the occurrence of an accident resulting in significant fuel domange is very low in the period during which the exemption is being sought.
1 In sommary, the justification provided above adequately demonstrates that the public health and safety would not be jeopa.-dized by approval of this schedular exemotion request for the first cycle of GGNS operation.
g -
q l
- l AECM-84/0415 Attachment i Pagell 111. The Requested Exemptions Will Not Endanger the Common Defense and Security The common defense and security are not implicated in this exemption request.
Only the potential impact on public health and safety is at issue.
IV. The Requested Exemption is in the Public interest The requested exemption is in the public interest in that any delay in commence-ment of the power ascension program would cause a day-for-day delay in the attainment of commercial operation and since, as shown above, the health and safety of the public will be adequately protected.
Grand Gulf Unit I is physically complete in all essential respects and is ready for power ascension to full power. Upon sottsfactory completion of the power ascension program in acwdonce with the license and technical specifications, the facility will be placed in commercial operation. The requested exemption discussed in Section ll above is schedular. In this instance, the delay associated with modifying the FWLCS or the feedwater check valves now rather than at the first refueling outage ranges from one month to several months. Modification of the FWLCS would require design changes and modifications, including procure-mer.: of certain safety related equipment. Since the conceptual design has yet to be selected, this option would likely require several months. Modification of the feedwater check valves could range from replacement of the disks to replacement of the valves. Procurement of these components would require at least one month.
In any case, a corresponding delay in commerclol operation of Grand Gulf Unit I would be occasioned by delay at this stage. Middle South Energy Inc., and South Mississippi Electric Power Association own undivided ownership interests of 90%
AECM-84/0415 Attachment i Page 12 and 10%, respectively, in Grand Gulf Nuclear Station Unit 1. Any delay in the commercial operation of Grand Gulf Nuclear Station Unit I would cause the cost of the unit to increase at the rate of more than $20 million per month. Under standard ratemaking practices these costs would eventually have to be borne by ,
ratepayers of the offected utilities. This substantial financial impact of a delay in commercial operation on the owners of ';rond Gulf Nuclear Station Unit I and the customers of the utilities which' will receive the output is not warranted inasmuch as, as shown above, the public health and safety are adequately j protected.
i i
i
AECM-84/0415 Attachment 2 Pagel ENVIRONMENTAL IMPACT ASSESSMENT
!. PROPOSED EXEMPTION Mississippi Power & Light Company (MP&L) requests a schedular, partial exemption to the acceptance criteria for . local leakrate testing of certain .
containment isolation valves. The subject acceptance criteria are contained in Section Ill.C.3 of 10 CFR 50, Appendix J and pertains to Type B and C leakage testing of containment penetrations and volves.
The requested exemption is restricted to the plant's two feedwater piping containment penetrations, each of which are provided with three (3) isolation barriers in series. The requested exemption would establish criteria in which the leakage rate for a path would be that associated with the isolation barrier (valve) exhibiting the highest leakage rate. The requested exemption also provides a conservative acceptance criteria for the combined measured leakage rate from the remaining isolation barriers (volves) in the feedwater penetration.
The subject regulation, if strictly interpreted, would require the significantly conservative addition of leakage rates from all barriers in each penetration. The requested exemption proposes a more realistic, yet adequately conservative, application of this regulation in proposing the above described method and criteria for combining and accepting barrier leakage for two subject penetro-tions.
Additionally, the exemption would allow consideration of the m irgin in meeting Type A test by slightly exceeding the leakage criteria for Type B & C tests.
However, the overall leakage, Lo, would not exceed that assumed in the radiological accident analyses.
a ..
AECM-84/0415 Attachment 2 Page 2
- 11. ENVIRONMENT IMPACT ASSESSMENT There are no environmental impacts of the requested exemption. The requested exemption establishes criteria by which leakage of isolation barriers in the feedwater piping penetrations are combined and accepted. The subject accept-ance criterion is described in Section Ill.C.3,10 CFR 50, Appendix J.
The total leakage allowed is not in access of that already accounted for in the accident analyses. Each penetration is provided by design with three principal, isolation barriers, all of which have been or will be subjected to local leakrate testing with air or nitrogen in accordance with Appendix J. Given the worst case single failure of one isolation valve in the penetrations, combined with the lowest check volve leakage from the other feedwater penetration and the Type A leakage, the total leakage is less than La. Therefore, by the granting of the requested exemption and the implementation of the proposed interpretation of the cceeptance criterion specified in Section Ill.3.C of Appendix J, the radio-logical consequences of analyzed accidents involving containment leakage are no different from those previously analyzed.
As discussed in Section ll.A of the preceding attachment, MP&L will take necessary actions to come into literal compliance with the subject regulation by the startup following the first refueling outage. In addition, also discussed in Section ll.A, the current analyses indicate only a small period of time (less than one hour) when leakage through the feedwater penetrations would be expected.
Furthermore, consideration of various aspects of the feedwater piping system's design and reliability in maintaining a water seal between the main condenser and the feedwater penetrations provides additional assurance that containment integrity is maintained consistent with the accident analyses. These factors, combined with the low probability of the occurance of an accident resulting in significant fuel damage, MP&L concludes that there is an overall low likelihood that containment Integrity would be challenged in the period during which the exemption is bding sought.
c; -
AECM-84/0415 Attachment 2 1 Page 3 l 1
I No aspects of the requested exemptior..suggest an increase in the probability of a radiological release in excess of that already analyzed or of an event that would lead to on increase in the consequences of analyzed events.
Further the requested exemption does not otherwise significantly affect radio-logical plant effluents, occupational exposure, non-radiological effluents, or any -
other non-radiological consequences.
In conclusion, based on the above discussion, supported by information presented or referenced in Section ll of Attachment I of this submittal, MP&L has determined that the requested exemption, if granted, has no adverse environmental impact.
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