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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 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 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 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 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 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 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 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 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 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 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 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) 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 ML20211K6061999-08-31031 August 1999 Informs That Plant Has No Candidates to Take 991006 Generic Fundamentals Exam 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 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 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 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 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 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 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 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 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 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 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 ML20216J7101999-09-26026 September 1999 Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator License Examinations 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) 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 ML20211K6061999-08-31031 August 1999 Informs That Plant Has No Candidates to Take 991006 Generic Fundamentals Exam 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 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 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 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) 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 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 ML20207A9901999-02-24024 February 1999 Informs That Util Has No Candidates from GGNS to Nominate for Participation in Planned Gfes,Scheduled for 990407 ML20207A8161999-02-24024 February 1999 Forwards 1998 Annual Operating Rept for Ggns,Unit 1. Listed Attachments Are Encl 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-90-0172, 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-90-0169, 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-90-0174, 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) AECM-90-0124, Forwards Rev 1 to Relief Request I-00018 Correcting Valve Number & Description of One Component.Review & Approval Requested Prior to 9010011990-07-0909 July 1990 Forwards Rev 1 to Relief Request I-00018 Correcting Valve Number & Description of One Component.Review & Approval Requested Prior to 901001 AECM-90-0120, Responds to NRC 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 Operators1990-06-29029 June 1990 Responds to NRC 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 AECM-90-0107, 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 Performed1990-06-22022 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 AECM-90-0112, 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 Flow1990-06-14014 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 AECM-90-0106, Forwards Rev 9 to GGNS-TOP-1A, Operational QA Manual, for Evaluation1990-06-11011 June 1990 Forwards Rev 9 to GGNS-TOP-1A, Operational QA Manual, for Evaluation AECM-90-0110, Forwards List of Directors & Officers of Entergy Operations, Inc.Operation of All Plants Transferred to Entergy on 9006061990-06-0808 June 1990 Forwards List of Directors & Officers of Entergy Operations, Inc.Operation of All Plants Transferred to Entergy on 900606 AECM-90-0097, Forwards Bimonthly Status Repts Re Security Boundary Upgrade Project for Period Ending 9005311990-06-0808 June 1990 Forwards Bimonthly Status Repts Re Security Boundary Upgrade Project for Period Ending 900531 AECM-90-0094, Forwards Updated Svc List to Be Used Re Plant Correspondence.Requests WT Cottle Be Primary Addressee for All Correspondence Concerning Plant1990-06-0707 June 1990 Forwards Updated Svc List to Be Used Re Plant Correspondence.Requests WT Cottle Be Primary Addressee for All Correspondence Concerning Plant AECM-90-0103, Forwards Nonproprietary ANF-90-060(NP), Criticality Safety Analysis for Grand Gulf Fuel Storage Racks W/ANF-1.4 Fuel Assemblies1990-06-0707 June 1990 Forwards Nonproprietary ANF-90-060(NP), Criticality Safety Analysis for Grand Gulf Fuel Storage Racks W/ANF-1.4 Fuel Assemblies 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 AECM-90-0100, Forwards 1989 Annual Financial Repts for Sys Energy Resources,Inc & South Mississippi Electric Power Assoc1990-06-0606 June 1990 Forwards 1989 Annual Financial Repts for Sys Energy Resources,Inc & South Mississippi Electric Power Assoc AECM-90-0104, 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 9006061990-06-0606 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 AECM-90-0102, 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 Review1990-05-31031 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 AECM-90-0091, Forwards Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Jan-Mar 19901990-05-24024 May 1990 Forwards Degraded Core Accident Hydrogen Control Program, Quarterly Status Rept for Jan-Mar 1990 AECM-90-0093, 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 Required1990-05-23023 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 AECM-90-0088, Forwards Final Response to Generic Ltr 89-04, Guidance on Developing Acceptable Inservice Testing Programs & Rev 4 to Pump & Valve Inservice Testing Program1990-05-18018 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 AECM-90-0089, 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 Plant1990-05-17017 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 AECM-90-0082, 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 Monitoring1990-05-0909 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 AECM-90-0078, Forwards Rev 4 to Fire Hazards Analysis. Design Changes Include Installation of Alternate DHR Sys & Access Hatch in Pipe Chase1990-05-0909 May 1990 Forwards Rev 4 to Fire Hazards Analysis. Design Changes Include Installation of Alternate DHR Sys & Access Hatch in Pipe Chase AECM-90-0072, Forwards Response to Recommendations Re Areas of Concern Noted in NRC SER & 900316 Request for Addl Info Re Design Criteria for Cable Tray Supports in Turbine Bldg1990-05-0909 May 1990 Forwards Response to Recommendations Re Areas of Concern Noted in NRC SER & 900316 Request for Addl Info Re Design Criteria for Cable Tray Supports in Turbine Bldg AECM-90-0086, Requests Extension of 90 Days to Provide Addl Time for Securities & Exchange Commission Review Re Implementation of Amend 65 to License NPF-291990-05-0404 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 AECM-90-0067, 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 Specs1990-05-0404 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 AECM-90-0081, Responds to NRC 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 Contamination1990-04-30030 April 1990 Responds to NRC 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 AECM-90-0075, 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,Per1990-04-30030 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 AECM-90-0041, Forwards Response to NRC 900329 Request for Addl Info Re Boraflex Gap Analysis & Surveillance Program.Boraflex Gap Measurement Results Will Be Compared to Acceptance Criteria Described in AECM-90/0068 for Both Cycle 4 & 5 Analysis1990-04-30030 April 1990 Forwards Response to NRC 900329 Request for Addl Info Re Boraflex Gap Analysis & Surveillance Program.Boraflex Gap Measurement Results Will Be Compared to Acceptance Criteria Described in AECM-90/0068 for Both Cycle 4 & 5 Analysis AECM-90-0079, Advises That Util Will Not Be Able to Provide Complete Supplemental Summary Rept on Dcrdr by 900430,as Indicated in Util .Supplemental Rept Will Be Submitted by 9009301990-04-30030 April 1990 Advises That Util Will Not Be Able to Provide Complete Supplemental Summary Rept on Dcrdr by 900430,as Indicated in Util .Supplemental Rept Will Be Submitted by 900930 AECM-90-0057, Responds to Generic Ltr 89-15, Emergency Response Data Sys. Util Volunteers to Participate in Emergency Response Data Sys1990-04-30030 April 1990 Responds to Generic Ltr 89-15, Emergency Response Data Sys. Util Volunteers to Participate in Emergency Response Data Sys AECM-90-0063, Provides Addl Info Re Plant IGSCC Programs in Response to Generic Ltr 88-01.Table 3C Noted in 890714 Response to Ltr Modified to Reflect Correct Schedule,Per Inservice Insp Program.Revised Table 3C,reflecting Insp of Weld N6-A Encl1990-04-23023 April 1990 Provides Addl Info Re Plant IGSCC Programs in Response to Generic Ltr 88-01.Table 3C Noted in 890714 Response to Ltr Modified to Reflect Correct Schedule,Per Inservice Insp Program.Revised Table 3C,reflecting Insp of Weld N6-A Encl AECM-90-0071, Forwards Rev 0 to Temporary Procedure 04-1-01-P41-1-Temp-10, Contingency Long-Term Cooling for Standby Svc Water Basin a1990-04-16016 April 1990 Forwards Rev 0 to Temporary Procedure 04-1-01-P41-1-Temp-10, Contingency Long-Term Cooling for Standby Svc Water Basin a 1990-09-08
[Table view] |
Text
r rC MISSISSIPPI POWER & LIGHT COMPANY -
Helping Build Mississippi
~'
P. O. B O X 16 4 0 J A C K S O N. MISSISSIPPI 3_E 2 0 5
/
.c PROOVCTION OEPARTWENY August 10, 1979 Of fice of Inspection & Enforcement U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 ATTENTION:
Mr. J.
P. O'Reilly, Director Centlemen:
S UB.1ECT : Grand Gulf Nuclear Station File 0272/0260/0275/M-001.0/L-860.0 IE Bulletin 79-02 Revision No. 1 Evaluation AECM-79/89 on July 6,1979, Mississippi Power & Light Company submitted its response to IE Bulletin 79-02 Revision No. 1 entitled " Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts."
At the request of Mr. William Ang of your staff, we have revised our response to clarify several items.
Our revised response is attached.
Also, at the request of Mr. Ang, we will be forwarding Reference 3 of the attached response to you in the near future.
Yours truly,
-b c
L. F. Dale Nuclear Project Manager RTE:dwe Attachment cc:
Mr. N.
L. Stampley Mr. R. B. McGehee Mr. T.
B. Conner Mr. John G.
Davis, Director Division of Inspection & Enforcement U. S. Nuclear Regulatory Commission Washington, D. C.
20555 n
,.5 0d02 'i0 N90918nlQ Member Middle South Utilities System
? r i. A Y Wig
f' This is in response to NRC Bulletin 79-02 (Revision No. 1), dated June 21, 1979, requiring all licensees and permit holders for nuclear power plants to review the design and installation procedure used for Seismic Category I pipe suppott base plates which are anchored by expansion bolts.
The following types of supports have been reviewed for the Grand Gulf Nu-clear Power Station (GG).
a.
Pipe Anchors b.
Main Steam Relief Valve (MSRV) Supports c.
Seismic Category I Pipe Hangers d.
Supports for piping systems 2-inch in diameter or less The basic document governing the design and installation of expansion an-chors on GG is the Specification 9645-C-103.1.
of expansion anchors is governed by Specification 3645-M-205.0.
Procurement Both documents were prepared by Bechtel Power Corporation.
The concepts which are presented in this response will be applied to all future designs and revisions for Units 1 and 2.
Seismic Category I systems are defined in Regulatory Guide 1.29, Revision 2, with exceptions noted in Grand Gulf's FSAR Appendix 3A.
OUESTION #1 Verify that pipe support base plate flexibility was accounted for in the calcule: ion of anchor bolt loads.
In lieu of supporting analysis justifying the assumption of rigidity, the base plates should be considered flexible if the unstiffened distance between the member welded to the plate and the edge It is of the base plate is greater than twice the thickness of the plate.
recognized that this criteria must be justified and the justification sub-mitted as part of the response to the Bulletin.
If the base plate is LSLil determined to be flexible, then recalculate the bolt loads using an appropriate analysis.
If possible, this is to be done prior to testing of anchor bolts. These calculated bolt loads are referred to hereafter as the bolt design loads. A description of the analytical model used to verify that pipe sup;- rt base plate flexibility is accounted for in the calculation of anchor bolt loads is to be submitted with your response to the Bulletin.
It has been noted that the schedule for analytical work on base plate fley!.bility for some facilities extends boyond the Bulletin reporting time frame of July 6, 1979.
For those facilities for which an anchor bolt testing program is required (i.e., sufficient QC documentation does not exist), the anchor bolt.esting program should not be delayed.
RESPONSE
All pipe anchors and MSRV supports with base plates using expancion anchors were re-analyzed to account for plate flexibility and bolt stiffness. The effects of shear-tension interaction, minimum edge distance and bolt spacing were included in the original designs and were also included in the re-analysis of the base plates.
One of the following methods used during the re-analysis to solve for tensile bolt loads:
1.
An empirical method based on the results of finite element analysis using ANSYS computer programs for base plates with eight (8) or less bolts. (ATTACHMENT 1) 2.
Finite Element Analysis Method using the ASSYS computer program.
(ATTACHMENT 2) 3.
An empirical method based on the results of finite element analysis using ANSYS computer programs for base plates with twelve (12) bolts.
(ATTACHMENT 3).
ry o 9 3 <3 sJdD A* t M Other acceptable methods of structural analysis with conservative 4.
assumptions.
Shear bolt loads were computed by conventional m,thods.
One of the follow-ing shear-tension interaction equations was used during the re-analysis:
T
[S Where:
T = Design Tension Load 4[ 1.0
+ 1 T
k 5^
T = Allowable Tension Load
~
4 A
OR S = Design Shear Load 3
T 5l3
( S et S = Allowable Shear Load T
4 ( g--
- 1. 0 A
s 4
Both of these equations are more conservative than the circular interaction equation generally used for this applicatien. Any consideration of shear force is conservative due to friction between the base plate and the concrete surface.
During the design of Seismic Category I pipe hangers, effects of plate flexibility, shear-tension interaction, bolt spacing and minimum edge dis-tances were considered to assure a conservative initial design. Re-analysis was performed for selected plates using methods No. 1 as outlined above to assure that the original design approach was conservative.
Listed below are the total number of hangers checked for each bolt pattern and the total number of hangers which employ those patterns.
POPU LATION NUMBER OF HANGER BOLT PATTERN (TOTAL NUMBER OF SKETCHES ANALYZED HANGERS) 4 - bolt 304 3
6 - bolt 15 2
8 - bolt 5
3 3b51.43 In each case the original design was found to be conservative.
two inch and under piping, a For design of base plates used to support chart analysis was used.
This analysis method is contained in Bechtel Power Corporation Document No. M-18, "Of fice and Field Engineering User's 17.
Manual for Routing and Supporting Two Inch and Under Piping", Rev.
This chart analysis has been shown to be very conservative.
It should be noted that all methods of analysis conservatively ignored the ef f ects of redistribution of tensile bolt loads.
This phenomenon occurs as stiffness.
the bolt design loads increase which decreases the individual bolt Prying action was considered negligible in the re-analysis for the following reasons:
1.
Due te the low stiffness of expansion bolts, it is unlikely that prying action would exist.
2.
If prying action does occur, it typically exists at the plate corners which results in the following:
Where the anchorage system capacity is governed by the a.
concrete shear cone, the prying action would result in an application of an external compressive load on the cone and would not therefore affect the anchorage capacity.
b.
Where the bolt pull out determines the anchorage capacity, the additional load carried by the bolt due to the prying action will be self-limiting since the bolt stiffness de-creases with the increasing load.
At higher loads the bolt extension will be such that the corners of the base plate will lift of f and the prying action will be relieved.
This phenomenon has been found to occur when the bolt stiffness in the Finite Element Analysis were varied from a high to a low cc r sJs)D.y a n.c. t Q to correspond to typical values of initial bolt
- value, stif fness and values of stif f ness at a point beyond allow-able design loads.
c.
A point load will result in an infinitely high compressive stress in the concrete at the point of prying.
In this case, the high stress probably would lead to local concrete crushing which will relieve the prying action.
QUESTION #2 2.
Verify that the concrete expansion anchor bolts have the following minimum factor of safety between the bolt design load and the bolt ultimate capacity determined from static load tests (e.g. anchor bolt manufacturer's) which simulate the actual conditions of installation (i.e., type of concrete and its strength properties):
Four - For wedge and sleeve type anchor bolts, a.
b.
Five - For shell type anchor bolts.
The bolt ultimate capacity should account for the effects of shear-tension interaction, minimum edge distance and proper bolt spacing.
If the minimum factor of safety of four for wedge type anchor bolts and five for shell type anchors can not be shown then justification must be provided.
RESPONSE
Grand Gulf uses only the wedge type anchor bolts for Category I piping systems.
For all cases, a safety factor of four was applied to the ultimate capacity of the bolt.
These ultimate capacities were obtained f rom static load tests as furnished by the manufacturer.
The factor of safety of four is higher than the factor of safety considered acceptable as specified in Reference 1.
WSIM
-s-
As stated in the responses to Questio:i #1, the ef fects of shear-tension interaction, minimum edge distance and proper bolt spacing were accounted for.
QUESTION #3 3.
Describe the design requirements if applicable for anchor bolts to withstand cyclic loads (e.g. seismic loads and high cycle operating loads).
RESP 0SSE In the original design of the piping systems Bechtel considered deadweight, thermal stresses, seismic loads, and dynamic loads (including steam hammer in the main steam systems) in the generation of the static equivalent pipe support design loads. To the extent that these loads include cyclic considera-tions, these effects would be included in the design of the hangers, base plates and anchorages.
The safety factors used for concrete expansion anchors, installed on supports for safety related piping systems, were not increased for loads which are cyclic in nature. The use of the same safety factor for cyclic and stacic loads is based or the FFTF Tests (Reference 2).
The test results indicate:
1.
The expansion anchors successfully withstood two million cycles of long term fatigue loading at a maximum intensity of 0.20 of the static ultimate capacity. When the maximum load intensity was steadily increased beyond the aforementioned value and cycled for 2,000 times at each load step, the observed failure load was about the same as the static ultimate capacity.
2.
The dynamic load capacity of the expansion anchors, under simulated seismic loading, was about the same as their correspond-ing static ultimate capacities.
yx-
+-
s> J D J.: H QUESTION #4 Verify f rom existing QC documentation that design requirements have been met for each anchor bolt in the following areas:
(a)
Cyclic loads have been considered (e.g. anchor bolt preload is equal to or greater than bolt design load).
In the case of the shell type, assure that it is not in contact with the back of the support plate prior to preload testing.
(b)
Specified design size and type is correctly installed (e.g.
proper embedment depth).
In sufficient documentation does not exist, then initiate a testing program that will assure that minimum design requirements have been met with respect to sub-items (a) and (b) above.
A sampling technique is acceptable.
On acceptable technique is to randomly select and test one anchor bolt in each base plate (1. e. some supports may have more than one base plate).
The test should provide verification of sub-items (a) and (b) above.
If the test f ails, all other bolt s on that base plate should be similarly tested.
In any event, the test program should assure that each Seismic Category 1 system will perform its intended function.
The preferred test method to demonstrate that bolt preload has been accomplished is using a direct pull (tensile test) equal to or greater than design load.
Recognizing this method may be difficult due to accessi-an alternative test method such as torque testing bility in some arc-c may be used.
If torque testing is used it must be shown and substantiated that a correlation between torque and tension exists.
If manufacturer's data for the specific bolt used is not available, or is not used, then site specific data must be de loped by qualification tests.
o, J v.'J ;'."]
- c:.
1 a
Bolt test values of one-fourth (wedge type) or one-fif th (shell type) of bolt ultimate capacity may be used in lieu of individually calculated bolt design loads where the test value can be shown to be conservative.
The purpose of Bulletin 79-02 and this revision is to assure the In all cases operability of each seismic Category I piping system.
an evaluation to confirm system operability must be performed.
If a base plate or anchor bolt failure rate is identified at one unit of multi-unit site which threatens operability of safety r4 lated piping systems of that unit, continued operation of the remaining units at that site must be immediately evaluated and reported to the NRC.
The evaluation must consider the generic applicability of the identified failures.
can be Appendix A describes two sampling methods for testing that used.
Other sampling methods may be used but must be justified.
These options may be selected on a system by system basis.
Justification for omitting certain bolts from sample testing which are in high radiation areas during an outage must be based on other testing or analysis which substantiates operability of the af fected system.
Bolts which are f ound during the testing program not to be preloaded to a load equal to or greater than bolt design load must be properly preloaded or it must be shown that the lack of preloading is not detri-mental to cyclic loading capacility.
If it can be established that a tension load on any of the bolts does not exist for all loading cases then no preload or testing of the bolts is required.
If anchor bolt testing is done prior to completion of the analytical work on base plate flexibility, the bolt testing must be performed to at least the original calculated bolt Icad.
For testing purposes Od8],f}.@
factors may be used to conservatively estimate the potential increase in the calculated bolt load due to base plate flexibility. After completion of the analytical work on the base plates the conserva-tism of these factors must be verified.
raised from the For base plate supports using expansion anchors, but supporting surf ace with grout placed under the base plate, for testing purposes it must be verified that leveling nuts were not used.
If leveling nuts were used, then they must be backed off such that they in contact with the base plate before applying tension or are not torque testing.
Bulletin No. 79-02 requires verification by inspection that bolts are Parameters properly installed and are of the specified size and type.
which should be included are embedment depth, threat engagement, plate bolt hole size, bolt spacing, edge distance to the side of a concrete member and full expansion of the shell for shell type anchor bolts.
If piping systems 2 -inch in diameter or less were computer analyzed If a chart then they must be treated the same as the larger piping.
analysis method was used and this method can be shown to highly conserva-t iv e, then the proper installation of the base plate and anchor bolts should be verified by a sampling inspection.
The parameters inspected should include those described in the preceding paragraph.
If small diameter piping is not inspected: then justification of system oper-ability must be provided.
[Ibb.1lO
/JPENDIX A SAMPLING METHODS Item 4 of this Bulletin states that for anchor bolt testing purposes a sampling program is acceptable. Two sampling methods are discussed below, but other methods may be used if justified.
Test one bolt on each plate as originally recommended a.
in Bulletin No. 79-02.
If the tesc fails, all other bolts on that base plate should be similarly tested.
A high failure rate should be the basis for increased testing, b.
Randomly select and test a statistical sample of the bolts to provide a 95 percent confidence level that less than 5 percent defective anchors are installed in any one seismic Category I system. The sampling program should be done on a system by system basis.
RESPONSE
Cyclic loads were considered in the design of the base plate as de-a.
scribed in the response to Question #3.
The FFTF tests and other tests have demonstrated the capability of expansion anchors to withstand dy-namic and fatigue type cyclic loads.
In Reference 3 the preload in the dynamic tests was varied from " finger tight" to a value approximately equal to one quarter of the static ultimate capacity.
Furthermore, due to creep relaxation of expansion bolt anchorages after installation, it is not possible to predict the preload in expansion anchors at any given time.
In conclusion, if the initial installation torque accomplishes the purpose of setting the wedge, then the ultimate capacity of the bolt is not affected by the amount of preload present in the bolt at the time of cyclic loading.
?N50.$.530 b.
The design size and type of expansion anchor is specified on the design drawing for each pipe support. Proper embedment depth is controlled by Specification 9645-C-103.1.
In reference to sub-item a. above, a field testing procedure for installation torque is used.
This procedure involves testing a certain number of anchors with calibrated torque wrench and is governed by Section 6.0 of Specifi-cation 9645-C-103.1.
The testing frequency and acceptance criteria are described in Sections 6.2 and 6.3 as follows:
6.2 Testing Frequency Expansion anchors shall be tested at the following frequency:
6.2.1 For each support test at least one anchor, but not less than 10 percent of the total expansion anchors used for the support.
For supports which have less than five expansion anchors and are installed in continuous repetitious patterns, test 10 percent without regard for the number of supperts. Groups selected for 10 percent testing shall not exceed 20 anchors.
Select expansion anchors for test at random so that they are representative of the group in which they are located and of the installation conditions.
6.2.2 Accept all the anchors in the test group if the anchors tested according to the procedures in Paragraph 6.2.1 meet the acceptance criteria given in Paragraph 6.3.
6.2.3 Test an additional two, but not less than 30 percent of the expansion anchors from the same group if those tested according to the procedures in Paragraph 6.2.1 do not meet the acceptance criteria given in Paragraph 6.3.
Eb.$$Ifi_
6.2.4 Accept all remaining expansion anchors in the group as satisfactory if those tested according to the procedures given in Paragraph 6.2.3 neet the acceptance criteria as given in Paragraph 6.3.
6.2.5 Test all remaining expansion anchors in the group if one of those tested according to the procedures in Paragraph 6.2.3 does not meet the ecceptance criteria as given in Paragraph 6.3.
Accept each anchor on an individual basis.
6.3 Acceptance Criteria Accept the tested expansion anchor if during testing the concrete did not break and/or the anchor did not slip excessively (1/4 inch maximum) or become loose, and the minimum torque or tension values of Paragraph 6.0 are met.
The results of the torque testing are documented and maintained at the jobsite.
In reference to sub-item (b) above, design size and type are governed by Section 6.0 of Specification 9645-C-103.1.
This section requires expansion bolt in-stallation to be verified for location (elevation, spacing and edge distance),
anchor type maximum bolt projection and anchor diameter and length. These items are checked with the design drawings.
The anchor length, which is specified on the design drawings, ensures proper embedment depth.
Documentation is kept at the jobsite.
Specification 9645-C-103.1 invokes the quality assurance criteria in accordance with 10 CFR 50, Appendix B.
Under the quality assurance program existing during construction, failure to meet the acceptance criteria would create a condition whereby the work would be either corrected or rejected.
O... 9 t cvs) D.L s E *
REFERENCES:
(1) Proposed Addition to:
Code Recuirements for Nuclear Safety Related Concrete Structures (ACI 349-76), Title No.
75-35.
(2) " Drilled in Expansion Bolts Under Static and Alternating Load", Prepared by Bechtel Power Corporation, San Francisco, California for the U. S. Atomic Energy Commission, Hartford Engineering Development Laboratory, Richland, Washington, BR-5853-C-4, Rev. 1, October 1976.
(3) " Static and Dynamic Loading of 3/8-Inch Concrete Anchors", Prepared by Pacific Gas and Electric Company Department of Engineering Research.
ObbIlie3 ATTACHMENTS:
(1) Model for base plates with eight (8) or less bolts.
(2) ANSYS computer model and example.
(3) Model for base plates with twelve (12) bolts. 33[Ya55'E