Letter Sequence Other |
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MONTHYEARML20097A3331996-02-0101 February 1996 Forwards Response to NRC 951213 RAI Re Toed 950623 Response to GL 95-03, Circumferential Cracking of SG Tubes Project stage: Other ML20112E7371996-06-0303 June 1996 Submits Rev of Response to GL 95-03, Circumferential Cracking of Steam Generator Tubes Project stage: Other ML20141E8581997-05-15015 May 1997 Discusses Completion of Licensing Action for GL 95-03, Circumferential Cracking of SG Tubes, Project stage: Other 1996-06-03
[Table View] |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217P2061999-10-26026 October 1999 Forwards for First Energy Nuclear Operating Co Insp Rept 50-346/99-17 on 990928-1001.Insp Was Exam of Activities Conducted Under License Re Implementation of Physical Security Program.No Violations Identified ML20217N3851999-10-20020 October 1999 Forwards RAI Re Licensee 990521 Request for License Amend to Allow Irradiated Fuel to Be Stored in Cask Pit at Davis-Besse,Unit 1.Response Requested within 60 Days from Receipt of Ltr ML20217N2321999-10-15015 October 1999 Requests NRC Approval to Use Alternative to Requirements of 10CFR50.55a(f)(4)(ii).Licensee Requests Extension to Specified Schedule for Implementing Updates to IST Program ML20217G9201999-10-14014 October 1999 Discusses Utils Request for Approval of Quality Assurance Program Changes PY-CEI-NRR-2438, Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl1999-10-14014 October 1999 Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl ML20217F8371999-10-0808 October 1999 Forwards Insp Rept 50-346/99-10 on 990802-0913.One Violation Occurred Being Treated as NCV ML20217A5641999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Davis-Besse on 990901.Informs That NRC Plans to Conduct Addl Insps to Address Questions Raised by Issues Re Operator Errors & Failure to Commit to JOG Topical Rept on MOV Verification ML20212L0691999-09-30030 September 1999 Forwards,For Review & Comment,Copy of Preliminary ASP Analysis of Operational Condition Discovered at Unit 1 on 981014,as Reported in LER 346/98-011 ML20216J6701999-09-24024 September 1999 Forwards Post Examination Documentation for Written Operator Initial License Examination Administered at Davis-Besse Nuclear Power Station on 990920.Without Encls ML20212D3501999-09-21021 September 1999 Forward Copy of Final Accident Sequence Precursor Analysis of Operational Event at Plant,Unit 1 on 980624,reported in LER 346/98-006 05000346/LER-1998-001, Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached1999-09-0909 September 1999 Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached ML20216E5961999-09-0707 September 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1,safety Features Actuation Sys Instrumentation & Associated Bases 3/4.3.1 & 3/4.3.2,reactor Protection Sys & Safety Sys Instrumentation ML20211P3001999-09-0707 September 1999 Forwards FEMA Transmitting FEMA Evaluation Rept for 990504 Emergency Preparedness Exercise at Davis-Besse Nuclear Power Plant.No Deficiencies Identified.One Area Requiring C/A & Two Planning Issues Identified ML20211K6681999-08-30030 August 1999 Forwards Copies of Certified Personal Qualification Statement - Licensee (NRC Form 398) for Operator Candidates Listed Below.Without Encls ML20211K6611999-08-30030 August 1999 Forwards Copies of Operator License Renewal Applications for Individuals Listed.Operators Have Successfully Completed Appropriate Operator Requalification Training Program at Dbnps.Without Encls ML20211K0951999-08-30030 August 1999 Forwards Request for Addl Info Re Fire & Seismic Analyses of IPEEE for Davis-Besse Nuclear Power Station,Unit 1. Response Requested within 60 Days ML20211H0201999-08-25025 August 1999 Forwards semi-annual FFD Rept for 990101-0630 for DBNPS, Unit 1,IAW 10CFR26.71(d) ML20211D1171999-08-20020 August 1999 Forwards Insp Rept 50-346/99-09 on 990623-0802.Violations Identified & Being Treated as Noncited Violations ML20211G3911999-08-20020 August 1999 Forwards Update to Estimated Info for Licensing Action Requests Through 010930,re Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates PY-CEI-NRR-2411, Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl1999-08-19019 August 1999 Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl ML20211J9201999-08-13013 August 1999 Urges NRC to Find Funds for Stockpiling Radiation Pills for Residents Living Near Plant ML20211B0161999-08-13013 August 1999 Forwards SE Accepting Evaluation of Second 10-year Interval Inservice Insp Program Request for Relief Numbers RR-A16, RR-A17 & RR-B9 for Plant,Unit 1 ML20210T1061999-08-12012 August 1999 Forwards Preliminary NRC Forms 398 & 396 for Listed Candidates,Per Operator License Exam Scheduled for Week of 990913.Encl Withheld ML20210S6071999-08-11011 August 1999 Provides Final Response to NRC RAI Re GL 98-01, Y2K Readiness of Computer Systems at Npps ML20210P8051999-08-0909 August 1999 Forwards Insp Rept 50-346/99-15 on 990712-16.No Violations Noted.However,Several Deficiencies Were Identified with Implementation of Remp,Which Collectively Indicated Need for Improved Oversight of Program IR 05000346/19980211999-08-0606 August 1999 Refers to NRC Insp Rept 50-346/98-21 Conducted on 980901- 990513 & Forwards Nov.Two Violations Identified Involving Failure to Maintain Design of Valve & Inadequate C/A for Degraded Condition Cited in Encl NOV 05000346/LER-1998-009, Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl1999-08-0606 August 1999 Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl ML20210H6101999-07-30030 July 1999 Informs That Region III Received Rev 21 to Various Portions of Davis-Besse Nuclear Power Station Emergency Plan.Revision Was Submitted Under Provisions of 10CFR50.54(q) in Apr 1999 ML20210H0491999-07-28028 July 1999 Forwards Application for Amend to License NPF-3,revising TS 3/4.7.5.1, Ultimate Heat Sink, to Allow Plant Operation in Modes 1-4 with Water Temp Less than or Equal to 90 F ML20210G5521999-07-28028 July 1999 Provides Addl Response to 980923 OL Licensing Exam Rept 50-346/98-301 Re OL Exam Administered in Aug 1998.Results of Root Cause Investigation & Corrective Actions,Discussed ML20210G3831999-07-27027 July 1999 Forwards Application for Amend to NPF-3,changing TSs 6.4, Training, 6.5.2.8, Audits, 6.10, Record Retention, 6.14, Process Control Program & 6.15, Odcm ML20211P3071999-07-26026 July 1999 Forwards Final Rept for 990504 Biennial Radiological Emergency Preparedness Exercise for David-Besse Power Station.No Deficiencies Identified for Any Jurisdiction During Exercise ML20210G4391999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1, Safety Features Actuation Sys Instrumentation, & Associated Bases 3/4.3.1 & 3/4.3.2, Reactor Protection Sys & Safety Sys Instrumentation ML20210G7151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising TSs 3/4.3.3.1, Radiation Monitoring Instrumentation, 3/4.3.3.2, Instrument - Incore Detectors & 3/4.3.3.9, Instrumentation - Waste Gas Sys Oxygen Monitor ML20210G5151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs for Implementation of 10CFR50,App J,Option B for Type B & C Containment Leakage Rate Testing ML20210G3211999-07-26026 July 1999 Forwards Written OL Exam & Supporting Matl for Exam to Be Administered at DBNPS During Week of 990913.Listed Encls Withheld from Public Disclosure Until After Exam Complete ML20210C4381999-07-20020 July 1999 Forwards Insp Rept 50-346/99-08 on 990513-0622.Unidentified RCS Leak Approached TS Limit of 1 Gallon Per Minute Prior to Recently Completed Maint Outage.Three Violations of NRC Requirements Identified & Being Treated as NCVs ML20209G3681999-07-15015 July 1999 Advises That Info Submitted in & 990519 Affidavit Re Design & Licensing Rept,Davis-Besse,Unit 1 Cask Pit Rack Installation Project,Holtec Intl, HI-981933,marked Proprietary,Will Be Withheld from Public Disclosure ML20207H6401999-07-0909 July 1999 Discusses Closure of TAC MA0540 Re Util Responses to RAI on GL 92-01,rev 1,suppl 1, Rv Structural Integrity. Staff Has Revised Info in Rvid & Releasing It as Rvid Version 2 ML20209D1341999-07-0808 July 1999 Forwards Notice of Withdrawal of Application for Amend to Operating License.Proposed Change Would Have Modified Facility TSs Pertaining to Allowable as-found Pressure Lift Setting Tolerance of Two Pressurizer Code Safety Valves 05000346/LER-1998-012, Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached1999-07-0707 July 1999 Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached ML20209C3981999-07-0101 July 1999 Responds to NRC Re Violations Noted in Insp Rept 50-346/98-21.Corrective Actions:Developed Rev to Boric Acid Control Program & Work Process Guideline on Plant Leakage ML20209B5821999-06-24024 June 1999 Provides Justification for Rev to Completion Date for One of Insp follow-up Items Cited in Insp Rept 50-346/98-03, Designated as Inspector follow-up Item 50-346/97-201-10 ML20196G1251999-06-23023 June 1999 Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20196E5321999-06-17017 June 1999 Forwards Addl Info Re Relief Request RR-A16 to Support NRC Approval of Relief Request ML20195K2751999-06-16016 June 1999 Forwards Safety Evaluation Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20196A6601999-06-16016 June 1999 Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1 ML20195F9071999-06-10010 June 1999 Forwards Application for Amend to NPF-3,changing Tech Specs 3/4.6.4.4, Hydrogen Purge Sys, TS 3/4.6.5.1, Shield Bldg Emergency Ventilation Sys & TS 3/4.7.6.1, Crevs ML20195F8851999-06-0707 June 1999 Withdraws 950929 License Amend Application,Proposing Mod to Allowable as-found Pressure Lift Setting Tolerance of Two Pressurizer Code Safety Valves ML20207G0751999-06-0707 June 1999 Forwards Insp Rept 50-346/99-04 on 990323-0513.Violations Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217N2321999-10-15015 October 1999 Requests NRC Approval to Use Alternative to Requirements of 10CFR50.55a(f)(4)(ii).Licensee Requests Extension to Specified Schedule for Implementing Updates to IST Program PY-CEI-NRR-2438, Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl1999-10-14014 October 1999 Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl ML20216J6701999-09-24024 September 1999 Forwards Post Examination Documentation for Written Operator Initial License Examination Administered at Davis-Besse Nuclear Power Station on 990920.Without Encls 05000346/LER-1998-001, Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached1999-09-0909 September 1999 Forwards Rev 1 for LER 1998-001,which Updates Corrective Actions & Revises Completion Date Re Implementation of Changes to Plant Emergency Operating Procedure.List of Commitments Attached ML20216E5961999-09-0707 September 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1,safety Features Actuation Sys Instrumentation & Associated Bases 3/4.3.1 & 3/4.3.2,reactor Protection Sys & Safety Sys Instrumentation ML20211K6611999-08-30030 August 1999 Forwards Copies of Operator License Renewal Applications for Individuals Listed.Operators Have Successfully Completed Appropriate Operator Requalification Training Program at Dbnps.Without Encls ML20211K6681999-08-30030 August 1999 Forwards Copies of Certified Personal Qualification Statement - Licensee (NRC Form 398) for Operator Candidates Listed Below.Without Encls ML20211H0201999-08-25025 August 1999 Forwards semi-annual FFD Rept for 990101-0630 for DBNPS, Unit 1,IAW 10CFR26.71(d) ML20211G3911999-08-20020 August 1999 Forwards Update to Estimated Info for Licensing Action Requests Through 010930,re Administrative Ltr 99-02, Operating Reactor Licensing Action Estimates PY-CEI-NRR-2411, Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl1999-08-19019 August 1999 Informs That Firstenergy Nuclear Operating Co Has Developed Corporate QA Program Manual for Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant,As Discussed on 990318 Between Util & Nrc.Revised USAR Pages,Encl ML20211J9201999-08-13013 August 1999 Urges NRC to Find Funds for Stockpiling Radiation Pills for Residents Living Near Plant ML20210T1061999-08-12012 August 1999 Forwards Preliminary NRC Forms 398 & 396 for Listed Candidates,Per Operator License Exam Scheduled for Week of 990913.Encl Withheld ML20210S6071999-08-11011 August 1999 Provides Final Response to NRC RAI Re GL 98-01, Y2K Readiness of Computer Systems at Npps 05000346/LER-1998-009, Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl1999-08-0606 August 1999 Forwards LER 98-009-01,IAW 10CFR50.73(a)(2)(ii)(B). Commitments Made by Util Are Encl ML20210G5521999-07-28028 July 1999 Provides Addl Response to 980923 OL Licensing Exam Rept 50-346/98-301 Re OL Exam Administered in Aug 1998.Results of Root Cause Investigation & Corrective Actions,Discussed ML20210H0491999-07-28028 July 1999 Forwards Application for Amend to License NPF-3,revising TS 3/4.7.5.1, Ultimate Heat Sink, to Allow Plant Operation in Modes 1-4 with Water Temp Less than or Equal to 90 F ML20210G3831999-07-27027 July 1999 Forwards Application for Amend to NPF-3,changing TSs 6.4, Training, 6.5.2.8, Audits, 6.10, Record Retention, 6.14, Process Control Program & 6.15, Odcm ML20210G4391999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs 3/4.3.2.1, Safety Features Actuation Sys Instrumentation, & Associated Bases 3/4.3.1 & 3/4.3.2, Reactor Protection Sys & Safety Sys Instrumentation ML20210G7151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising TSs 3/4.3.3.1, Radiation Monitoring Instrumentation, 3/4.3.3.2, Instrument - Incore Detectors & 3/4.3.3.9, Instrumentation - Waste Gas Sys Oxygen Monitor ML20210G5151999-07-26026 July 1999 Forwards Application for Amend to License NPF-3,revising Tech Specs for Implementation of 10CFR50,App J,Option B for Type B & C Containment Leakage Rate Testing ML20211P3071999-07-26026 July 1999 Forwards Final Rept for 990504 Biennial Radiological Emergency Preparedness Exercise for David-Besse Power Station.No Deficiencies Identified for Any Jurisdiction During Exercise ML20210G3211999-07-26026 July 1999 Forwards Written OL Exam & Supporting Matl for Exam to Be Administered at DBNPS During Week of 990913.Listed Encls Withheld from Public Disclosure Until After Exam Complete 05000346/LER-1998-012, Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached1999-07-0707 July 1999 Forwards LER 98-012-01,which Is Being Submitted to Provide Addl Info Re 981018 Occurrence.Commitment List Attached ML20209C3981999-07-0101 July 1999 Responds to NRC Re Violations Noted in Insp Rept 50-346/98-21.Corrective Actions:Developed Rev to Boric Acid Control Program & Work Process Guideline on Plant Leakage ML20209B5821999-06-24024 June 1999 Provides Justification for Rev to Completion Date for One of Insp follow-up Items Cited in Insp Rept 50-346/98-03, Designated as Inspector follow-up Item 50-346/97-201-10 ML20196G1251999-06-23023 June 1999 Responds to NRC RAI Re GL 98-01, Y2K Readiness of Computer Sys at Nuclear Power Plants ML20196E5321999-06-17017 June 1999 Forwards Addl Info Re Relief Request RR-A16 to Support NRC Approval of Relief Request ML20196A6601999-06-16016 June 1999 Forwards Master Decommissioning Trust Agreements Revised After 1990 for Ohio Edison Co,Cleveland Electric Illuminating Co,Toledo Edison Co & Pennsylvania Power Co Re Bvnps,Units 1 & 2,DBNPS,Unit 1 & Perry Unit 1 ML20195F9071999-06-10010 June 1999 Forwards Application for Amend to NPF-3,changing Tech Specs 3/4.6.4.4, Hydrogen Purge Sys, TS 3/4.6.5.1, Shield Bldg Emergency Ventilation Sys & TS 3/4.7.6.1, Crevs ML20195F8851999-06-0707 June 1999 Withdraws 950929 License Amend Application,Proposing Mod to Allowable as-found Pressure Lift Setting Tolerance of Two Pressurizer Code Safety Valves ML20207F4231999-06-0202 June 1999 Forwards Rev 1 to DBNPS Emergency Preparedness Evaluated Exercise Manual 990504, IAW 10CFR50.4.NRC Evaluated Exercise Has Been Rescheduled for 991208,since NRC Did Not Evaluate 990504 Exercise ML20207E9561999-05-28028 May 1999 Forwards Update to NRC AL 98-03,re Estimated Info for Licensing Activities Through Sept 30,2000 ML20207E2521999-05-28028 May 1999 Forwards Rev 18,App A,Change 1 to Davis-Besse Nuclear Power Station,Unit 1,industrial Security Plan IAW Provisions of 10CFR50.54(p).Encl Withheld,Per 10CFR73.21 ML20207E7801999-05-21021 May 1999 Forwards Application for Amend to License NPF-3,allowing Use of Expanded Spent Fuel Storage Capacity.Proprietary & non- Proprietary Version of Rev 2 to HI-981933 Re Cask Pit Rack Installation Project,Encl.Proprietary Info Withheld ML20206N0231999-05-0606 May 1999 Forwards License Renewal Applications for Davis-Besse Nuclear Power Station,Unit 1 for ML Klein,Cn Steenbergen & CS Strumsky.Without Encls ML20206D2421999-04-28028 April 1999 Forwards Combined Annual Radiological Environ Operating Rept & Radiological Effluent Release Rept for 1998. Rev 11, Change 1 to ODCM & 1998 Radiological Environ Monitoring Program Sample Analysis Results Also Encl PY-CEI-NRR-2382, Forwards 1998 Annual Rept for Firstenergy Corp, for Perry Nuclear Power Plant & Davis-Besse Nuclear Power Station.Form 10-K Annual Rept to Us Securities & Exchange Commission for FY98 Also Encl1999-04-21021 April 1999 Forwards 1998 Annual Rept for Firstenergy Corp, for Perry Nuclear Power Plant & Davis-Besse Nuclear Power Station.Form 10-K Annual Rept to Us Securities & Exchange Commission for FY98 Also Encl ML20206B8831999-04-17017 April 1999 Forwards 1634 Repts Re Results of Monitoring Provided to Individuals at Davis-Besse Nuclear Power Station During 1998,per 10CFR20.2206.Without Repts ML20205K5641999-04-0707 April 1999 Forwards Response to NRC 980415 RAI Re GL 96-06, Assurance of Equipment Operability & Ci During Design-Basis Accident Conditions. Rept FAI/98-126, Waterhammer Phenomena in Containment Air Cooler Swss, Encl ML20205K3871999-04-0707 April 1999 Forwards Copy of Application of Ceic,Oec,Ppc & Teco to FERC, Proposing to Transfer Jurisdictional Transmission Facilities of Firstenergy Operating Companies to American Transmission Sys,Inc.With One Oversize Drawing ML20205J1171999-03-29029 March 1999 Forwards Rev 1 to BAW-2325, Response to RAI Re RPV Integrity, Per GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity. Rev Includes Corrected Values in Calculations PY-CEI-NRR-2377, Submits Decommissioning Repts for Bvps,Units 1 & 2,Davis- Besse Nuclear Power Station,Unit 1 & Perry Nuclear Power Plant,Unit 1,per 10CFR50.75(f)(1)1999-03-29029 March 1999 Submits Decommissioning Repts for Bvps,Units 1 & 2,Davis- Besse Nuclear Power Station,Unit 1 & Perry Nuclear Power Plant,Unit 1,per 10CFR50.75(f)(1) ML20205F5961999-03-27027 March 1999 Forwards Comments on Preliminary Accident Sequence Percursor (ASP) Analysis of 980624 Operational Event at Dbnps,Unit 1, as Transmitted by NRC Ltr ML20205D4791999-03-26026 March 1999 Forwards Rept Submitting Results of SG Tube ISI Conducted in Apr 1998.Rept Includes Description of Number & Extent of Tubes Inspected,Location & Percent wall-thickness Penetration for Each Indication of Imperfection ML20205L2031999-03-26026 March 1999 Submits Correction to Dose History of Tj Chambers.Dose Records During 1980-1997 Were Incorrectly Recorded Using Wrong Social Security Number.Nrc Form 5 Not Encl ML20205C7371999-03-25025 March 1999 Certifies That Dbnps,Unit 1,plant-referenced Simulator Continues to Meet Requirements of 10CFR55.45(b) for Simulator Facility Consisting Solely of plant-referenced Simulator.Acceptance Test Program & Test Schedule,Encl ML20205E3551999-03-19019 March 1999 Requests That Proposed Changes to TS 6.8.4.d.2 & TS 6.8.4.d.7 Be Withdrawn from LAR Previously Submitted to NRC ML20204J6361999-03-17017 March 1999 Forwards Firstenergy Corp Annual Rept for 1998 & 1999 Internal Cash Flow Projection as Evidence of Util Guarantee of Retrospective Premiums Which May Be Served Against Facilities PY-CEI-NRR-2375, Forwards List That Details Util Insurers,Policy Numbers & Coverage Limits for Two Power Plants,Per Requirements of 10CFR50.54(w)(3) Re Reporting of Property Insurance Coverage1999-03-15015 March 1999 Forwards List That Details Util Insurers,Policy Numbers & Coverage Limits for Two Power Plants,Per Requirements of 10CFR50.54(w)(3) Re Reporting of Property Insurance Coverage ML20204E6821999-03-12012 March 1999 Requests That Listed Changes Be Made to NRC Document Svc List for Davis-Besse Nuclear Power Station,Unit 1 1999-09-09
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f**9 cswEason ENEaGY 300 Mod 6 son Annue John P. S W E ]
Toledo, OH 43852-0001 Vk:e Preement - Nuclear 419 249 2300 Davis-Besse Docket Number 50-346 License Number NPF-3 Serial Number 2354 February 1, 1996 United States Nuclear Regulatory Commission j Document Control Desk !
Washington, D. C. 20555 )
1
Subject:
Response to NRC Request for Additional Information Regarding the )
Toledo Edison Response to Generic Letter 95-03, "Circumferential l Cracking of Steam Generator Tubes" (TAC No. M92238)
Ladies and Gentler.en: <
On April 28, 1995, the L. S. Nuclear Regulatory Commission issued Generic Letter (GL) 95-03, "Circumferent.al i Cracking of Steam Generator Tubes,* which requested steam generator tube information regarding the Davis-Besse Nuclear Power ]
Station. Toledo Edison (TE) responded by letter dated June 23, 1995 (Serial '
Letter Number 2304). The NRC's letter dated December 13, 1995 (TE Log Number 4656) requested additional information and contained an enclosure with eight questions. The additional information requested is provided in the attachment to this letter.
Should you have any questions or require additional information, please contact Mr. James L. Freels, Manager - Regulatory \ffairs, at (419) 249-2366.
Very truly yours, A \
FWK/llh Attachment I
cc: L. L. Gundrum, DB-1 NRC/NRR Project Manager I H. J. Miller, Regional Administrator, NRC Region III S. Stasek, DB-1 NRC Senior Resident Inspector Utility Radiological Safety Board 9602050261 960201 PDR ADOCK 05000346 P PDR 7 f Operating Companies.
D Cleveland Electnc liluminating
(
g Toledo Edison
. Docket Number 50-346
. License Number NPF-3
- Serial Number 2354
' Attachment Page 1 QUESTION 1:
Discuss the design differences between the Davis-Besse steam generators and the generic design information provided in the B&W Owners Group response, if any.
RESPONSE TO QUESTION 1:
) The Babcock and Wilcox Owners Group (BWOG) generic response said a Once-Through j Steam Generator (OTSG) typically has 15,531 tubes. The Davis-Besse Nuclear
- Power Station (DBNPS) OTSGs have 15,457 tubes each.
The BWOG generic response stated that the Tube Support Plates (TSPs) have broached holes. All TSPs in the DBNPS OTSGs have broached holes except the 15th TSP which has both broached and drilled holes. In the DBNPS OTSGs, 1,621 of the 15,457 tubes have drilled holes at the 15th TSP. The drilled holes are located around the periphery.
The BWOG generic response said the steam temperature is about 570 degrees Fahrenheit. The DBNPS steam temperature is approximately 591 degrees Fahrenheit.
QUESTION 2:
Dented Regions including dented tube support plates:
Electric Power Research Institute (EPRI) report NP-6201 "PWR Steam Generator Examination Guidelines: Revision 3," dated November 1992, indicated that B&W plants have experienced denting at tube support plates and in the lower tubesheet. Circumferential indications have been observed at dented areas in recirculating steam generators. If denting has been observed at Davis-Besse and it is in a location potentially susceptible to circumferential cracking, please submit the information requested in Generic Letter (GL) 95-03 per the guidance contained in the GL. If a voltage threshold is used for determining the threshold for examining dents, provide the calibration procedure used (e.g.,
2.75 volts on 4-20% through-wall ASME holes at 550/130 mix).
EPRI report NP-6201 indicates that the fifteenth tube support plate contains both broached holes and drilled holes. The drilled holes being prone to denting. Please clarify whether all of the tube support plates are of the broached hole designs or whether a number of them contain drilled holes.
. Discuss whether denting has been limited to the drilled hole locations, if applicable, or if it has been observed at other support plate intersections (i.e., broached holes).
. Docket Number 50-346
. License Number NPF-3 Serial Number 2354
' Attachment Page 2 RESPONSE TO QUESTION 2:
The DBNPS does consider " dings," regardless of their location in the OTSG, to be potentially susceptible to circumferential cracking. The information requested in the Generic Letter is provided below:
- 1) Evaluate recent operating experience with respect to the detection and sizing ,
of circumferential indications to determine the applicability to their plant.
" Dings" are considered to be potentially susceptible to circumferential cracking because of the associated increase in stress levels within the tubing. Note that tube diameter reductions in OTSGs are generally called s
" dings." The term " ding" has been used to distinguish them from the more severe classical denting observed in recirculating steam generators. Dings in OTSGs are most prevalent at the secondary face of the Upper Tubesheet (UTS) and Lower Tubesheet (LTS). Although not as common, dings have been detected at Tube Support Plates (TSP) and freespan locations also. Outside diameter degradation has been detected at dings in OTSGs.
The region of the OTSG from the 14th TSP and above is where dings would be most susceptible to circumferential cracking. This region would be most susceptible because the tubes are at the highest temperature there. The steam temperature in this region is a minimum of approximately 590 degrees ,
Fahrenheit and the reactor coolant temperature is a minimum of approximately 605 degrees Fahrenheit.
- 2) On the basis of the evaluation in Item (1) above, past inspection scope and results, susceptibility to circumferential cracking, threshold of detection, expected or inferred crack growth rates, and other relevant factors, develop a safety assessment justifying continued operation until the next scheduled steam generator tube inspections are performed.
The DBNPS has a relatively small number of dings spread throughout the OTSGs.
The current number of dings reported in in-service tubes and their location !
are as follows: '
SG 1 l
- 1) 14th TSP and above 11
- 2) TSPs 1 through 13 20
- 3) Spans 1 through 13 78
- 4) LTS secondary face 0 ;
- 5) LTS 13 :
Total 122 SG 2 1
- 1) 14th TSP and above 15 j
- 2) TSPs 1 through 13 1
- 3) Spans 1 through 13 9
- 4) LTS secondary face 82 i
- 5) LTS 12 l Total 119 l
. Docket Number 50-346
. Ificense Number NPF-3 Serial Number 2354
' Attachment Page 3 A limited number of dings in the region most susceptible to circumferential cracking were examined with 3-coil Motorized Rotating Pancake Coil (MRPC) probes during the last steam generator tube inspection. No circumferential indications were identified.
The area of greatest concern for OTSG tube dings at the DBNPS has been the freespan of the periphery tubes in the region of the abandoned and secured internal auxiliary feedwater header. This header is located in the span between the 15th TSP and the upper tube sheet secondary face. The dings were caused by loose parts (dowel pins, subsequently retrieved) contacting the tubes and from contact by the header itself. This event occurred in 1981. Damaged tubes in thi. area have been plugged or sleeved and, therefore, are not included in the data above.
The DBNPS OTSGs have dings on a small fraction of the inservice tubes. Of these, a small number are in locations that may make them more susceptible to cracking. As previously stated, 3-coil MRPC examinations of a limited number of dings in the most susceptible region for circumferential cracking during the previous steam generator inspection found no circumferential indications.
This indicates that circumferential cracking has not developed as a result of 4 dings present in the DBNPS OTSGs. I Continued operation until the next refueling outage, currently scheduled for April, 1996, is justified by the good performance of the DBNPS OTSGs. This good performance is shown by the low Reactor Coolant System leakage, typically less than 0.1 GPD. Continued operation is further justified by the relatively small fraction of dings in the DBNPS OTSGs and the absence of circumferential crack indice.tions.
- 3) Develop plans for the next steam generator tube inspections as they pertain to the detection of circumferential cracking. The inspection plans should address, but not be limited to, scope (including sample expansion criteria, if applicable), methods, equipment, and criteria (including personnel training and qualification).
The next planned DBNPS steam generator inspection is scheduled for April /May 1996. The currently reported dings in the region at the 14th TSP and above will be examined. In addition, a 20% sample of the remaining currently reported dings will be examined from throughout the steam generators.
If circumferential crack-like eddy current indications are reported at a ding, an engineering evaluation would determine the need for inspection expansion. The evaluation would consider the physical parameters affecting the tubing. These parameters may include tube temperature at the ding location, heat of tube material, and ding location within the tube bundle.
In addition, the evaluation would consider the results from supplemental I examinations as well as the size and quantity of indications. For new dings reported during the next scheduled outage, an engineering evaluation would !
determine the need for supplemental examinations and the increased scope necessary to bound any potential problems. The same criteria as discussed ,
above will be used in the engineering evaluation. l l
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. Docket Number 50-346
. L'icense Number NPF-3 Seridl Number 2354 .
Attachment l Page 4 The technique for the detection of circumferential cracking at dings will use a probe which has been qualified for detection of circumferential cracks per Appendix H of EPRI Report NP-6201, "PWR Steam Generator Examination Guidelines: Revision 3." The analysts will be qualified to SNT-TC-1A Level II or higher, as well as to site-specific guidelines with respect to detection of circumferential cracking. l The DBNPS does not use a voltage threshold for performing supplemental examinations on dings.
As explained above in the response to Question 1, all TSPs in the DBNPS OTSGs have broached holes except the 15th TSP which has drilled holes instead of broached holes around the periphery. Of the tubes in-service with dings, none have occurred at drilled hole locations.
QUESTION 3:
Expansion transition examinations:
Clarify if the eight tubes that did not receive the post-roll stress relief are still in service.
Clarify the inspections performed during the last outage at the expansion transition region. Address the probe used and the number of tubes inspected.
Provide the criteria to be used for determining whether expansion of the inspections for expansion transition indications is necessary.
RESPONSE TO QUESTION 3:
The eight tubes which did not receive post-roll stress relief are still in service.
Although the standard 0.510" bobbin probe (as part of the required Technical Specification inspection) was used to examine expansion transitions (non-post-roll stress relieved and post-roll stress relieved) during the last outage (October / November 1994), no supplemental examinations were performed at expansion transitions.
If circumferential crack-like eddy current indications are repcrted at an expansion transition, the criteria to evaluate the need for inspection expansion will be developed by an engineering evaluation. The evaluation would consider the physical parameters affecting the tubing. These parameters may include tube loads, tube temperature, steam generator operating history, heat of tube material, and tube location within the tube bundle. In addition, the evaluation would consider the results from supplemental examinations as well as the size and quantity of indications.
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. Docket Number 50-346
. Ificense Number NPF-3 Serial Number 2354
-Attachment Page 5 QUESTION 4:
Lane / Wedge Region:
Clarify the inspection scope in the lane / wedge region during the last steam !
generator tube inspections (including the probe type and number (and/or percentage) of tubes inspected).
Provide the criteria to be used for determining whether the expanded inspection scope around any identified indications adjacent to the sleeved lane / wedge region is bounded. l RESPONSE TO QUESTION 4: l The DBNPS has installed preventive sleeves in the tubes in the lane / wedge region. The inspection scope of these sleeves during the last steam generator tube inspection (October / November 1994) was as follows:
In steam generator number 1, 13.6% of the previously installed in-service sleeves in the lanc/ wedge region had the three joints in the sleeve examined.
The joints were examined with the " optimized" rotating cross wound probe. The examination revealed no indications. In steam generator number 2, preventive sleeving was installed in the lane / wedge region. The entire lane / wedge region was sleeved. All of the newly installed sleeves were examined with the cross wound / bobbin probe. No indications were reported.
If circumferential crack-like eddy current indications are reported in an unsleeved border tube, an engineering evaluation would determine the need for )
inspection scope expansion. The evaluation would consider the location of the tube within the region and the thermal / hydraulic co.ditions that the tube experiences. Tubes which experience similar thermal / hydraulic conditions in the lane / wedge region would be examined. A one tube wide border around the tubes with similar thermal / hydraulic conditions would be examined to assure bounding.
QUESTION 5:
Recently, several tubes have been pulled from B&W once through steam generators (OTSGs). Discuss any analyses performed on these pulled tubes for monitoring the development of circumferential cracking. For example, discuss the destructive and nondestructive examinations performed on these pulled tubes in the laboratory at the expansion transition area.
Docket Number 50-346 o L'icense Number NPF-3 Serial Number 2354
-Attachment Page 6 RESPONSE TO QUESTION 5:
Recent tube pulls from Crystal River 3 (1992 and 1994) and Oconee 3 (1994) have been performed to determine the cause of eddy current indications that were observed in areas of the steam generator where damage mechanisms are not well understood. These included both freespan indications and TSP indications, primarily in the boiling region of the OTSG. These indications did not have circumferential characteristics, therefore, specific analyses to monitor the development of circumferential cracking was not performed. The tubesheet expansion region was not an area of interest due to lack of observed degradation in this area. It was drilled out during the removal process to ease the removal of the rest of the tube. The tubesheet expansion area was therefore not available for laboratory examination.
QUESTION 6:
Clarify whether the inspection method to be used at Davis-Besse is qualified for the detection of circumferential cracks per Appendix H of Electric Power Research Institute (EPRI) report NP-6201 or whether a site specific qual'.fication program will be used. If using site specific qualification procedures, state the differences and provide the justification for these criteria including a discussion of pulled tube data to support the detectability of circumferential cracks in the field.
RESPONSE TO QUESTION 6:
The examination of unsleeved areas of SG tubing believed to be susceptible to circumferential cracking will utilize probes currently listed as qualified for detection of circumferential cracking per the requirements of Appendix H of EPRI Report NP-6201.
No inspection technique is currently qualified per Appendix H for the detection of circumferential cracks in OTSG sleeves. Examb.ation of sleeves during the Spring 1996 outage at the DBNPS will be conducted using a plus point probe, which Toledo Edison believes to be the best technique currently available for detection of circumferential cracking. This probe is currently qualified per Appendix H for detection of circumferential cracking in Westinghouse HEJ sleeves. Appendix H qualification for this technique for OTSG sleeves is currently being pursued in a project funded by the B&W Owners Group. This project is expected to result in successful qualification for detection of circumferential cracking per Appendix H by the end of 1996.
QUESTION 7:
Discuss the number and types of sleeves used at Davis-Besse along with their installation dates (i.e., month / year).
Clarify the extent of the inspections performed on these sleeved tubes during the last outage (e.g., 100% of all sleeve joints were examined with the
" optimized" rotating cross wound probe).
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Docket Number 50-346 L"icense Number NPF-3 Serihl Number 2354 Attachment Page 7 RESPONSE TO QUESTION 7:
There are 212 80" BWNT Alloy 690 sleeves in-service in steam generator number 1.
All 212 sleeves were installed in March / April 1993. 39 of these sleeves have minor scoring in unexpanded sections from installation tooling. There are 199 80" BWNT Alloy 690 sleeves in-service in steam generator number 2. All 199 were installed in October / November 1994.
The inspections performed during the last outage (October / November 1994) for steam generator number 1 were as follows:
62 of the 212 sleeves installed in steam generator number 1 were inspected using the bobbin cross wound probe in the unexpanded sections. Of these 62 sleeves, 31 (called sample set A in this letter) were also inspected using the optimized rotating cross wound probe in the expanded sections. The sample set A sleeves that received inspection by two different probes were part of the Technical Specification required sample.
39 of the 212 sleeves installed in steam generator number 1 received minor scoring in the unexpanded sections caused by tooling during installation. Eight of these sleeves with minor scoring were included in sample set A. The remaining 31 sleeves (called sample set B in this letter) were selected for the purpose of monitoring the minor installation tool marks.
Of the 62 sleeves inspected (sample sets A and B), one indication, in the upper roll, was identified in an installed sleeve of sample set A. A supplemental examination using a motorized rotating single coil pancake probe was performed.
After examining the data, the indication was dispositioned as permeability.
The inspections performed during the last outage (October / November 1994) for steam generator number 2 were as follows:
All 199 sleeves were examined with the bobbin cross wound probe following installation. No indications were identified.
QUESTION 8:
During the Maine Yankee outage in July / August 1994, several weaknesses were identified in their eddy current program as detailed in NRC Information Notice 94-88, " Inservice Inspection Deficiencies Result in Severely Degraded Steam Generator Tubes." In Information Notice 94-88, the staff observed that several circumferential indications could be traced back to earlier inspections when the data was reanalyzed using terrain plots. These terrain plots had not been generated as part of the original field analysis for these tubes. For the Rotating Pancake Coil (RPC) examinations performed at your plant at locations susceptible to circumferential cracking during the previoas inspection (i.e.;
previous inspection per your Generic Letter 95-03 response), discuss the extent to which terrain plots were used to analyze the eddy current data. If terrain plots were not routinely used at locations susceptible to circumferential cracking, discuss whether or not the RPC eddy current data has been reanalyzed using terrain mapping of the data. If terrain plots were not routinely used during the outage and your data has not been reanalyzed with terrain mapping of the data, discuss your basis for not reanalyzing your previous RPC data in light of the findings at Maine Yankee.
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. Docket Number 50-346
- Uicense Number NPF-3 Serial Number 2354
- Attachment Page 8 Discuss whether terrain plots will be used to analyze the RPC eddy current data at locations susceptible to circumferential cracking during your next steam generator tube inspection (i.e., the next inspection per your Generic Letter 95-03 response).
RESPONSE TO QUESTION 8:
Non-rotating probes (i.e., bobbin) were the primary probes used for inspection in the majority of the areas of the DBNPS OTSGs during the last outage. The only pre-planned area identified as susceptible to circumferential cracking that was inspected last outage with rotating probes were sleeves. DBNPS Site Specific Analysis Guidelines require the use of terrain plots in addition to the Lissajous display when examining the areas of the sleeve roll expansions and sleeve ends when using the " optimized" rotating cross wound probe. A three coil MRPC was also used to examine a limited number of dings identified with the bobbin probe. Although not required by the analysis guidelines at the time, terrain plots were used for the evaluation of the dings to ensure detection enhancement.
The DBNPS will require through Analysis Guidelines that terrain plots be used when using any rotating probes in areas susceptible to circumferential cracking for any future inspections.
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