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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for Approval of Quality Assurance Program Changes PY-CEI-NRR-2438, Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl1999-10-14014 October 1999 Informs That DBNPS & Pnpp Staffs Have Modified or Withdrawn Several of Positions Proposed within Re Request for Approval of Qap.Revised Positions Encl ML20217F8371999-10-0808 October 1999 Forwards Insp Rept 50-346/99-10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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.",, OWNERS GROUP I j .d::"'-llllh.,
Duke Pnwer Company Oconee 1,2,3 '~~
Entergy operations, Jnc. ANO-1 Tole @ Edison Corrpeny Onvin Besse 1 Fborsda PowerCorporation CrystalRiver3 '
Tennessee Veney Authorty
- ""**' corpararm w \\ g u ll ,
Bswearrow,movee Benebate1.2 Wcrking Together to Economically Provide Reliable and Safe Electrical Power Suite 525 e 1700 Rockville Pike e Rockville, MD 20852 e (301) 230-2iOO l
June 13,1995 i OG-95-1519 U.S. Nuclear Regulatory Commission l
Attn: Document Control Desk j
Washington, DC 20555-0001 '
i
Subject:
B&W Owners Group Generic Response to GL 95-03 Gentlemen:
The B&W Owners Group (BWOG) has reviewed NRC . Generic Letter 95-03 regarding
]
circumferemial cmcking 'of steam generator tubes and is submitting the attached generic response for all utilities with operating B&W Once-TLrough Steam Generators (OTSG). This generic BWOG response will be referenced and incorporated in each Utility's individual response to GL 95-03.
The circumferential cracking issue addressed.in GL 95-03 has some applicability to the B&W OTSG's. However, by virtue of the OTSG design and tubing material, the tubes have not been susceptible to many of the cracking mechanisms known to exist in other steam generator designs.
Preventive sleeving has minimized the occurrence of circumferential fatigue failures. Past tube inspections have concentrated on susceptible areas and circumferential cracking has not been identified as a significant degradation mechanism at B&W plants. The next inspection of OTSG's will continue to concentrate on these areas. As required by Generic Letter 95-03, the attached information provides the justification for continued operation of the B&W plants to their next scheduled steam generator inspection outage.
i Very truly yours, ed Y John A. Selva B&W Owners Group Steering Committee Chairman c: J.D.- Woodward (SGMP) i T.E.' Tipton (NEI) !
9506300309 950613 C. Welty (EPRI) PDR ADOCK 05000346 p PDR
June 13, 1995 Generic Response to GL 95-03 for B&W OTSG Relative to circumferqntial Crackina of Steam Generator Tubes I Anolicability of B&W OTSGs to Circumferential Crackina Issue 1.0 Introduction NRC Generic Letter 95-03 "Circumferential Cracking of Steam Generator Tubes" dated April 28, 1995 addressed circumferential cracking of steam generator tubes. It notified all holders of PWR operating licenses of recent steam generator tube inspection findings at Maine Yankee and the safety significance of these findings. The letter also requested a review of recent operating experience at each plant with respect to the detection and sizing of circumferential indications, in addition to a safety assessment to justify continued operation until the next scheduled steam generator tube inspection.
This document is a generic safety assessment which applies to the licensees whose plants use once-through steam i generators (OTSGs).
2.0 OTSG Backaround Information .
The OTSG is a vertical once-through steam generator i containing typically 15,531 5/8-inch diameter Alloy 600 straight tubes, each over 56-feet long (see attached Figure l 1). The tubes are aligned and' supported by 15 carbon steel I tube support plates (tsp) which have broached holes to l provide three-point support and a space for' axial flow around each tube. During manufacture each tube was roller expanded 1-inch deep at each primary face of the 24-inch thick tubesheets for tube-to-tubesheet welding. Each fully assembled steam generator underwent a furnace stress relief heat treatment process at 1100 to 3150 degrees for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. I At full power, the primary coolant enters the upper head at about 604 degrees F and exits the lower head at about 554 degrees F, while approximately 450 degree F feedwater is preheated in an annulus around the tube bundle and then boils and superheats as it flows axially up around the tubes. Superheated steam leaves the OTSG at about 570 degrees F.
The first OTSG went into operation at Oconee 1 in 1974, and the 14 now in operation have accumulated EFPYs ranging from 9 to 16 years. All OTSGs started up using AVT water treatment and full-flow condensate polishing, and the feedwater chemistry has generally been maintained well below the maximum EPRI guideline values.
3.0 Susceptible Circumferential Crack Locations The areas of the OTSG tubes which may be susceptible to circumferential cracking are (1) those which may be subjected to high-cycle fatigue, and (2) those roll transitions which have residual tensile stresses which are primarily in an axial orientation. These criteria indicate four regions which may be susceptible. However, it should be noted that service-induced circumferential cracking has been observed only in region (a) below,
- a. The uppermost span of tubes in the lane / wedge region,
- b. Tube-to-tubesheet expansion transitions, including 189 total tubes which have not been stress-relieved and are therefore expected to be more susceptible to cracking,
- c. Tube kinetic expansion transitions, which have not been stress relieved (TMI-1 UTS only), and
- d. Tube / sleeve expansions at sleeve roll transitions.
Each of these regions is discussed separately.
3.1 The Uonermost Soan of Tubes in the Lane /wedae Reaion This region consists of tubes in the lane / wedge region at elevations between the upper tubesheet secondary face (UTSF) and the uppermost tube support plate (15th TSP). The inspection lane is an un-tubed channel which extends from the periphery to the center of the tube bundle along Row 76 of the tube layout. The lane / wedge region has been defined as a region of tubes which abuts the outer half of this lane and then flares out into a wedge shape near the periphery of the tube bundle, per the attached Figure 2.
The steam flow in the uppermost tube span is primarily across the tubes as it turns to exit the bundle. The open lane is a cooler area of the generator which in turn lets wet steam deposit contaminants higher in the bundle, contributing to local degradation. The lower flow resistance of the lane / wedge region results in higher crossflow velocities, and this high velocity steam may cause flow-induced vibration of these tubes. This vibration could result in high cycle fatigue failure (Reference NRC l Information Notice 91-43).
i
E W .em Mn .
1 1
- I
- 3.2 Tube-to-Tubesheet Exnansion Transitions
^ ' "
.f~ lDuring manufacture of.the OTSGs each tube was roller ,, :
expanded 1-inch. deep:at the primary face of both tubesheets in preparation for-the tube-to-tubesheet welds. This" roller l expansion process produces a diameter _changeLin the tube _and~ -l leaves'the transition. area with residual stresses which are. (
primarily tensile on the OD surface. These cold-worked !
areas are expected to be_more susceptible to cracking:than the balance of the tube-which has not been cold-worked. ..
-Later the entire assembled OTSG was subjectedito a furnace f.
stress relief process, which reduced the residual stresses
^
J '
in the transitions and thermally treated the tubes.
However, a few tubes in each plant were re-expanded during _;
the shop hydrostatic: tests as a temporary seal until the !!
leaking tube could be rewelded. These deeper' expansions did .,
not have a post-roll' stress relief, which makes.them more l likely candidates for cracking. . .There are a total of 189 known non-stress relieved transitions existing in all the OTSGs according to manufacturing records, including:47 in l the UTS (hot 3eg upper tubesheet) and.142 in the LTS (cold leg lower tubesheet). Most of these expansions are in the ',{
Oconee-1 plant, including 29 located in the UTS. j Stress relief of Alloy 600 is neither required nor_ l prohibited by ASME Section III Code. However, this 12-hour j furnace stress relief at 1100 to 1150 F constitutes a :
f thermal treatment which both reduces the residual stresses and enhances the resistance of; Alloy 600 to caustic ;
corrosion. Therefore the roll transition joints which did' !
not get the standard treatment are likely to be'the location of the first tube transition failures in the OTSGs.1 Unlike1 :
an RSG, the tunes are normally in compression during operation, which reduces the likelihood of circumferential- ,
cracking in the' transitions. ~
l- Three tubes in the Oconee-3 B-0TSG were explosively expanded
, full length in both upper and lower tubesheets. It could i not be determined from available records whether the expansions were performed before or after the: full bundle I stress relief. !I 3.3 Tube Kinetic Expansion Transitions (TMI-1 UTS'only)
, , During an extended cold shutdown period, the tubes of the i TMI-1 steam generators were found to have many ID-initiated IGA assisted cracks near the top of the upper tubesheet
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, (Reference NRC Information Notices 82-14 and 84-18). This ;
was a singular event due to a reduced sulphur excursion ;
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during the shutdown. Some tubes were plugged, and all tubes remaining in service were kinetically expanded part-length through the upper tubesheet to create a new load-carrying and essentially leak-tight joint. The lower transition of the new expanded length has not been subjected to a stress relief process. These kinetic expansion transitions are similar to the rolled tube transitions discussed above, and could be considered to be potential sites for circumferential cracking.
3.4 Tube / Sleeve Exoansions at Sleeve Roll Transitions Mechanical sleeves have been installed in all operating B&W plants in order to mitigate tube leaks due to high cycle fatigue and to repair tubes with other UTS or 15th TSP degradation. Sleeves are in service that are fabricated from both Alloy 600TT and Alloy 690TT material. The mechanical sleeve has roller expanded joints at each end to seal it into the parent tube. These tube expansion transitions have had no stress relief process. Thus, consistent with earlier discussions, this is a potential site for circumferential cracking. However, the orientation of the crack can not be reliably predicted because the resultant of the residual and operational stresses may vary.
Six tubes were sleeved at Oconee-1 in 1979 on a test basis with hydraulically expanded sleeves. Four of these had short sleeves at both UTSF and 15th TSP, and the other two were sleeved only at UTSF. The four sleeved tubes in the lane region have subsequently been plugged for reasons unrelated to the sleeves, but two of the double-sleeved tubes remain in service. These hydraulic expansions are also potential sites for circumferential cracking.
4.0 Scone of Past Tube Inspections Past tube inspections have concentrated upon the susceptible regions described above. The susceptible regions have been inspected as follows:
4.1 The upper span of any unsleeved tubes in the lane / wedge region has been examined by MRPC at the UTSF and 15th TSP.
4.2 Duke Power has inspected all the non-stress relieved UTS 1 roll transitions in the oconee-1 plant during the most l recent inspection outages using MRPC. I I
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4'.3 Kinetic transitions in the. lane / wedge region at TMI-l have I been frequently examined using an 8x1 array pancake coil probe.
l 4.4 All plants have examined a portion of their in-service tube l sleeves'using both standard and crosswound bobbin coils at l each scheduled inspection. In' addition, some sites have '
l recently used rotating plus point and rotating crosswound techniques for examination of sleeve rolled joints. During the Spring 1995 outage ANO-1 used the plus point probe to ,
inspect the rolled joints of approximately 222 Alloy 600 sleeves, which included the 50 sleeves which have the .
longest service life of any rolled sleeves in an operating .
B&W OTSG. Ten of these sleeves have now seen over 6 EFPY of operation. ,
5.0 Past Tube Inspection Results The results of past tube inspections related to circumferential cracking are listed for the four susceptible :
locations described above: !
e 5.1 Uppermost Span of Tubes in the Lane / wedge Region There have been 41 OTSG tube leaks which have been attributed to circumferential cracks caused by high cycle fatigue in the lane / wedge region. These cracks generally occur where the bending is restrained at the upper edge of the 15th TSP and at the secondary face of the upper tube 1 sheet, initiating in any local discontinuities. Tube failure due to high cycle fatigue usually progresses rapidly ,
after crack initiation, with primary to secondary leakage typically occurring within a period of hours.
The OTSG plant licensees have relied upon the installation f of preventative sleeves in the top span of these tubes to minimize tube failures. The 80-inch mechanical sleeve stiffens the tube and provides an inner pressure boundary.
All OTSG plants now have completed sleeving tubes in their defined lane / wedge regions. The annual rate of OTSG tube leaks decreased as the preventive-sleeving program neared -
completion, confirming *he effectiveness of this program.
Note that sleeving mitigates the consequences of fatigue failures, and it is not dependent upon inspection by non-destructive examination.
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5'. 2 Tube Roller Expansion Transitions None of the rolled tube transitions have leaked or shown service-induced eddy current indications to date, whether stress relieved or not.
5.3 Tube Kinetic Expansion Transitions There have been no confirmed service-induced tube cracks or '
leaks found in the TMI-1 kinetic expansions to date.
5.4 Tube / sleeve Expansions at Sleeve Roll Transitions To date, no sleeved tubes have been removed from service due to service-induced degradation in the sleeved area of the parent tube or sleeve.
II. Justification for Continued Oper3 tion The B&W plants by virtue of their design and tubing material have not been susceptible to many of the circumferential cracking mechanisms known to exist in other plants. The OTSG plants have not identified any service-induced circumferential cracking in the tubes or sleeves other than fatigue-induced cracks in the upper span of the lane / wedge region. Degradation in this region has been adequately addressed by preventative sleeving, as discussed in Section I. Inspections of the other susceptible regions have not identified circumferential cracks.
The inspection techniques used a't B&W plants are capable of ;
reliable detection of circumferential cracking. These {
techniques have been shown to prevent uncontrolled tube i degradation. l If a tube should develop a circumferential crack and sever I at a rolled or kinetic expansion transition, the i consequences of this failure are minimized because the j severed ends of the tube would be captured within the tubesheet. In a similar manner, the sleeve prolongation beyond the rolled joint would keep the ends of the tube from separating if a tube should sever at a rolled sleeve j transition. The consequences of a circumferential crack ;
which severs a tube are minimized by these design features.
Identified circumferential cracking has been successfully repaired at B&W plants. The inspection techniques which have been used will reliably detect the onset of e
circumferential cracking where it is known to occur. The potential for the rupture of a tube is very low, and the consequences of a severed tube are mitigated by the capture provisions of the design. Therefore, based upon the demonstrated success of preventative sleeving and the lack of observed failures in other areas believed to be most susceptible to circumferential cracking, there is no known risk incurred by continuing to operate the OTSG plants.
Continued operation of the OTSG plants until the next scheduled tube inspection is safe and is therefore justified.
III. Plans for the Next Inspection for Circumferential Crackina 1.0 Scone of the Insoections The specific locations considered to be most susceptible to circumferential cracking in B&W OTSGs will be inspected at the next scheduled tube inspection outage as follows:
1.1 The Uppermost Span of Tubes in the Lane / Wedge Region Tubes in the lane / wedge region have been preventively sleeved in the uppermost span to mitigate the consequences !
of fatigue failures. A one-tube wide border of un-sleeved tubes in service around the lane / wedge region of sleeved tubes will be examined to confirm that the sleeved region is still adequately bounded. This examination will consist of inspection of the 15th TSP and UTSF areas of each tube.
Examination of sleeved tubes is addressed in item III.1.4 below.
If cracking is confirmed in the uppermost span of the border tubes surrounding the sleeved lane / wedge region, the inspection scope will be expanded until the area of cracking is demonstrated to be bounded.
1.2 Tube-to-Tubesheet Expansion Transitions Each of the rolled tube transitions which has not been stress relieved will be examined if it has not been sleeved or plugged . This includes transitions within both UTS and LTS. Note that these transitions in OTSG tubes are deep i
within the tubesheet where the signals are not distorted by dents or the face of the tubesheet.
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4 If cracking is-confirmed in the rolled expansion transitions which have not been stress relieved, sample expansion into the stress relieved transitions will be evaluated.
1.3 Tube Kinetic Expansion Transitions The examination of unsleeved tubes in the lane region at TMI-1 will include inspection of the kinetic expansion trans.itions for those tubes (approximately 280 tubes per OTSG). Indicfcions of circumferential cracking in the kinetic expansion transitions will be cause for additional examinations as necessary to bound that occurrence.
Leak testing by bubble test is scheduled for both TMI-1 OTSGs during the next refueling outage; such leak testing is capable of identifying upper tubesheet kinetic expansion transition crack leaks for follow-up eddy current examination and repairs as required. Note that previous leak tests since 1985 have consistently found that there-were no leaking kinetic expansion transitions.
1.4 Tube / sleeve Expansions at Sleeve Roll Transitions.
A 20 percent random sample of all sleeved tubes will be selected for examination. This examination will include tubesheet and freespan expansions in the pressure boundary portions of the sleeve and parent tube.
If cracking is found in the sleeved tubes, the sample will be expanded to 100% of the installed sleeves or until the expanded inspection scope demonstrates that the cracking is localized to a specific area or population of tubes.
2.0 Inspection Methods and Eauipment The technique to be used for the detection of circumferential cracking in OTSG tubes and sleeves 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", or other site-specific qualification.
The analysts will be qualified to SNT-TC-1A Level II or higher, as well as to site-specific guidelines with respect t to detection of circumferential cracking. ;
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The circumferential cracking issue is.applicablecto the B&W. .
OTSGs. By virtue of the design'and the tubing material, the j tubes have not been susceptible to many of the cracking j mechanisms known'to exist in other. steam generator designs. l' Preventive sleeving has' minimized the occurrence of fatigue'-
failures. Past tube inspections have concentrated upon !
certain areas ~of the tubes which may be susceptible', and i circumferential cracking has.not been identified as a i significant degradation mechanism at B&W plants.-
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The inspection techniques which have been used are capable ,
of reliable detection of circumferential cracking. The i potential for the rupture of a tube remains acceptably low :
and the. consequences of a rupture.are mitigated by the j design features. The next inspect' ions of OTSGs will- s continue to concentrate upon areas which may be susceptible. l to circumferential cracking. . :Thus,Las required by Generic Letter 95-03, it is concluded that continued operation of t the.OTSG plants until the next scheduled tube inspection ~1s .i justified. ,
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