ML15299A149
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ML15299A149 | |
Person / Time | |
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Site: | Davis Besse |
Issue date: | 07/30/2012 |
From: | Liano R, Munshi J, Reilly R Bechtel Power Corp |
To: | Office of Nuclear Reactor Regulation |
Shared Package | |
ML15299A142 | List:
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References | |
L-15-328 25593-000-G83-GEG-00016-000 | |
Download: ML15299A149 (90) | |
Similar Documents at Davis Besse | |
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Category:Report
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Response to NRC Request for Information Pursuant to 10 CFR 50.54 (F) Regarding the Flooding Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident ML14007A6702014-02-21021 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14042A2942014-02-19019 February 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Davis-Besse Nuclear Power Station, TAC No.: MF0961 ML13340A1592013-11-26026 November 2013 Davis-Besse Nuclear Power Station Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report Revision 1, Appendix a ML13340A1472013-11-26026 November 2013 Davis-Besse Nuclear Power Station & Perry Nuclear Power Plant - Response to RAI Associated with Seismic Aspects of Recommendation 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident (TAC Nos. 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[Table view]Submittal of Relief Request for Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements L-22-253, Submittal of Pressure and Temperature Limits Report, Revision 52023-01-10010 January 2023 Submittal of Pressure and Temperature Limits Report, Revision 5 L-22-211, Technical Specification 5.6.6 Steam Generator Tube Inspection 180-Day Report2022-09-29029 September 2022 Technical Specification 5.6.6 Steam Generator Tube Inspection 180-Day Report L-22-216, Submittal of Pressure and Temperature Limits Report. Revision 42022-09-27027 September 2022 Submittal of Pressure and Temperature Limits Report. Revision 4 L-22-149, Post Accident Monitoring Report2022-06-23023 June 2022 Post Accident Monitoring Report ML22202A4362022-04-0808 April 2022 Enclosure F: Updated Inputs to 52 EFPY P-T Operating Curves ML22202A4372022-03-0202 March 2022 Enclosure G: Framatome Inc. Document 86-9344713-000, Davis-Besse Reactor Vessel Embrittlement Fluence Reconciliation Through 60 Years ML21322A2892021-12-0909 December 2021 Approval of Plant-Specific Analysis of Certain Reactor Vessel Internal Components in Accordance with License Renewal Commitment No. 53 ML20302A3022020-09-25025 September 2020 1 to Technical Requirements Manual ML19255H0992019-10-10010 October 2019 Staff Assessment of Flooding Focused Evaluation L-19-189, 54010-CALC-01, Davis-Besse Nuclear Power Station: Evaluation of Risk Significance of Permanent ILRT Extension.2019-07-29029 July 2019 54010-CALC-01, Davis-Besse Nuclear Power Station: Evaluation of Risk Significance of Permanent ILRT Extension. 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Midwest Laboratory Radiological Environmental Monitoring Program L-17-270, Notification of Emergency Core Cooling System (ECCS) Evaluation Model Change Pursuant to 10 CFR 50.462017-09-0101 September 2017 Notification of Emergency Core Cooling System (ECCS) Evaluation Model Change Pursuant to 10 CFR 50.46 ML17086A0322017-03-31031 March 2017 Enclosure B to L-17-105, Areva Report ANP-3542NP, Revision 1, Time-Limited Aging Analysis (TLAA) Regarding Reactor Vessel Internals Loss of Ductility at 60 Years ML17026A0082016-12-31031 December 2016 Areva Report ANP-3542NP, Time-Limited Aging Analysis (TLAA) Regarding Reactor Vessel Internals Loss of Ductility for Davis-Besse Nuclear Power Station, Unit No. 1 at 60 Years (Non Proprietary) L-16-229, Submittal of Pressure and Temperature Limits Report, Revision 32016-07-28028 July 2016 Submittal of Pressure and Temperature Limits Report, Revision 3 L-15-288, Response to NRC Letter. Request for Information, Per 10 CFR 50.54(f) Regarding Recommendations 2.1. 2.3. and 9.3. of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident2015-10-0202 October 2015 Response to NRC Letter. Request for Information, Per 10 CFR 50.54(f) Regarding Recommendations 2.1. 2.3. and 9.3. of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident L-14-289, Pressure and Temperature Limits Report. Revision 22014-09-22022 September 2014 Pressure and Temperature Limits Report. Revision 2 ML14134A5172014-05-30030 May 2014 Staff Assessment of the Seismic Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident L-14-148, CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2014-05-19019 May 2014 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models ML14112A3152014-04-21021 April 2014 Review of Draft Plant-Specific Supplement 52 to the Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding ML14007A6702014-02-21021 February 2014 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Order EA-12-049 (Mitigation Strategies) ML14042A2942014-02-19019 February 2014 Mega-Tech Services, LLC, Technical Evaluation Report Regarding the Overall Integrated Plan for Davis-Besse Nuclear Power Station, TAC No.: MF0961 L-13-154, CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models2013-05-28028 May 2013 CFR 50.46 Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models ML13009A3752012-12-12012 December 2012 Enclosure B to L-12-444, Calculation No. 32-9195651-000, Equivalent Margins Assessment of Davis-Besse Transition Welds for 52 EFPY - Non-Proprietary. ML13008A0612012-08-10010 August 2012 Davis-Besse Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C, Area Walk-By Checklists, Sheet 21 of 139 Through End L-15-328, Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 7 of 72012-07-30030 July 2012 Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 7 of 7 ML15299A1502012-07-30030 July 2012 Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 6 of 7 ML15299A1492012-07-30030 July 2012 Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 5 of 7 ML15299A1482012-07-30030 July 2012 Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 4 of 7 ML15299A1472012-07-30030 July 2012 Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 3 of 7 ML15299A1462012-07-30030 July 2012 Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 2 of 7 ML15299A1442012-07-30030 July 2012 Enclosure B, Bechtel Report No. 25593-000-G83-GEG-00016-000, Effect of Laminar Cracks on Splice Capacity of No. 11 Bars Based on Testing Conducted at Purdue University and University of Kansas for Davis-Besse Shield Building. Part 1 of 7 ML12209A2602012-07-26026 July 2012 Attachment 31, Fauske & Associates, Inc. Technical Bulletin No. 1295-1, BWR MSIV Leakage Assessment: NUREG-1465 Vs. MAAP 4.0.2 ML1017400422010-06-0404 June 2010 0800368.407, Rev. 0, Summary of Design and Analysis of Weld Overlays for Reactor Coolant Pump Suction and Discharge, Cold Leg Drain, and Core Flood Nozzle Dissimilar Metal Welds for Alloy 600 Primary Water Stress Corrosion Cracking Mitigati L-10-132, 0800368.408, Revision 0, Summary of Weld Overlay Ultrasonic Examinations for Reactor Coolant Pump Suction and Discharge Nozzle Welds, Core Flood Nozzle Welds, and Cold Leg Drain Nozzle Welds2010-04-25025 April 2010 0800368.408, Revision 0, Summary of Weld Overlay Ultrasonic Examinations for Reactor Coolant Pump Suction and Discharge Nozzle Welds, Core Flood Nozzle Welds, and Cold Leg Drain Nozzle Welds ML1002501322010-01-11011 January 2010 0800368.404, Revision 1, Leak-Before-Break Evaluation of Reactor Coolant Pump Suction and Discharge Nozzle Weld Overlays for Davis-Besse Nuclear Power Station, Enclosure B ML11301A2222008-12-0101 December 2008 Reference: Combined Heat and Power Effective Energy Solutions for a Sustainable Future ML0821900132008-08-0707 August 2008 Monthly Operating Reports Second Quarter 2008 L-08-105, Reactor Head Inspection Report2008-04-11011 April 2008 Reactor Head Inspection Report L-08-005, Submittal of the 2007 Organizational Safety Culture and Safety Conscious Work Environment Independent Assessment Report for Davis-Besse2008-01-27027 January 2008 Submittal of the 2007 Organizational Safety Culture and Safety Conscious Work Environment Independent Assessment Report for Davis-Besse ML0726105652007-09-17017 September 2007 Confirmatory Order, 2007 Independent Assessment of Corrective Action Program (FENOC) ML0708602822007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Appendix B, Crack Driving Force and Growth Rate Estimates ML0708602812007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Appendix a, Finite Element Stress Analysis of Davis-Besse CRDM Nozzle 3 Penetration ML0708602802007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 10. the Unique Nature of the Davis-Besse Nozzle 3 Crack and the RPV Head Wastage Cavity ML0708602762007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 9. Cfd Modeling of Fluid Flow in CRDM Nozzle and Weld Cracks and Through Annulus ML0708602712007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 8. Stress Analysis and Crack Growth Rates for Davis-Besse CRDM Nozzles 2 and 3 ML0708602842007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Appendix C, Cfd Analysis 2024-06-05 |
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