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Category:Graphics incl Charts and Tables
MONTHYEARML20142A3972020-05-0505 May 2020 Annual Radiological Environmental Operating Report, Figure 32, Wind Rose Annual Average 10M ML20142A3982020-05-0505 May 2020 Annual Radiological Environmental Operating Report, Figure 31, Wind Riose Annual Average 75M ML20142A3992020-05-0505 May 2020 Annual Radiological Environmental Operating Report, Figure 30, Wind Rose Annual Average 100M ML16056A1392016-03-11011 March 2016 Correction to the U.S. Nuclear Regulatory Commission Analysis of Licensees' Decommissioning Funding Status Reports ML15239B2122015-09-0303 September 2015 Interim Staff Response to Reevaluated Flood Hazards Submitted in Response to 10 CFR 50.54(f) Information Request - Flood-Causing Mechanism Reevaluation ML15238B8652015-09-0303 September 2015 FHRR MSFHI Tables 1 and 2 ML14307B7072014-12-10010 December 2014 Supplemental Information Related to Development of Seismic Risk Evaluations for Information Request Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-T L-13-257, Notification of Completion of a License Renewal Commitment Related to the Review of License Renewal Application TAC No. ME4640) and License Renewal Application Amendment No. 45)2013-07-23023 July 2013 Notification of Completion of a License Renewal Commitment Related to the Review of License Renewal Application TAC No. ME4640) and License Renewal Application Amendment No. 45) ML12167A2622012-05-29029 May 2012 Enclosure - Davis-Besse 17RFO License Condition 2.C(7) SG Circumferential Crack Report ML1214500142012-05-23023 May 2012 NFPA 805 LAR Status Matrix - May 2012 ML1015201172010-05-28028 May 2010 Nozzle 4 - 52M Deposit Depth ML1005396232010-02-22022 February 2010 Lessons Learned Task Force Action Plan Regarding Stress Corrosion Cracking Dec 2009 (Final Update) ML0824903002008-10-0202 October 2008 ITSB Draft R, to Davis-Besse Amendment Conversion Improved Technical Specifications with Beyond Scope Issues ML0824902752008-10-0202 October 2008 Draft a, Attachment to Davis Besse Amendment Conversion Improved Technical Specifications with Beyond Scope Issues ML0824902802008-10-0202 October 2008 ITSB Draft La, to Davis Besse Amendment Conversion Improved Technical Specifications with Beyond Scope Issues ML0824902952008-10-0202 October 2008 ITSB Draft M, to Davis-Besse Amendment Conversion Improved Technical Specifications with Beyond Scope Issues ML0824902892008-10-0202 October 2008 ITSB Draft L, to Davis-Besse Amendment Conversion Improved Technical Specifications with Beyond Scope Issues L-08-094, Response to Request for Additional Information Regarding Measurement Uncertainty Recapture Power Uprate Amendment Application2008-03-12012 March 2008 Response to Request for Additional Information Regarding Measurement Uncertainty Recapture Power Uprate Amendment Application ML0727405102008-02-11011 February 2008 Lltf Action Plan Feb 2008 Update - Public ML0803700812008-02-0101 February 2008 Lltf Status of Recommendations February 2008 ML0728203072007-10-11011 October 2007 Electronic Distribtion Initiative Letter, Licensee List, Electronic Distribution Input Information, Division Plant Mailing Lists 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 ML0708602532007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 3. Background ML0708602612007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 5. Worldwide Industry Response to CRDM and Other Alloy 600 Nozzle Cracking ML0708602682007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 7. FENOC/Davis-Besse Response to CRDM Cracking and Boric Acid Corrosion Issues 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 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 ML0708602642007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 6. Boric Acid Wastage of Carbon Steel Components in Us PWR Plants ML0702603952007-01-31031 January 2007 Summary of Conference Telephone Call Regarding the Spring 2006 Steam Generator Inspections - Handouts ML0618002582006-06-27027 June 2006 Closure of Davis-Besse Operational Improvement Plan ML0611401302006-03-16016 March 2006 Ltr Fm J. Gutierrez, Morgan Lewis to S. Brock Requesting to Withhold Materials from Public Disclosure Which Was Submitted by FENOC - Affidavits of Terry Young and Warren Bilanin ML0525707982005-09-12012 September 2005 Documentation of Completed Post-Restart Process Plan for Transition of Davis-Besse from the IMC 0350 Oversight Process to the Reactor Oversight Program (ROP) ML0523801412005-08-31031 August 2005 Status of Davis-Besse Lessons Learned Task Force Recommendations - Last Update: August 31, 2005 ML0411301992004-04-21021 April 2004 Oversight Panel Final Restart Checklist Public Release Memo ML0414503382004-02-13013 February 2004 Oversight Panel Open Action Item List ML0414503372004-02-13013 February 2004 IMC 0350 Panel Process Plan ML0414503262004-02-13013 February 2004 Restart Action Matrix - Open Items Only ML0308000432003-01-0101 January 2003 Firstenergy Nuclear Operating Company, NRC Allegations and Employee Concerns Program Contact Trends, January 2002 - January 2003 ML0224601492002-08-29029 August 2002 Undated Data Chart, CDF, LERF, Rem ML0224601482002-08-29029 August 2002 Heat 69 Crack Growth - Bounding Weibull with 1.5 Shape Facto with Handwritten Notes ML0224601422002-08-29029 August 2002 Chart, Long-Term Results for Integrated Model with Handwritten Notes ML0224003452002-08-27027 August 2002 Failure Frequency Curve for Davis-Besse Conditions ML0224003402002-08-27027 August 2002 Draft Graphs for Flaw Curves and Stress Intensity Factors ML0530703682002-03-13013 March 2002 Timeline of Key Events Related to Reactor Vessel Head Boric Acid Wastage Rev.3, 3-13-02 ML0530704022001-04-20020 April 2001 Normal Sump Pump Samples 2020-05-05
[Table view] Category:Report
MONTHYEARL-23-188, Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments2023-08-0707 August 2023 Energy Harbor Nuclear Corp., Supplement to Application for Order Consenting to Transfer of Licenses and Conforming License Amendments 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 IR 05000346/20210902021-12-16016 December 2021 Reissue Davis-Besse NRC Inspection Report (05000346/2021090) Preliminary White Finding 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. ML22262A1522019-05-0101 May 2019 Framatome Inc., Document ANP-2718NP, Revision 007, Appendix G Pressure-Temperature Limits for 52 EFPY for the Davis-Besse Nuclear Power Station ML22202A4332019-04-30030 April 2019 Enclosure C: Framatome ANP-2718NP, Rev. 7, Appendix G Pressure-Temperature Limits for 52 EFPY for the Davis-Besse Nuclear Power Station L-18-108, Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformance Identified in Response to Regulatory Issue Summary 2015-06, Tornado Missile....2018-04-12012 April 2018 Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15-002 for Tornado-Generated Missile Protection Non-Conformance Identified in Response to Regulatory Issue Summary 2015-06, Tornado Missile.... ML18149A2812018-02-16016 February 2018 2017 ATI Environmental Inc. 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 L-17-088, Independent Spent Fuel Storage Installation Changes, Tests and Experiments2017-03-27027 March 2017 Independent Spent Fuel Storage Installation Changes, Tests and Experiments 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-16-148, Fatigue Monitoring Program Evaluation of Reactor Coolant Pressure Boundary Components for Effects of the Reactor Coolant Environment on Fatigue Usage (I.E., Environmentally-Assisted Fatigue)2016-04-21021 April 2016 Fatigue Monitoring Program Evaluation of Reactor Coolant Pressure Boundary Components for Effects of the Reactor Coolant Environment on Fatigue Usage (I.E., Environmentally-Assisted Fatigue) 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 ML15230A2892015-08-25025 August 2015 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force Review L-14-401, First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In2014-12-19019 December 2014 First Energy Nuclear Operating Company (FENOC) Expedited Seismic Evaluation Process (ESEP) Reports Response to NRC Request for Information Pursuant to 10 CFR50.54(f) Regarding Recommendation.1 of the Near-Term Task Force (NTTF) Review of In ML14353A0602014-11-0303 November 2014 2734296-R-010, Rev. 0, Expedited Seismic Evaluation Process (ESEP) Report Davis-Besse Nuclear Power Station L-14-289, Pressure and Temperature Limits Report. Revision 22014-09-22022 September 2014 Pressure and Temperature Limits Report. Revision 2 L-14-259, Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051)2014-08-28028 August 2014 Firstenergy Nuclear Operating Company'S (Fenoc'S) Third Six-Month Status Report in Response to March 12, 2012 Commission Order Modifying Licenses with Regard to Reliable Spent Fuel Pool Instrumentation (Order Number EA-12-051) ML14141A5252014-06-30030 June 2014 Staff Assessment of the Flooding Walkdown Report Supporting Implementation of Near-Term Task Force Recommendation 2.3 Related to the Fukushima DAI-ICHI Nuclear Power Plant Accident L-14-167, Report of Facility Changes, Tests and Experiments for the Period Ending May 26, 20142014-06-18018 June 2014 Report of Facility Changes, Tests and Experiments for the Period Ending May 26, 2014 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 L-14-104, Firstenergy Nuclear Operating Co. 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 Accident2014-03-11011 March 2014 Firstenergy Nuclear Operating Co. 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. MF0116 & MF0 ML13340A1632013-10-0909 October 2013 Davis-Besse Nuclear Power Station Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report Revision 1, Appendix C to Appendix G ML13340A1622013-10-0909 October 2013 Davis-Besse Nuclear Power Station Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report Revision 1, Appendix B (2 of 2) ML13340A1602013-10-0909 October 2013 Davis-Besse Nuclear Power Station Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report Revision 1, Appendix B (1 of 2) ML13340A1582013-10-0909 October 2013 Davis-Besse Nuclear Power Station Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report Revision 1 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 L-13-157, Generic Safety Issue 191 Resolution Plan2013-05-15015 May 2013 Generic Safety Issue 191 Resolution Plan 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. L-12-347, FENOC Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident2012-11-27027 November 2012 FENOC Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Flooding Aspects of Recommendation 2.3 of Near-Term Task Force Review of Insights from Fukushima Dai-ichi Accident ML13135A2452012-08-10010 August 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B - Seismic Walkdown Checklists (Swcs), Sheet 202 of 379 Through Sheet 379 of 379 ML13135A2442012-08-10010 August 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B - Seismic Walkdown Checklists (Swcs), Sheet 1 of 379 Through Sheet 201 of 379 ML13135A2432012-08-10010 August 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix a - Resumes and Qualifications ML13135A2422012-08-10010 August 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Cover Through Page 176 2023-08-07
[Table view] Category:Technical
MONTHYEARL-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. ML22262A1522019-05-0101 May 2019 Framatome Inc., Document ANP-2718NP, Revision 007, Appendix G Pressure-Temperature Limits for 52 EFPY for the Davis-Besse Nuclear Power Station ML22202A4332019-04-30030 April 2019 Enclosure C: Framatome ANP-2718NP, Rev. 7, Appendix G Pressure-Temperature Limits for 52 EFPY for the Davis-Besse Nuclear Power Station ML18149A2812018-02-16016 February 2018 2017 ATI Environmental Inc. 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) 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 ML0708602682007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 7. FENOC/Davis-Besse Response to CRDM Cracking and Boric Acid Corrosion Issues ML0708602612007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 5. Worldwide Industry Response to CRDM and Other Alloy 600 Nozzle Cracking ML0708602562007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 4. the Davis-Besse March 2002 Event ML0708602532007-03-15015 March 2007 Review and Analysis of the Davis-Besse March 2002 Reactor Pressure Vessel Head Wastage Event, Section 3. Background 2023-01-10
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Text
Appendix A Finite Element Stress Analysis of Davis-Besse CRDM Nozzle 3 Penetration
Appendix A Finite Element Stress Analysis of Davis-Besse CRDM Nozzle 3 Penetration Exponent developed a three-dimensional finite element model of the control rod drive mechanism (CRDM) Nozzle-3 penetration in the Davis Besse reactor pressure vessel (RPV) head for evaluation of weld residual and operational stresses in and around the J-groove weld. The modeled loading history included the buttering deposit, nozzle expansion, thirteen weld passes, the hydrostatic test and normal operating conditions.
The stress results can be used to predict crack growth in the nozzle wall.
Finite Element Model The finite element model of the Nozzle-3 penetration, shown in Figure A.1, represents a portion of the hemispherical head extending half the distance from Nozzle 3 to neighboring Nozzles 1, 6 and 11. Advantage was taken of a plane of geometrical symmetry passing through the nozzle centerline along a meridian of the hemispherical RPV head. This plane of symmetry allowed for simulation of the full nozzle using a model representing only half of the nozzle. Appropriate boundary conditions applied on the symmetry (or cut) plane cause the half-symmetry model to behave just like the full nozzle. This common modeling technique optimizes solution efficiency by avoiding redundant calculations.
The finite element model was developed from dimensional information obtained from the following Babcock & Wilcox (B&W) fabrication drawings and quality control records:
- Quality Control Inspection Drawing No. 156631 E-R24 BN63097.001 B0T0 1106 DB05 A-1
- Quality Control Inspection Drawing No. 154628 E45 Unlike most other similar B&W reactors, the Davis-Besse head was fabricated without counter bores in the CRDM nozzle penetrations6. This means that the nozzle shrink-fit interference was not uniform around the circumference of the nozzle due to the curvature of the head. The interference contact zone extended higher on the nozzle on the uphill side and lower on the nozzle on the downhill side.
Material Properties Four different materials models were used in the analysis. In each case, material properties such as the conductivity, specific heat, thermal expansion coefficient, elastic modulus and strength data were modeled as a function of temperature.
The hemispherical head was modeled as ASME SA-533, Grade B, Class 1 alloy steel.
The properties were taken from the ASME Boiler and Pressure Vessel Code7. The elastic modulus is given as a function of temperature in Figure A.3, the specific heat as a function of temperature in Figure A.4, the thermal conductivity as a function of temperature in Figure A.5, the thermal expansion as a function of temperature in Figure A.6 and strength as a function of the temperature in Figure A.7. The head material was modeled as elastic-perfectly plastic with temperature-dependent yield strength based on the yield curve in Figure A.7. The input stress-strain relations for the head material are shown in Figure A.8.
The cladding on the inside of the head is modeled as Type 308 stainless steel. The elastic modulus as a function of temperature is given in Figure A.9. Yield, tensile, and flow stress as a function of temperature for Type 308 stainless steel cladding is given in Figure A.10. Poisons ratio was taken as 0.29 over the entire temperature range. The density of Type 308 stainless steel is 499.4 lb/ft3. The thermal expansion coefficient as a function of temperature is shown in Figure A.11, the specific heat as a function of temperature in Figure A.12 and the thermal conductivity as a function of temperature in Figure A.13.
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The nozzle is modeled as the nickel-chrome-iron alloy Inconel 600. The stress-strain data for Inconel 600 was adapted from the ASME Code and Special Metals product brochures.
The 0.2%-offset yield points were taken from Special Metals data and Heat M3935 Cert.
The stress-strain data of Alloy 600 at 2300°F was taken from EPRI TR-103696, Fig. 7-3.
The final input stress-strain curves are shown in Figure A.14. Elastic modulus as a function of temperature was taken from the ASME Code and is shown in Figure A.15 and Poissons ratio as a function of temperature is shown in Figure A.16. Temperature-dependent thermal properties for Inconel 600 were obtained from the ASME Code. The coefficient of thermal expansion is given in Figure A.17, the thermal conductivity in Figure A.18, the specific heat in Figure A.19 and the emissivity Figure A.20. Density varies little with temperature, so a constant value of 528.8 lb/ft3 was used.
The J-groove weld and buttering was modeled as Inconel 182 (Alloy 182) weld filler metal. The thermal and elastic properties of the weld filler metal are the same as the INCONEL Alloy 600 properties. The work hardening behavior for Inconel 182 is shown in Figure A.21.
Loading History The Davis-Besse RPV head assembly was fabricated in a number of steps. Several key steps influenced the stress states in and around the CRDM nozzle J-groove welds. The hemispherical alloy-steel head was clad on the inside with a layer of Type 308 stainless steel weld metal. Following this, the CRDM nozzle penetrations were bored through the head and cladding. Then the J-grooves were cut on the inside end of each CRDM penetration to accept the multi-pass welds which join the CRDM nozzles to the head and create a watertight seal. The surface of the J-grooves was then buttered (or clad) with a layer of Alloy 182 weld metal. At this point, the entire head was given a thermal stress relief treatment at 1125°F. The CRDM nozzles were matched to the head penetrations and each was ground to final dimensions that provided a targeted interference fit of 0.0013 to 0.0015 inch. To install each slightly oversized nozzle into its head penetration, the nozzle was cooled to around -320°F in liquid nitrogen to shrink it and then inserted into the penetration. As the nozzle warmed to ambient temperature, it expanded into BN63097.001 B0T0 1106 DB05 A-3
contact with the head, locking it in position. Lastly, the multi-pass J-groove welds were applied to join the nozzles with the head and provide a watertight seal.
There was no post-welding thermal stress relief treatment given to the head after the J-groove welds were applied. Therefore, tensile residual stresses from weld cooling remained in and around the J-groove welds after head assembly fabrication. Prior to service, the entire reactor pressure vessel must pass an ASME-Code required hydrostatic pressure test to 3,125 psi (1.25 times its design pressure of 2,500 psi). This pressure test leads to some localized yielding in and around the J-groove welds due to their already high tensile residual stresses from welding. This yielding during the pressure test results in a slight reduction in the residual stresses.
In service, the reactor pressure vessel was subjected to the full-load operational pressure and temperature of the primary cooling water. During Cycle 13 at Davis-Besse, the RCS pressure averaged about 2155 psi and the temperature averaged just over 605°F.
The finite element model of Nozzle 3 was analyzed in the following steps to simulate the actual sequence of fabrication, testing, and operational loading:
- solidification and cooling of the J-groove weld buttering from 2240°F to 70°F,
- stress relief of the head, cladding, and J-groove buttering at 1125°F,
- cooling of the nozzle to -320°F, installation of the nozzle into the head, and warming of the nozzle back to 70°F to achieve the interference fit,
- sequential application of thirteen weld passes, each deposited at 2240°F and cooled below a maximum inter-pass temperature of 350°F,
- pressurization of the wetted internal surfaces up to the hydrostatic test pressure of 3,125 psi and back to zero pressure, and BN63097.001 B0T0 1106 DB05 A-4
- pressurization to the operating pressure of 2,155 psi at a temperature of 605°F.
Throughout the simulation, the boundary planes of the modeled head segment were constrained to allow only radial displacement. Such boundary conditions cause the model to behave as if it were part of the complete hemispherical head.
The analysis of the Nozzle-3 model involved both transient thermal and thermal-mechanical solutions. A coupled analysis technique was employed wherein a transient thermal solution and a thermal-mechanical solution is obtained for each small increment of load application. This type of incremental coupled analysis was necessary to appropriately simulate the nonlinear response (in both stress and displacement) of the nozzle to temperature and pressure loads.
J-groove Weld Buttering To simulate the buttering weld deposit and cool down, the elements representing the buttering were initially set to 2240°F. The remote cut boundaries of the truncated head portion of the model were held at 70°F during the weld deposition and cool down stages.
Radiation boundary conditions were applied to the exposed faces of the buttering.
Convective boundary conditions were applied to the exposed faces of the buttering and the top and bottom faces of the head. For both the head and nozzle, turbulent free convective film coefficients were used with an ambient-air sink temperature of 70°F.
Heat was also allowed to conduct away from the weld through the head. The buttering weld was allowed to cool to near ambient temperature. During the buttering analysis, the nozzle portion of the model was deactivated to simulate the fact that the nozzle was not yet installed.
Head Stress Relief After the buttering step, the temperature of the entire model was raised uniformly to the stress-relief temperature of 1125°F and then reduced back to 70°F. Since the yield strength of the materials is lower at higher temperatures, this step reduced the residual stresses around the buttering weld. However, the stress relief does not remove all BN63097.001 B0T0 1106 DB05 A-5
residual stress from the head, and the buttering and head cladding layers remained in tension after this step.
Nozzle Installation with Interference Fit Following the stress relief step, the nozzle portion of the model was activated and shrunk by cooling to -320°F. Frictional contact interfaces on the outside of the nozzle and the inside of the head penetration bore were activated, and the nozzle was warmed back to 70°F, engaging contact between the nozzle and head to accurately simulate the interference fit. The coefficient of friction between the head and nozzle was taken as 0.2.
At the end of the nozzle heating, there was complete contact between the nozzle and head.
J-groove Weld Deposition After the nozzle heat up, the thirteen weld passes were deposited sequentially starting at the root of the J-groove. The elements representing each weld pass were activated at 2240°F and allowed to cool to below 350°F before the next pass was deposited. The boundaries of the truncated head portion of the model remote from the nozzle were again held at 70°F for these thermal loading steps. Radiation boundary conditions were applied to the exposed faces of the weld and buttering as well as the adjacent, exposed exterior faces of the nozzle. Convective boundary conditions were applied as before on the exposed faces of the weld and buttering as well as the top and bottom of the head.
Hydrostatic Testing and Full-load Operation Following completion of the J-groove weld deposition steps, the hydrostatic test was simulated by ramping up the pressure on the inner surfaces of the nozzle and head to 3,125 psi and back down to zero. In the final step of the analysis, to simulate the full-load operating condition, pressure on the inner surfaces of the nozzle and head was ramped up the operating pressure of 2155 psi and the temperature of the entire model was simultaneously raised to 605°F.
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Stress Results Residual stresses are present in the region near the weld from the welding process. Under full-load operating conditions, the hoop stress in the vicinity of the downhill side of the weld on the symmetry plane is shown in Figure A.22. This stress tends to cause axial cracking. The stresses in this region are used to determine the driving force for axial crack growth. As can be seen from Figure A.22, both the weld and the buttering have yielded. Some small regions of the nozzle and the head that are in the immediate vicinity of the weld have also yielded but the stress decays quickly with distance from the weld.
The axial stress in the same region is shown in Figure A.23. The axial stress is about half of the hoop stress in this region.
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Figure A.1 Half-symmetry finite element model of Davis-Besse CRDM Nozzle-3 head penetration.
Figure A.2 Close-up view of 13-pass J-groove weld in Nozzle-3 model.
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SA-533 Gr.B Cl.1 Alloy Steel Elastic Modulus 35000 30000 Elastic Modulus (ksi) 25000 20000 15000 10000 5000 0
0 500 1000 1500 2000 2500 Temperature (°F)
Figure A.3 Elastic modulus of ASME SA-533 Grade B Steel as a function of temperature7,8.
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SA-533 Grade B Class 1 Alloy Steel Specific Heat 0.45 0.40 Specific Heat (BTU/lb/°F) 0.35 0.30 0.25 0.20 0.15 ASME Code Data from Reference 8 used 0.10 d to extrapolate ASME Code 0.05 curve above 1500 F.
0.00 0 500 1000 1500 2000 2500 Temperature (°F)
Figure A.4 Specific heat as a function of temperature for SA-533 Grade B Class 1 alloy steel 7,8.
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SA-533 Grade B Class 1 Alloy Steel Thermal Conductivity 40.0 Data for low and medium carbon steels from Thermal Conductivity (BTU/hr/ft/°F)
Reference 8 used to extrapolate ASME Code curve 35.0 above 1500 F.
30.0 25.0 ASME Code data 20.0 15.0 10.0 5.0 0.0 0 500 1000 1500 2000 2500 Temperature (°F)
Figure A.5 Thermal conductivity as a function of temperature for SA-533 Grade B Class 1 Alloy Steel 7,8.
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SA-533 Grade B Class 1 Alloy Steel Coefficient of Thermal Expansion 10 Thermal Expansion Coeff. (10 /°F)
-6 9
(Extrapolation >1500F) 8 7
6 5
0 500 1000 1500 2000 2500 Temperature (°F)
Figure A.6 Coefficient of thermal expansion for SA-533 Grade B Class 1 Alloy Steel 7 BN63097.001 B0T0 1106 DB05 A-12
SA-533 Grade B Class1 Alloy Steel Yield and Tensile Strength 90 80 Yield Strength Tensile Strength 70 60 Stress (ksi) 50 40 30 20 10 0
0 500 1000 1500 2000 2500 Temperature (°F)
Figure A.7 Strength as a function of temperature for SA-533 Grade B Class 1 Alloy Steel BN63097.001 B0T0 1106 DB05 A-13
SA-533 Gr.B Cl.1 Alloy Steel Stress Strain Behavior 50 0°F, 70°F 200°F 400°F 600°F 40 800°F Stress (ksi) 30 1000°F 20 1200°F 10 1500°F 0
0 0.001 0.002 0.003 0.004 0.005 0.006 Strain Figure A.8 Stress Strain behavior of SA-533 Grade B Class 1 Alloy Steel BN63097.001 B0T0 1106 DB05 A-14
Elastic Modulus Type 308 Stainless Steel Cladding 30000 25000 Elastic Modulus (ksi) 20000 15000 10000 5000 0
0 500 1000 1500 2000 2500 Temperature (°F)
Figure A.9 Elastic modulus as a function of temperature for 308 stainless steel cladding BN63097.001 B0T0 1106 DB05 A-15
Strength of Type 308 Stainless Steel Cladding 60 Tensile 50 40 Strength (ksi)
Flow 30 Yield 20 10 0
0 500 1000 1500 2000 2500 Temperature (°F)
Figure A.10 Strength as a function of temperature for Type 308 Stainless Steel BN63097.001 B0T0 1106 DB05 A-16
Thermal Expansion of Type 308 Stainless Steel Cladding 15 Thermal Expansion Coefficient (x10-6 /°F) 12 9
6 3
0 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.11 Coefficient of thermal expansion as a function of temperature for series 308 stainless steel BN63097.001 B0T0 1106 DB05 A-17
Specific Heat of Type 308 Stainless Steel Cladding 0.15 0.13 Specific Heat (Btu/lb/°F) 0.10 0.08 0.05 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.12 Specific Heat as a function of temperature for series 308 stainless steel BN63097.001 B0T0 1106 DB05 A-18
Thermal Conductivity of Type 308 Stainless Steel Cladding 20.00 Thermal Conductivity (Btu/hr/ft/°F) 15.00 10.00 5.00 0.00 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.13 Thermal Conductivity as a function of temperature for series 308 stainless steel BN63097.001 B0T0 1106 DB05 A-19
M A In c o n e l 6 0 0 S tre s s -S tra in D a ta 90 E la stic slo p e s (7 0 °F , 8 0 0 °F , 1 2 0 0 °F , 1 6 0 0 °F )
80 7 0 °F 70 8 00 °F 60 Stress (ksi) 50 12 0 0 °F 40 0 .2 % -o ffse t yie ld p o ints fro m S p e cia l M e ta ls d a ta a nd H e a t M 3 9 3 5 30 C e rt. (R T )
20 16 0 0 °F 10 0
0 .0 0 0 .0 1 0 .0 2 0 .0 3 0 .0 4 0 .05 0 .06 S tra in Figure A.14 Stress strain data for Alloy 600 BN63097.001 B0T0 1106 DB05 A-20
Inconel Alloy 600 Elastic Modulus 35 30 Elastic Modulus (ksi x 1000) 25 20 15 10 5
0 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.15 Elastic modulus as a function of temperature for Inconel Alloy 600 BN63097.001 B0T0 1106 DB05 A-21
Inconel Alloy 600 Poisson's Ratio 0.40 0.35 Poisson's Ration 0.30 0.25 0.20 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.16 Poisson's ratio as a function of temperature for Inconel Alloy 600 BN63097.001 B0T0 1106 DB05 A-22
Thermal Expansion of INCONEL Alloy 600 12 Thermal Expansion Coefficient (x10-6 /°F) 9 6
3 0
0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.17 Coefficient of thermal expansion as a function of temperature for INCONEL Alloy 600 BN63097.001 B0T0 1106 DB05 A-23
Thermal Conductivity of INCONEL Alloy 600 18 Thermal Conductivity (Btu/hr/ft/°F) 15 12 9
6 3
0 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.18 Thermal Conductivity as a function of temperature for INCONEL Alloy 600 BN63097.001 B0T0 1106 DB05 A-24
Specific Heat of INCONEL Alloy 600 0.20 Specific Heat (Btu/lb/°F) 0.15 0.10 0.05 0.00 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.19 Specific Heat as a function of temperature for INCONEL Alloy 600 BN63097.001 B0T0 1106 DB05 A-25
Emmissivity of INCONEL Alloy 600 1.00 Emmissivity 0.80 0.60 0 500 1000 1500 2000 2500 3000 Temperature (°F)
Figure A.20 Emmissivity as a function of temperature for INCONEL Alloy 600 BN63097.001 B0T0 1106 DB05 A-26
Figure A.21 Elastic-perfectly plastic stress-strain model for Alloy-182 weld filler BN63097.001 B0T0 1106 DB05 A-27
Figure A.22 Hoop stress under full-load operation BN63097.001 B0T0 1106 DB05 A-28
Figure A.23 Axial stress under full-load operation BN63097.001 B0T0 1106 DB05 A-29
References
- 1. Control Rod Mechanism Housing, Drawing No. 154632 E-R2, Babcock &
Wilcox Company, May 14, 1971.
- 2. Closure Head Assembly, Drawing No. 154631 E-R4, Babcock & Wilcox Company, September 15, 1971.
- 3. Closure Head Sub-Assembly, Drawing No. 154628 E-R5, Babcock & Wilcox Company, April 12, 1978.
- 4. Quality Control Inspection Drawing No. 156631 E-R2, April 24, 1972.
- 5. Quality Control Inspection Drawing No. 154628 E-R2, March 9, 1972.
- 6. Closure Head Assembly, Drawing No. 154631 E-R4, Babcock & Wilcox Company, September 15, 1971.
- 7. ASME Boiler and Pressure Vessel Code, 2001 Edition,Section II, Materials, Part D, Properties, American Society of Mechanical Engineers, 2001.
- 8. Structural Alloys Handbook, 1996 Ed., CINDAS/Purdue University.
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