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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
BWXT SERVICES, INC. 1140-025 02-24 72 BWXT SERVICES, INC. 1140-025-02-24 72 Clad Profile - Bottom Surface 4.21 14
-0.16
-0.18
-0.2
-0.22 y -0.24
-0.26
-0.28
-0.3
-0.32
-0.34
-2
,Q, X 2 Figure 3.5.6: Colorized representation of bottom surface of cladding (RCS side).
Red area denotes area of maximum deflection.
5
.. X~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
BWXT SERVICES, INC. 1140-025-02-24 73 0O 30" 0.29" 0.28" U I026 0.27" 0.25" 0 24" U 0.23" 0.22" 0.21i 0 20" U
Figure 3.5.7: Exposed cladding surface with superimposed colorized thickness data.
The thinnest areas of cladding occurred adjacent to the J-groove weld toward the 90° (0.202") and 270° (0.208") locations. U U
U U
U
U U
BWXT SERVICES, INC. 1140-025-02-24 74 U
U U
U U
U U
Figure 3.6.1: Photograph showing crack in exposed cladding in deflected region. U Red dot indicates (0, 0) coordinates for thickness measurements.
U U
U U
U U
BWXT SERVICES, INC. 1140-025-02-24 75 Figure 3.6.2: Mosaic showing cracking in exposed cladding. 2X
BWXT SERVICES, INC. 1 140-025-02-24 76 BWXT SERVICES, INC. 1140-025-02-24 76 Figure 3.6.3: Higher magnification view of cladding crack in bulged region. -4X
BWXT SERVICES, INC. 1 140-025-02-24 77 BWXT SERVICES, INC. 1140-025-02-24 77 Figure 3.6.4: Higher magnification photograph of crack near center bulge. 25X
BWXT SERVICES, INC. 1140-025-02-24 78 BWXT SERVICES, INC. 1140-025-02-24 78 Figure 3.6.5: Low magnification photograph of the cavity underside. The red circle shows the area of maximum deflection. The white dots were applied at Davis-Besse for the UT thickness measurements. The two parallel lines toward the left indicate the Nozzle 11 position, which is outside the photo on the left. This maximum bulging area is shown at higher magnification in the next figure.
BWXT SERVICES, INC. 1140-025-02-24 79 BWXT SERVICES, INC. 1140-025-02-24 79 Figure 3.6.6: Higher magnification photographs of cavity underside in area of maximum deflection. No cracks were observed in this area.
BWXT SERVICES, INC. 1 140-025-02-24 80 Top view of J-groove weld crack near 100. -. 4X Figure 3.6.7: Photographs showing axial crack in nozzle #3 J-groove weld near 10°.
BWXT SERVICES, INC. 1140-025-02-24 81 BWXT SERVICES, INC. 1140-025-02-24 81 Figure 3.6.8: Axial cracks located on J-groove weld bore near 1800. 0.75X
BWXT SERVICES, INC. 1140-025-02-24 82 BWXT SERVICES, INC. 1140-025-02-24 82 LOOuUMI9 LuWa[uraIu -U.;7A Figure 3.6.9: Low magnification photographs of cavity sidewall.
BWXT SERVICES, INC. 1140-025-02-24 83 BWXT SERVICES, INC. 1140-025-02-24 83 Figure3.6.10: Highermagnification photographs taken nearthe nozzle#3 bore.
BWXT SERVICES, INC. 1140-025-02-24 84 Figure 3.6.11: Photographs showing the cavity sidewall near 2700 (top) and looking up at the cavity nose (bottom).
BWXT SERVICES, INC. 1140-025-02-24 85 S s A Xr ES , a, e # U AS' okX ail /i ' _ r qr Ma S
<Inkin ;Sm J 'is<,
' t _ 5 M i.*' 4/
,,,..liN 6
,,C..0< ,SME,, l w . - w it. no, t Atn k our 4 =° lay l w - l *_ _
Figure 3.6.12: Photographs showing undercut regions of cavity near 300 (top) and 3150 (bottom).
p C031
BWXT SERVICES, INC. 1140-025-02-24 86 Pc.1 _
o.5".I Pc.B2 SS Cd I 0.25T Side View I Weld 28 . 22 5AP C . PcB2C1 L t.+PC.62C283 Direction 2.25-'
I PC.B2A ,-- -- -- - -- -- - --- Pc.02B
/~~ ~ ~~~~~~~~~~~
_-*TP.2C2B3t
/ t Pc.82C2A * -B C2122
.- :PCB2C2tal Top View of Pc.B2 I - 3.5" 12.5" Section Line - - -
Orawig not to Scale Figure 4.1.1: Schematic showing section locations and sample identifications for Block "B".
BWXT SERVICES, INC. 1140-025-02-24 87 Polished Surface I ! ! , ! ', X' Pc.B2C2B3
~~~~~~~~Pc.B2C2B2 i Pc.B2C2 Top View Pc.B2C2B1 I ~~~~~~~~~~~~~~i; Low Alloy Stool O.75-i_.._........._......_.._.._.._.._.._..iI Specimen B2C2A2 -
J- e-@@-......................
Specimen B2C2AI - SS Clad l_.._._........ ...... ..... ..... _.
Pc.B2C2 Side View I1Z I
-3.0" - + 0.5"H-Weld Direction Section Liic - - - -
EDM Section -*-
Drawing not to Scale Figure 4.1.2: Schematic showing tensile specimen locations. The polished surface of piece B2C2B3 (met sample) is also indicated.
BWXT SERVICES, INC. 1140-025-02-24 88 BWXT SERVICES, INC. 1140-025-02-24 88 Figure 4.2.1: Macro photograph of Sample B2C2B3 showing the low alloy steel stainless steel interface and various grain structures within the low alloy steel HAZ.
BWXT SERVICES, INC. 1 140-025-02-24 89 BWXT SERVICES, INC. 1140-025-02-24 89 Figure 4.2.2: Typical low alloy steel microstructures near bond. 2% nital
BWXT SERVICES, INC. 1140-025-02-24 90 BWXT SERVICES, INC. 1140-025-02-24 90 Figure 4.2.3: Typical low alloy steel microstructures 1/4" from bond. 2% nital
BWXT SERVICES, INC. 1 140-025-02-24 91 1140-025-02-24 91 BWXT SERVICES, INC.
Figure 4.2.4: Typical low alloy steel microstructures -1/2" from bond. 2% nital
BWXT SERVICES, INC. 1140-025-02-24 92 1' r= . *-r .- *.. V
".5.
..-i.. 3 ,.....* *
- 5. -
V..
PS V 33 (45VV IV
. 5..# tJ*. *
I OOX 0* * . . -. . . Si
a
- .V5.Z*t 43
- '55. -
-.5-
.3
.-0 1
55
.fs-- -
.. f* - Si. J . . -
- 4. .4 .- . .
I
- L) 1** '4
- /44 .44 4,
--4.
1
(.5 375X Figure 4.2.5: Typical cladding microstructures near bond.
(Acetic-nitric-hydrochloric etch)
BWXT SERVICES, INC. 1140-025-02 24 93 1140-025-02-24 93 BWXT SERVICES, INC.
Figure 4.2.6: Typical cladding microstructures away from bond.
(Acetic-nitric-hydrochloric etch)
BWXT SERVICES, INC. 1140-025-02-24 94 BWXT SERVICES, INC. 1140-025-02-24 94 Knoop Hardness (500 gram) 350.0 -
300.0-
= 250.0 c200.0- Al i 150.0-0 Stainless Clad Low Alloy Steel
,, 100.0 -
50.0 -
0.0 - l l l 0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 0.55 0.60 0.65 0.70 0.75 Distance, inches
.6 w It oy" teel 4o Figure 4.2.7: Microhardness data and low magnification photograph for Sample B2C2B3.
BWXT SERVICES, INC. 1140-025-02-24 95 1140-025-02-24 95 BWXT SERVICES, INC.
Figure 4.3.1: Tensile specimen design (dummy test specimen shown).
BWXT SERVICES, INC. 1 140-025-02-24 96 Tensile Test Comparison 60
_50 -
t 40 a)
"O 30 C
CD 20 -
101 0
0 5 10 15 20 25 30 35 Engineering Strain (%)
l-B2C2A1 -B2C2A2 B2C2A1 B2C2A2 UTS 54,800 psi 57,100 psi 2% Offset YS 30,500 psi 31,300 psi Elongation 28.7% 28.7%
Reduction in Area 39.3% 34.3%
Figure 4.3.2: Tensile test results for specimen B2C2A1 (near RCS) and specimen B2C2A2 (near low alloy steel).
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BWXT SERVICES, INC. 1140-025-02-24 97
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- T - -J r ip e Specimen :C2A1 -1.8X Specimen B2C2A2 1.8X Figure 4.3.3: Low magnification photographs of tensile specimens after test.
BWXT SERVICES, INC. 1140-025-02-24 98
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B2C2A2 5X Figure 4.3.4: Higher magnification photographs showing fracture location of tensile specimens.
BWXT SERVICES, INC. 1140-025-02-24 99 Table 5.1: Sample identification listing for Piece A2A.
Sample Location Test Plan Met SEM A2A1 A2A section at -2250 No plan l A2A2 A2A section at -170°-1 900 Axial cracks at -1800, see 1 Table 5-2 A2A3 A2A section at -1350 No plan -
A2A4 A2A section at -240-350° No plan -
A2A5 A2A section at -900 Thin area of clad at 900 (met), see Table 5.3 1 Axial crack at 10°; 4 (2 open A2A6 A2A section at -3500-700 circ. cracks at 20-450, see 4 cracks, 2 Table 5.4 through Table 5.10 mounts)
A2A section, contains Clad cracks; undercut regions, 6 4 (2 open A
lA7 exposed clad see Tables 5.11 and 5.12 cmounts)
A2A8 A2A8 A2A section, contains cavity nose No plan --
A2B l Trimmed comer No plan
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BWXT SERVICES, INC. 1140-025-02-24 100 p
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p p
p Figure 5.1: Sectioning of Piece A2A (lower portion of cavity) showing new sample identifications.
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BWXT SERVICES, INC. 1140-025-02-24 101 Table 5.2: Sample identification listing for Piece A2A2.
PieceID Location Test Plan ] Met [SEM p
A2A2A Upper portion of nozzle #3 No plan bore A2A2131 Lower portion of nozzle 0 #3 No plan- - p bore at -190 A2A2B2 Lower portion of nozzle #3 No plan bore -1800 p
Contains axial cracks at 180° Met sample A2A2133 in the J-groove weld. through axial 1 in the J-groove weld. cracks p
A2A2B4 Lower portion or nozzle #3 No plan bore at -1700 p
p p
p p
p p
p Figure 5.2: Sectioning of Piece A2A2, looking at the ID of the J-groove weld bore.
p p
BWXT SERVICES, INC. 1140-025-02-24 102 Table 5.3: Sample identification listing for Piece A2A5.
Piece ID F Location Test Plan Met SEM A2A5A RV head near 900 No plan -- --
A2A5B RV head clad near 900 No plan -- --
Thin region of Met sample A2A5C clad near 900 through 1 --
l 2AlC thin region A2A5D J-groove weld and No plan -- --
clad near 900 p
p p
p p
p p
Figure 5.3: Piece A2A5 before sectioning. The mounted surface is indicated.
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