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Category:Report
MONTHYEARL-23-214, Submittal of Relief Request for Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements2024-06-0505 June 2024 Submittal of Relief Request for Impractical American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Examination Requirements L-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 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 2024-06-05
[Table view] Category:Miscellaneous
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 IR 05000346/20210902021-12-16016 December 2021 Reissue Davis-Besse NRC Inspection Report (05000346/2021090) Preliminary White Finding 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.... 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 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) 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-14-259, 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 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 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 ML13340A1592013-11-26026 November 2013 Davis-Besse Nuclear Power Station Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report Revision 1, Appendix a 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 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) 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 L-13-157, Generic Safety Issue 191 Resolution Plan2013-05-15015 May 2013 Generic Safety Issue 191 Resolution Plan 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 ML13008A0312012-08-10010 August 2012 Davis-Besse Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Cover Through Appendix a, Resumes and Qualifications, Page A-39 ML13135A2462012-08-10010 August 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix C - Area Walk-by Checklists (Awcs) Through End 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 L-12-283, Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B, Seismic Walk-Down Checklists, Sheet 364 of 461 Through Appendix C, Area Walk-By Checklists, Sheet 20 of 1392012-08-10010 August 2012 Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B, Seismic Walk-Down Checklists, Sheet 364 of 461 Through Appendix C, Area Walk-By Checklists, Sheet 20 of 139 ML13008A0642012-08-10010 August 2012 Davis-Besse Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B, Seismic Walk-Down Checklists, Sheet 244 of 461 Through Sheet 363 of 461 ML13008A0632012-08-10010 August 2012 Davis-Besse Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B, Seismic Walk-Down Checklists, Sheet 121 of 461 Through Sheet 243 of 461 ML13008A0622012-08-10010 August 2012 Davis-Besse Near-Term Task Force Recommendation 2.3 Seismic Walkdown Report, Appendix B, Seismic Walk-Down Checklists, Page B-1 Through Sheet 120 of 416 ML12213A3372012-07-27027 July 2012 SB-00-20: CR-2011-03346 Evaluation ML12213A3352012-07-27027 July 2012 Review of Technical Assessment Report No. 25539-200 COR-0000-00001 L-12-196, FENOC-Davis-Besse Nuclear Power Station, Unit 1 Docket No. 50-346, License No. NPF-3 Submittal of Contractor Root Cause Assessment Report-Section 72012-05-14014 May 2012 FENOC-Davis-Besse Nuclear Power Station, Unit 1 Docket No. 50-346, License No. NPF-3 Submittal of Contractor Root Cause Assessment Report-Section 7 ML12138A0582012-05-14014 May 2012 FENOC-Davis-Besse Nuclear Power Station, Unit 1 Docket No. 50-346, License No. NPF-3 Submittal of Contractor Root Cause Assessment Report-Section 3 ML12138A0812012-05-14014 May 2012 FENOC-Davis-Besse Nuclear Power Station, Unit 1 Docket No. 50-346, License No. NPF-3 Submittal of Contractor Root Cause Assessment Report-Section 6 ML12138A0492012-05-14014 May 2012 FENOC-Davis-Besse Nuclear Power Station, Unit 1 Docket No. 50-346, License No. NPF-3 Submittal of Contractor Root Cause Assessment Report-Section 2 ML12138A0672012-05-14014 May 2012 FENOC-Davis-Besse Nuclear Power Station, Unit 1 Docket No. 50-346, License No. NPF-3 Submittal of Contractor Root Cause Assessment Report-Section 4 ML12138A0732012-05-14014 May 2012 FENOC-Davis-Besse Nuclear Power Station, Unit 1 Docket No. 50-346, License No. NPF-3 Submittal of Contractor Root Cause Assessment Report-Section 5 L-12-071, 10 CFR 50.46 Report of Significant Changes or Errors in the Loss of Coolant Accident Evaluation Model2012-03-16016 March 2012 10 CFR 50.46 Report of Significant Changes or Errors in the Loss of Coolant Accident Evaluation Model ML12200A2122011-11-21021 November 2011 Containment Shield Building Issue ML12200A2102011-11-17017 November 2011 Containment System Primary Steel Containment an Shield Building ML12213A3482011-10-28028 October 2011 Table of Contents to Isolated Crack Indication Identified by Impulse Response Testing of the Shield Building L-11-319, Pressure and Temperature Limits Report2011-10-27027 October 2011 Pressure and Temperature Limits Report ML11301A2302011-06-0101 June 2011 Reference: State Energy Consumption Estimates 1960 Through 2009 ML1101007582011-01-10010 January 2011 Pressure Temperature Limits Report Attachment ML11301A2232010-07-31031 July 2010 Chp Regional Application Centers: Activities and Selected Results for Fiscal Year 2009 ML1015201162010-05-26026 May 2010 Field Report Control Rod Drive Mechanism Number 4 Ultrasonic Results ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants ML1005396232010-02-22022 February 2010 Lessons Learned Task Force Action Plan Regarding Stress Corrosion Cracking Dec 2009 (Final Update) L-09-068, Independent Spent Fuel Storage Installation - 10 CFR 72.48 Report of Dry Fuel Storage Facility Changes, Tests, and Experiments2009-02-27027 February 2009 Independent Spent Fuel Storage Installation - 10 CFR 72.48 Report of Dry Fuel Storage Facility Changes, Tests, and Experiments 2023-08-07
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Text
32 EFPY PTLR Page 1 of 9
FIRSTENERGY NUCLEAR OPERATING COMPANY DAVIS-BESSE UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT FOR THE EARLIER OF 32 EFFECTIVE FULL POWER YEARS OR APRIL 22, 2017
Prepared by: Dennis Blakely ___________________
Reviewed by: Kevin Burnworth ____________________
Approved by: Kevin Zellers ______________________
32 EFPY PTLR Page 2 of 9
FirstEnergy Nuclear Operating Company Davis-Besse Unit 1 32 Effective Full Power Years Pressure and Temperature Limit Report
1.0 Introduction This Pressure and Temperature Limit Report (PTLR) provides the information required by Davis-Besse Nuclear Power Station (DBNPS) Technical Specification 5.6.4 to ensure that the Reactor Coolant System Pressure Boundary is operated in accordance with its design.
The PTLR provides the RCS Operating Limits in Section 2.0, which satisfies Technical Specification 5.6.4.a. The Analytical Methods used to develop the limits, including determination of the vessel neutron fluence, are provided in Section 3.0, fulfilling Technical Specification 5.6.4.b. The information and formatting of Section 3 follows the guidance of Attachment 1 to Generic Letter 96-03. The PTLR requirements are provided in Section 4.0 of the report, fulfilling Technical Specification 5.6.4.c.
Revisions to the PTLR are to be submitted to the NRC after issuance.
2.0 RCS Pressure and Temperature Limits
- a. The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines and ramp rates shown on Figures 1, 2, and 3 (Reference 5.7) during heatup, cooldown, criticality, and in-service leak and hydrostatic (ISLH) testing with:
- 1. A maximum heatup of 50°F in any one hour period, and
- 2. A maximum cooldown of 100°F in any one hour period with a cold leg temperature of > 270°F and a maximum cooldown of 50°F in any one hour period with a cold leg temperature of < 270°F.
- b. During periods of low temperature operation (T avg <280 °F), Technical Specification 3.4.12 (Reference 5.3) provides additional requirements for RCS pressure and temperature limits. Those limits are maintained in the Technical Specifications because they are not determined using methods generically approved by the NRC.
32 EFPY PTLR Page 3 of 9 Figure 1: Composite Normal Heatup/Cooldown Limit - Hot Leg "A" Pressure Tap 020040060080010001200140016001800200022002400260050100150200250300350400450Pressure, psigHeatup/Cooldown LimitCriticality Limit Temperature, °F Notes: 1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.
- 2. Allowable cooldown rate at or above 270 °F is 100 °F/hr (Ramp), limited by a 15 °F step change followed by a 9-minute hold.
- 3. Allowable cooldown rate below 270 °F is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.
- 4. A maximum step temperature change of 15 °F is allowable when removing all RC pumps from operation with the DHR system operating. The step temperature change is defined as RC temp minus the DHR return temps to the reactor coolant system prior to stopping all pumps. 5. When the decay heat removal system (DH) is operating without any RC pumps operating, indicated DH return temperature to the reactor vessel shall be used.
- 6. The acceptable pressure and temperature combinations are below and to the right of the limit curve.
- 7. Instrument error is not accounted for in these limits.
A B C D E F G J K H I N O P L M Q R U V S T 32 EFPY PTLR Page 4 of 9 Figure 2: Composite Normal Heatup/Cooldown Limit - Hot Leg "B" Pressure Tap 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 260050100150200250300350400450Pressure, psigHeatup/Cooldown LimitCriticality Limit Temperature, °FNotes: 1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.
- 2. Allowable cooldown rate at or above 270 °F is 100 °F/hr (Ramp), limited by a 15 °F step change followed by a 9-minute hold.
- 3. Allowable cooldown rate below 270 °F is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.
- 4. A maximum step temperature change of 15 °F is allowable when removing all RC pumps from operation with the DHR system operating. The step temperature change is defined as RC temp minus the DHR return temps to the reactor coolant system prior to stopping all pumps. 5. When the decay heat removal system (DH) is operating without any RC pumps operating, indicated DH return temperature to the reactor vessel shall be used.
- 6. The acceptable pressure and temperature combinations are below and to the right of the limit curve.
- 7. Instrument error is not accounted for in these limits.
A B C D E F G J K N O P H I L M Q R U V S T 32 EFPY PTLR Page 5 of 9 Figure 3 Reactor Coolant System Pressure-Temperature Heatup and Cooldown Limits for In-Service Leak and Hydrostatic Tests 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 2400 260050100150200250300350400Pressure, psig"A" Tap Press"B" Tap Press Temperature, °F Notes: 1. Allowable heatup rate is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.
- 2. Allowable cooldown rate at or above 270 °F is 100 °F/hr (Ramp), limited by a 15 °F step change followed by a 9-minute hold.
- 3. Allowable cooldown rate below 270 °F is 50 °F/hr (Ramp), limited by a 15 °F step change followed by an 18-minute hold.
- 4. A maximum step temperature change of 15 °F is allowable when removing all RC pumps from operation with the DHR system operating. The step temperature change is defined as RC temp minus the DHR return temps to the reactor coolant system prior to stopping all pumps.
- 5. When the decay heat removal system (DH) is operating without any RC pumps operating, indicated DH return temperature to the reactor vessel shall be used.
- 6. The acceptable pressure and temperature combinations are below and to the right of the limit curve.
- 7. Instrument error is not accounted for in these limits.
A B C D E F 32 EFPY PTLR Page 6 of 9 3.0 Analytical Methods 3.1 The limits provided in Section 2 and Figures 1, 2, and 3 are valid until the Reactor Vessel has accumulated 32 Effective Full Power Years (EFPY) of fast (E> 1 MeV) neutron fluence or April 22, 2017, whichever comes first.
3.2 The neutron fluence is calculated (Reference 5.12 with Reference 5.13) consistent with Regulatory Guide 1.190 using the NRC-approved methodology described in BAW-2241P-A (Reference 5.5). Table 1 provides the neutron fluence values used in the adjusted reference calculations. The listed fluence values are based on 52 EFPY of operation. The limits in Section 2 are administratively limited as described in Section 3.1 based on the current Operating License of Davis-Besse Nuclear Power Station.
3.3 The Davis-Besse Reactor Vessel Material Surveillance Program complies with the requirements of Appendix H to 10 CFR 50 and is described in BAW-1543A (Reference 5.6). This information was approved by the NRC in the SER of Amendment 199 (Reference 5.1). The specimen capsule withdrawal schedule is contained within the supplements of the topical report. All plant specific specimen capsules have been withdrawn from the reactor vessel. The ART values were not calculated using surveillance data (Reference 5.14) since it was determined to be non-credible.
3.4 Low Temperature Overpressure Protection (LTOP) limits are addressed in Section 2.b, above, and Technical Specification 3.4.12 (Reference 5.3). Reference 5.7, which meets the requirements of ASME Section XI, Appendix G, discusses the methods used to determine the temperature at which LTOP must be active. The pressure limit was determined using ASME Section XI, Appendix G (Reference 5.9), as modified by the alternative rules provided in ASME Code Case N-588 and ASME Code Case N-640.
3.5 Table 1 provides the Adjusted Reference Temperature (ART) for each reactor beltline material. The ART values were calculated in accordance with Regulatory Guide 1.99, Revision 2. For welds in the reactor beltline region, the initial RT NDT values used (in part) to determine ART was calculated using an alternate methodology described in the NRC-approved BAW-2308, Revisions 1-A and 2-A (Reference 5.10). As stated in the NRC Safety Evaluations for the BAW-2308 topical reports, an exemption request to use the alternate initial RT NDT values must be submitted to the NRC. The exemption was granted, and the limits and conditions for using the methodology were confirmed by the NRC to be satisfied in the SER of Amendment ??? (Reference 5.8).
3.6 The Pressure-Temperature (P/T) limits of Section 2 and Figures 1, 2, and 3 (with applicability as stated in 3.1) were generated consistent with the requirements of 10 CFR 50 Appendix G and Regulatory Guide 1.99, Revision 2, using the methods described in BAW-10046A (Reference 5.4) and ASME Section XI, 32 EFPY PTLR Page 7 of 9 Appendix G (Reference 5.9), as modified by the alternative rules provided in ASME Code Case N-588 and ASME Code Case N-640.
3.6.1 The NRC has reviewed the methods described in BAW-10046A (Reference 5.4) and approved the topical report by issuance of a Safety Evaluation Report (SER) dated April 30, 1986. Section 1.2 of BAW-10046A states that it is applicable to all current B&W nuclear steam systems. 3.6.2 ASME Code Cases N-640 and N-588 have been incorporated into ASME Section XI, Appendix G, 2003 Addenda, which are the edition and addenda codified in 10 CFR 50.55a (effective May 27, 2008) and thus may be used per NRC Regulatory Issue Summary (RIS) 2004-04. Specific approval for application at DB NPS is included in Ref. 5.8.
3.7 The minimum temperature requirements of 10CFR50, Appendix G are included on Figures 1 and 2. Figure 3 provides the In-Service Leak and Hydrostatic (ISLH) Test Limits. These limits were calculated in accordance with the requirements of 10CFR50, Appendix G and ASME Code Section XI, Appendix G, 1995 Edition, with Addenda through 1996 and ASME Code Cases N-588 and N-640. 3.8 Davis-Besse has removed more than two surveillance capsules. The capsule test results have been evaluated and found to be non-credible (Reference 5.14). Consequently, ART calculations are not based on the surveillance data. The Adjusted ART - Predicted ART data scatter was less than 2, so the Regulatory Guide 1.99, Rev. 2 Chemistry Table values used in the ART calculations are conservative.
4.0 PTLR Requirements 4.1 The PTLR has been prepared in accordance with the requirements of Technical Specification 5.6.4 (see Reference 5.11). The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto. Davis-Besse will continue to meet the requirements of 10 CFR 50, Appendix G, and any changes to the Davis-Besse P/T limits will be generated in accordance with the NRC approved methodologies described in TS 5.6.4.
32 EFPY PTLR Page 8 of 9 Table1: Davis-Besse Nuclear Power Station Reactor Vessel Beltline Region Data (Applicable as noted in Section 3.1)
Reactor Vessel Location Material Identification Fluence @ 52 EFPY (Wetted Surface) (n/cm 2) (E> 1 MeV) ART @ 1/4 T (°F) @52 EFPY (Note1) ART @ 3/4 T (°F) @52 EFPY (Note1) Limiting Mat'l? (Yes/No) RT PTS (°F) (Note 2) Nozzle Belt Forging ADB 203 2.29E+18 74.8 64.8 No 81.2 Nozzle Belt to Upper Shell Weld (ID 9%) WF-232 2.29E+18 Note 3 Note 3 No 118.3 Nozzle Belt to Upper Shell Weld (OD 91%) WF-233 2.29E+18 100.4* 67.8* No Note 4 Upper Shell Forging AKJ 233 1.69E+19 71.8 57.3 No 79.4 Upper Shell to Lower Shell Weld WF-182-1 1.69E+19 156.2* 106.4* Yes 182.2* Lower Shell Forging BCC 241 1.70E+19 89.9 78.8 Yes 95.7 Note 1: Reported ART values are based on Regulatory Guide 1.99, Revision 2 (Ref. 5.15). P/T Limit calculation was based on a temperature value which is more conservative than the listed ART value. (Ref. 5.13) Note 2: Values from Ref. 5.16, which are based on the location specific clad to vessel interface fluence at 52 EFPY. Note 3: This weld material does not extend out to the 1/4T or 3/4T location. Note 4: This weld material is not present at the clad to vessel interface, so RT PTS does not apply to it.
- Based on the initial RT NDT provided in BAW-2308, Rev. 1A and 2A with NRC Safety Evaluation Reports (Ref. 5.8)
32 EFPY PTLR Page 9 of 9 5.0 References 5.1 Safety Evaluation by the NRC Office of Nuclear Reactor Regulation Related to Amendment No. 199 to Facility Operating License No. NPF-3 Davis-Besse Nuclear Power Station, Unit No. 1, attached to correspondence dated July 20, 1995. 5.2 Technical Specification 5.6.4, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)."
5.3 Technical Specification 3.4.12, "Low Temperature Overpressure Protection."
5.4 BAW-10046A, Revision 2 "Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50 Appendix G."
5.5 BAW-2241P-A, "Fluence and Uncertainty Methodologies," dated April 1999.
5.6 BAW-1543A, "Master Integrated Reactor Vessel Material Surveillance Program."
5.7 ANP-2718, Revision 3, "Appendix G Pressure-Temperature Limits for 52 EFPY, Using ASME Code Cases for Davis-Besse Nuclear Power Station," dated August 2010. 5.8 Safety Evaluation by the NRC Office of Nuclear Reactor Regulation Related to License Amendment Request 08-034 to Facility Operating License No. NPF-3 Davis-Besse Nuclear Powe r Station, Unit No. 1.
5.9 ASME Code Section XI, Appendix G, as modified by the alternate rules provided in ASME Code Case N-640 and ASME Code Case N-588. ASME Code Cases N-640 and N-588 have been incorporated into ASME Section XI, Appendix G, 2003 Addenda, which are the edition and addenda codified in 10 CFR 50.55a (effective May 27, 2008).
5.10 BAW-2308, Revision 1-A and Revision 2-A, "Initial RT NDT of Linde 80 Weld Materials," dated August 2005 (1-A) and March 2008 (2-A).
5.11 Calculation C-NSA-064.02-037, Revision 0, "Davis-Besse 52 EFPY PT Limits - Midland RV Closure Head," dated ????
5.12 AREVA Report 86-9015129-000, "DB1 - Cycles 13-15 Fluence Analysis Report," dated 4/21/2006.
5.13 AREVA Report 51-9123331-000, "Davis-Besse - EOL Fluence Reconciliation,"
dated 10/8/2009.
5.14 AREVA Document 32-9031157-000, "Davis-Besse Revised ART Values at 52 EFPY," dated 9/20/2006.
5.15 AREVA Document 32-9017744-003, "Davis-Besse ART Values at 52 EFPY,"
dated 10/29/2009.
5.16 AREVA Document 32-9123247-000, "RTPTS Values of Davis-Besse Unit 1 for 52 EFPY, Including Extended Beltline," dated 11/12/09.