Letter Sequence Other |
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Initiation
- Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request
- Acceptance...
- Supplement, Supplement, Supplement, Supplement, Supplement, Supplement
Results
Other: 3F0112-11, Corrections to the Extended Power Uprate LAR 309 Vendor Affidavit and Technical Report, 3F0113-08, Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Re, 3F0712-03, Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip., 3F0912-01, ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C, 3F1011-08, ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip, 3F1112-02, 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 250, ML11259A012, ML11300A226, ML11342A195, ML12031A032, ML12053A231, ML12107A068, ML12202A734, ML12205A357, ML12205A358, ML12240A119, ML12255A409, ML12262A385, ML12314A391, ML12314A392, ML12314A393, ML13032A540
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MONTHYEAR3F0811-04, Response to Request for Additional Information to Support NRC Probabilistic Risk Assessment Licensing Branch Acceptance Review of the CR-3 Extended Power Uprate LAR2011-08-11011 August 2011 Response to Request for Additional Information to Support NRC Probabilistic Risk Assessment Licensing Branch Acceptance Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0811-02, Response to Request for Additional Information to Support NRC Balance of Plant Branch Acceptance Review of the CR-3 Extended Power Uprate LAR2011-08-11011 August 2011 Response to Request for Additional Information to Support NRC Balance of Plant Branch Acceptance Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0811-01, Response to Request for Additional Information to Support NRC Instrumentation and Controls Branch Acceptance Review of the CR-3 Extended Power Uprate LAR2011-08-18018 August 2011 Response to Request for Additional Information to Support NRC Instrumentation and Controls Branch Acceptance Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0811-03, Response to Request for Additional Information to Support NRC Mechanical and Civil Branch Acceptance Review of the CR-3 Extended Power Uprate LAR2011-08-25025 August 2011 Response to Request for Additional Information to Support NRC Mechanical and Civil Branch Acceptance Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI ML11259A0122011-09-14014 September 2011 NRR E-mail Capture - Clarification for August 25 Emcb RAI Acceptance Review Letter Project stage: Other ML11286A0922011-10-11011 October 2011 Response to Request for Additional Information to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI ML11300A2262011-10-25025 October 2011 Feedwater Line Break Overpressure Protection Analysis to Support NRC Reactor Systems Branch Acceptance Review of the CR-3 Extended Power Update LAR and LAR Approval Schedule Project stage: Other 3F1011-08, ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip2011-10-25025 October 2011 ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip Project stage: Other ML11326A2312011-12-0707 December 2011 Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12205A3572011-12-15015 December 2011 Attachment a to 3F0712-03, CR-3 LOCA Summary Report - EPU/ROTSG/Mark-B-HTP, Revision 4 Project stage: Other 3F1211-05, Response to Request for Additional Information to Support NRC Health Physics and Human Performance Branch Technical Review of the CR-3 Extended Power Uprate LAR2011-12-15015 December 2011 Response to Request for Additional Information to Support NRC Health Physics and Human Performance Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F1211-06, Response to Request for Additional Information to Support NRC Vessels and Internals Integrity Branch Technical Review of the CR-3 Extended Power Uprate LAR2011-12-15015 December 2011 Response to Request for Additional Information to Support NRC Vessels and Internals Integrity Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F1211-10, Response to Request for Additional Information to Support NRC Fire Protection Branch Technical Review of the CR-3 Extended Power Uprate LAR2011-12-21021 December 2011 Response to Request for Additional Information to Support NRC Fire Protection Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0112-04, Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR2012-01-0505 January 2012 Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI ML12030A2092012-01-0505 January 2012 Response to Request for Additional Information to Support NRC Instrumentation and Control Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI ML11342A1952012-01-0606 January 2012 Notice of Consideration of Issuance of Amendment to Facility Operating License and Opportunity for a Hearing and Order Imposing Procedures for Document Access to Sensitive Unclassified Non-safeguards Information Project stage: Other ML12024A3002012-01-19019 January 2012 Response to Request for Additional Information to Support NRC Component Performance and Testing Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0112-06, Response to Request for Additional Information to Support NRC Piping and NDE Branch Technical Review of the CR-3 Extended Power Uprate LAR2012-01-19019 January 2012 Response to Request for Additional Information to Support NRC Piping and NDE Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0112-11, Corrections to the Extended Power Uprate LAR 309 Vendor Affidavit and Technical Report2012-01-31031 January 2012 Corrections to the Extended Power Uprate LAR #309 Vendor Affidavit and Technical Report Project stage: Other ML12003A2172012-02-0808 February 2012 Nuclear Generating Plant - Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12052A1302012-03-0202 March 2012 Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12053A2312012-03-0202 March 2012 Request for Withholding Proprietary Information from Public Disclosure Related Extended Power Uprate Technical Report Project stage: Other 3F0312-02, Response to Second Request for Additional Information to Support NRC Instrumentation and Controls Branch (Eicb) Technical Review of the CR-3 Extended Power Uprate LAR2012-03-19019 March 2012 Response to Second Request for Additional Information to Support NRC Instrumentation and Controls Branch (Eicb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Request 3F0312-01, Response to Request for Additional Information to Support NRC PRA Licensing Branch Technical Review of the CR-3 Extended Power Uprate LAR2012-03-22022 March 2012 Response to Request for Additional Information to Support NRC PRA Licensing Branch Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI ML12031A0322012-03-26026 March 2012 Request for Withholding Proprietary Information from Public Disclosure Related to Extended Power Uprate Technical Report Project stage: Other ML12107A0682012-04-0404 April 2012 NRR E-mail Capture - Crystal River, Unit 3 EPU LAR - Clarifications for Component Performance & Testing Branch (Eptb) Project stage: Other 3F0412-05, Response to Second Request for Additional Information to Support NRC Health Physics and Human Performance Branch (Ahpb) Technical Review of the CR-3 Extended Power Uprate LAR2012-04-0404 April 2012 Response to Second Request for Additional Information to Support NRC Health Physics and Human Performance Branch (Ahpb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Request 3F0412-04, Response to Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch (Emcb) Technical Review of the CR-3 Extended Power Uprate LAR2012-04-0404 April 2012 Response to Request for Additional Information to Support NRC Mechanical and Civil Engineering Branch (Emcb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0412-07, Response to Request for Additional Information to Support NRC Containment and Ventilation Branch (Scvb) Technical Review of the CR-3 Extended Power Uprate LAR2012-04-12012 April 2012 Response to Request for Additional Information to Support NRC Containment and Ventilation Branch (Scvb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0412-06, Response to Request for Additional Information to Support NRC Electrical Systems Branch (Eeeb) Technical Review of the CR-3 Extended Power Uprate LAR2012-04-16016 April 2012 Response to Request for Additional Information to Support NRC Electrical Systems Branch (Eeeb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0412-09, Response to Request for Additional Information to Support NRC Accident Dose Branch (Aadb) Technical Review of the CR-3 Extended Power Uprate LAR2012-04-26026 April 2012 Response to Request for Additional Information to Support NRC Accident Dose Branch (Aadb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI ML12102A0442012-05-0404 May 2012 Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12314A3912012-05-31031 May 2012 17877-0002-100, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment 2 Project stage: Other ML12205A3582012-05-31031 May 2012 Attachment D to 3F0712-03, Technical Report, ANP-3114(NP), Rev. 0, CR-3 EPU - Feedwater Line Break Analysis Sensitivity Studies. Project stage: Other 3F0612-05, Response to Request for Additional Information to Support NRC Nuclear Performance and Code Review Branch (Snpb) Technical Review of the CR-3 Extended Power Uprate LAR2012-06-18018 June 2012 Response to Request for Additional Information to Support NRC Nuclear Performance and Code Review Branch (Snpb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Response to RAI 3F0712-03, Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip.2012-06-30030 June 2012 Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip. Project stage: Other ML12184A3442012-07-0202 July 2012 NRR E-mail Capture - Crystal River, Unit 3 EPU LAR - Discussion/Clarification of Emcb Draft RAIs Project stage: Draft Other ML12174A2922012-07-0505 July 2012 Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12171A3472012-07-0505 July 2012 Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12205A3562012-07-17017 July 2012 Supplement to the Extended Power Uprate License Amendment Request #309 Project stage: Supplement 3F0712-06, Response to Second Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch (Esgb) Technical Review of the CR-3 Extended Power Uprate LAR2012-07-17017 July 2012 Response to Second Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch (Esgb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Request ML12194A4172012-07-17017 July 2012 Generating Plant - Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI 3F0712-07, Response to Third Request for Additional Information to Support NRC Health Physics and Human Performance Branch (Ahpb) Technical Review of the CR-3 Extended Power Uprate LAR2012-07-31031 July 2012 Response to Third Request for Additional Information to Support NRC Health Physics and Human Performance Branch (Ahpb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Request 3F0712-08, Response to Second Request for Additional Information to Support NRC PRA Licensing Branch (Apla) Technical Review of the CR-3 Extended Power Uprate LAR2012-07-31031 July 2012 Response to Second Request for Additional Information to Support NRC PRA Licensing Branch (Apla) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Request ML12125A1622012-08-0101 August 2012 Nuclear Generating Plant - Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12202A7342012-08-0101 August 2012 Request for Withholding Proprietary Information from Public Disclosure Related to Extended Power Uprate Technical Report Project stage: Other ML12202A0602012-08-0202 August 2012 Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI ML12213A3032012-08-0303 August 2012 Request for Additional Information for Extended Power Uprate License Amendment Request (Srxb) Project stage: RAI ML12219A1912012-08-16016 August 2012 Request for Additional Information for Extended Power Uprate License Amendment Request Project stage: RAI 3F0812-02, Response to Second Request for Additional Information to Support NRC Electrical Engineering Branch (Eeeb) Technical Review of the CR-3 Extended Power Uprate LAR2012-08-21021 August 2012 Response to Second Request for Additional Information to Support NRC Electrical Engineering Branch (Eeeb) Technical Review of the CR-3 Extended Power Uprate LAR Project stage: Request 2012-03-26
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Category:Report
MONTHYEARML24090A0072024-03-31031 March 2024 Enclosures 10 - 17: Crystal River, Unit 3, Response to Request for Additional Information ML24090A0042024-03-29029 March 2024 Enclosure 6, Part 1: Crystal River, Unit 3 - Appendix E – CHAR-05 Site Characterization of Buildings ML24030A7472024-02-13013 February 2024 Audit Report Attachment - Crystal River Unit 3 Nuclear Generating Plant LTP ML24030A7482024-02-12012 February 2024 Audit Report Cover Letter and Report - Crystal River Unit 3 Nuclear Generating Plant LTP ML23345A1882023-12-0606 December 2023 Fws to NRC Crystal River Species List of Threatened and Endangered Species That May Occur in Your Proposed Project Location or May Be Affected by Your Proposed Project 3F0324-03, Enclosure 6, Part 2: Crystal River, Unit 3, DD Survey 23-10-00962023-10-10010 October 2023 Enclosure 6, Part 2: Crystal River, Unit 3, DD Survey 23-10-0096 ML23160A2962023-06-0909 June 2023 Response to Crystal River, Unit 3 Supplemental Information Needed for Acceptance on the Application for a License Amendment Regarding Approval of the License Termination Plan ML23160A2972023-06-0909 June 2023 CR3 Site Characterization Survey Report (CHAR-01) Impacted Open Land Survey Areas 3F0623-02, Maintenance Support Building2023-06-0909 June 2023 Maintenance Support Building ML23160A2982023-06-0909 June 2023 Site Characterization Surveys ML23107A2732023-06-0707 June 2023 Orise Independent Survey Report Dcn 5366-SR-01-0 3F0522-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20222022-05-17017 May 2022 Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2022 3F0520-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20202020-05-18018 May 2020 Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2020 3F0320-01, NRC Commitment Change Report - March 20202020-03-17017 March 2020 NRC Commitment Change Report - March 2020 ML19343A8252019-12-0606 December 2019 Letter from Erika Bailey, Oak Ridge Institute for Science and Education, to John Hickman, NRC, Forwarding Independent Confirmatory Survey Summary and Results for the 3,854-Acre Area Partial Site Release at the Crystal River Energy Complex ML19022A0762019-01-22022 January 2019 Partial Site Release Request ML19029A0092018-11-0707 November 2018 Reference 16 - Defueled Safety Analysis Report DSAR-R002 ML18303A2942018-06-21021 June 2018 Golder Associates, Inc. - Citrus Combined Cycle Project - CFR 122.21(r) Report 3F0518-03, Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20182018-05-24024 May 2018 Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2018 ML16176A3392016-10-28028 October 2016 Decommissioning Lessons Learned Report and Transmittal Memorandum ML19029A0102016-06-28028 June 2016 Reference 3 - Crystal River, Unit 3, Historical Site Assessment Rev. 00 3F0616-02, Nrg Commitment Change Report - June 20162016-06-14014 June 2016 Nrg Commitment Change Report - June 2016 ML13343A1782013-12-31031 December 2013 Report P23-1680-001, Rev. 0, Site-Specific Decommissioning Cost Estimate for Crystal River Unit 3 Nuclear Generating Plant. 3F0113-08, Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Re2013-01-31031 January 2013 Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Res 3F1112-01, Alion Technical Report ALION-PLN-ENER-8706-02, Rev. 0, Crystal River 3: Bypass Fiber Quantity Test Plan2012-11-0707 November 2012 Alion Technical Report ALION-PLN-ENER-8706-02, Rev. 0, Crystal River 3: Bypass Fiber Quantity Test Plan 3F1112-02, 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 2502012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 250 ML12314A3932012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 81 of 250 Through Page 173 of 250 ML12314A3922012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 1 of 250 Through Page 80 of 250 3F0912-01, ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C2012-09-30030 September 2012 ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C 3F0712-03, Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip.2012-06-30030 June 2012 Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip. ML12314A3912012-05-31031 May 2012 17877-0002-100, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment 2 ML12205A3582012-05-31031 May 2012 Attachment D to 3F0712-03, Technical Report, ANP-3114(NP), Rev. 0, CR-3 EPU - Feedwater Line Break Analysis Sensitivity Studies. ML12284A1382012-05-25025 May 2012 Report EGS-TR-HC589-01, Seismic Qualification Test Report for Structural Verification Testing of Iccms Cabinet Assembly. 3F0512-01, NRC Commitment Change Report - May 20122012-05-14014 May 2012 NRC Commitment Change Report - May 2012 3F0112-04, Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR2012-01-0505 January 2012 Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR ML12205A3572011-12-15015 December 2011 Attachment a to 3F0712-03, CR-3 LOCA Summary Report - EPU/ROTSG/Mark-B-HTP, Revision 4 3F1211-14, Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis2011-12-14014 December 2011 Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis 3F1011-08, ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip2011-10-25025 October 2011 ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip ML11237A0682011-08-0505 August 2011 Siemens Technical Report CT-27438, Missile Probability Analysis Report Progress Energy Crystal River 3, Revision 1A ML11207A4442011-06-15015 June 2011 Attachment 7- Crystal River Unit 3 Extended Power Uprate Technical Report 3F0511-02, Response to Request for Additional Information Required for the Development of the Confirmatory LOCA and Non-LOCA Models2011-05-0606 May 2011 Response to Request for Additional Information Required for the Development of the Confirmatory LOCA and Non-LOCA Models ML1101906672010-10-0404 October 2010 Levy, Units 1 and 2, Cola (Sensitive Material), Rev. 2 - Levy County Emergency Plan Part 02 - Draft (Redacted) 3F0910-01, CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analysis2010-09-0808 September 2010 CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analysis ML1019304172010-05-0606 May 2010 Tritium Database Report ML1010603472010-04-0909 April 2010 5.2.2.4.4. Quality Control and Nondestructive Testing ML1028710882010-03-12012 March 2010 7.6 Vibration Due to Cutting Tendons ML1028711112010-02-25025 February 2010 7.11 Added Stress from Pulling Tendons ML1028711102010-02-23023 February 2010 6.3 Thermal Effects of Greasing ML1028804682010-02-19019 February 2010 7.9 Inadequate Hydro Blasting Nozzles Rate Part 3 ML1028711212010-02-19019 February 2010 7.10 Hydrodemolition Induced Cracking 2024-03-31
[Table view] Category:Technical
MONTHYEARML24090A0072024-03-31031 March 2024 Enclosures 10 - 17: Crystal River, Unit 3, Response to Request for Additional Information ML24090A0042024-03-29029 March 2024 Enclosure 6, Part 1: Crystal River, Unit 3 - Appendix E – CHAR-05 Site Characterization of Buildings ML24030A7472024-02-13013 February 2024 Audit Report Attachment - Crystal River Unit 3 Nuclear Generating Plant LTP ML24030A7482024-02-12012 February 2024 Audit Report Cover Letter and Report - Crystal River Unit 3 Nuclear Generating Plant LTP ML23345A1882023-12-0606 December 2023 Fws to NRC Crystal River Species List of Threatened and Endangered Species That May Occur in Your Proposed Project Location or May Be Affected by Your Proposed Project 3F0324-03, Enclosure 6, Part 2: Crystal River, Unit 3, DD Survey 23-10-00962023-10-10010 October 2023 Enclosure 6, Part 2: Crystal River, Unit 3, DD Survey 23-10-0096 3F0623-02, Maintenance Support Building2023-06-0909 June 2023 Maintenance Support Building ML23160A2982023-06-0909 June 2023 Site Characterization Surveys ML23160A2972023-06-0909 June 2023 CR3 Site Characterization Survey Report (CHAR-01) Impacted Open Land Survey Areas ML23160A2962023-06-0909 June 2023 Response to Crystal River, Unit 3 Supplemental Information Needed for Acceptance on the Application for a License Amendment Regarding Approval of the License Termination Plan ML23107A2732023-06-0707 June 2023 Orise Independent Survey Report Dcn 5366-SR-01-0 3F0522-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20222022-05-17017 May 2022 Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2022 3F0520-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20202020-05-18018 May 2020 Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2020 3F0320-01, NRC Commitment Change Report - March 20202020-03-17017 March 2020 NRC Commitment Change Report - March 2020 ML19343A8252019-12-0606 December 2019 Letter from Erika Bailey, Oak Ridge Institute for Science and Education, to John Hickman, NRC, Forwarding Independent Confirmatory Survey Summary and Results for the 3,854-Acre Area Partial Site Release at the Crystal River Energy Complex ML19022A0762019-01-22022 January 2019 Partial Site Release Request ML19029A0092018-11-0707 November 2018 Reference 16 - Defueled Safety Analysis Report DSAR-R002 ML18303A2942018-06-21021 June 2018 Golder Associates, Inc. - Citrus Combined Cycle Project - CFR 122.21(r) Report 3F0518-03, Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20182018-05-24024 May 2018 Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2018 ML19029A0102016-06-28028 June 2016 Reference 3 - Crystal River, Unit 3, Historical Site Assessment Rev. 00 ML13343A1782013-12-31031 December 2013 Report P23-1680-001, Rev. 0, Site-Specific Decommissioning Cost Estimate for Crystal River Unit 3 Nuclear Generating Plant. 3F1112-01, Alion Technical Report ALION-PLN-ENER-8706-02, Rev. 0, Crystal River 3: Bypass Fiber Quantity Test Plan2012-11-0707 November 2012 Alion Technical Report ALION-PLN-ENER-8706-02, Rev. 0, Crystal River 3: Bypass Fiber Quantity Test Plan ML12314A3932012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 81 of 250 Through Page 173 of 250 ML12314A3922012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 1 of 250 Through Page 80 of 250 3F1112-02, 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 2502012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 250 3F0912-01, ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C2012-09-30030 September 2012 ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C 3F0712-03, Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip.2012-06-30030 June 2012 Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip. ML12205A3582012-05-31031 May 2012 Attachment D to 3F0712-03, Technical Report, ANP-3114(NP), Rev. 0, CR-3 EPU - Feedwater Line Break Analysis Sensitivity Studies. ML12314A3912012-05-31031 May 2012 17877-0002-100, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment 2 ML12284A1382012-05-25025 May 2012 Report EGS-TR-HC589-01, Seismic Qualification Test Report for Structural Verification Testing of Iccms Cabinet Assembly. ML12205A3572011-12-15015 December 2011 Attachment a to 3F0712-03, CR-3 LOCA Summary Report - EPU/ROTSG/Mark-B-HTP, Revision 4 3F1011-08, ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip2011-10-25025 October 2011 ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip ML11237A0682011-08-0505 August 2011 Siemens Technical Report CT-27438, Missile Probability Analysis Report Progress Energy Crystal River 3, Revision 1A ML11207A4442011-06-15015 June 2011 Attachment 7- Crystal River Unit 3 Extended Power Uprate Technical Report ML1101906672010-10-0404 October 2010 Levy, Units 1 and 2, Cola (Sensitive Material), Rev. 2 - Levy County Emergency Plan Part 02 - Draft (Redacted) ML1010603472010-04-0909 April 2010 5.2.2.4.4. Quality Control and Nondestructive Testing ML1028711122010-02-19019 February 2010 7.9 Inadequate Hydro Blasting Nozzles Rate Part 1 ML1028804682010-02-19019 February 2010 7.9 Inadequate Hydro Blasting Nozzles Rate Part 3 ML1028711212010-02-19019 February 2010 7.10 Hydrodemolition Induced Cracking ML1028711132010-02-19019 February 2010 7.9 Inadequate Hydro Blasting Nozzles Rate Part 2 ML1028711452010-02-18018 February 2010 7.8 Excessive Water Jet Pressure Part 1 ML1028711462010-02-18018 February 2010 7.8 Excessive Water Jet Pressure Part 2 ML1028804582010-02-18018 February 2010 7.8 Excessive Water Jet Pressure Part 3 ML1028711482010-02-0707 February 2010 7.2 Vibration Induced by Hydro Blasting Part 1 ML1028711492010-02-0707 February 2010 7.2 Vibration Induced by Hydro Blasting Part 2 ML1028702832010-02-0303 February 2010 6.6 Original Tensioning T(2) ML1028702912010-02-0202 February 2010 6.5 Inadequate RetensioningT1 ML1028704062010-01-27027 January 2010 Email - from: Dyksterhouse, Don (Don.Dyksterhouse@Pgnmail.Com) to: Lake, Louis Dated Wednesday, January 27, 2010 Design Basis Calculations Attachments: 0102-0135-02 Concrete Strength and Elastic.... Ro Final.Pdf; 0102-0135-03 Ro Final.Pdf; ML1028706922010-01-22022 January 2010 Ctl 059169 Final Report.Pdf ML1028805162010-01-16016 January 2010 Email - from: Miller, Craig L (Craig.Miller@Pgnmail.Com) to: Lake, Louis; Thomas, George; Carrion, Robert; 'Trowe@Wje.Com'; Sealey, Mac Cc: Williams, Charles R. Dated Saturday, January 16, 2010 1:19PM Subject: Failure Mode 2.6 for Review.. 2024-03-31
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ANP-3052 Revision 0 October 2011 CR-3 EPU Feedwater Line Break.Analysis with Failure of First Safety Grade Trip AREVA Inc.
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ANP-3052 Revision 0 Record of Revision Revision PageslSectionslParagraphs No. Changed Brief Description I Change Authorization 0 All Initial Release 4 4-4- 1-4-
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ANP-3052 Revision 0 Table of Contents 1.0 Introduction ........................................................................................................................................ 7 2.0 Analytical Methodology ............................................................................................................. 8 3.0 Analysis Inputs .................................................................................................................................. 9 4.0 Results / Conclusions ...................................................................................................... 12 5.0 References ...................................................................................................................................... 29 Page 3
ANP-3052 Revision 0 List of Tables Table 3-1: Input to Feedwater Line Break Analysis with First Trip Failed ............................................ 10 Table 4-1: Sequence of Events for FWLB with First Trip Failed .......................................................... 13 Table 4-2: Results for FWLB with First Trip Failed ............................................................................... 13 Page 4
ANP-3052 Revision 0 List of Figures Figure 4-1: FWLB with Fail First Trip - RCS Pressure ............................................................... 14 Figure 4-2: FWLB with Fail First Trip - Reactor Power ........................................................................ 15 Figure 4-3: FW LB with Fail First Trip - Reactivity ................................................................................ 16 Figure 4-4: FWLB with Fail First Trip - Primary System Temperatures ............. ...................... 17 Figure 4-5: FWLB with Fail First Trip - Indicated Pressurizer Level ...................................................... 18 Figure 4-6: FWLB with Fail First Trip - Pressurizer Collapsed Liquid Level ......................................... 19 Figure 4-7: FWLB with Fail First Trip - Pressurizer Surge Line Flow ...................................... 20 Figure 4-8: FWLB with Fail First Trip - RCS Volumetric Flow Rate ............................................. 21 Figure 4-9: FWLB with Fail First Trip - Pressurizer Safety Valve Flow ....................... 22 Figure 4-10: FWLB with Fail First Trip - SG Secondary Side Liquid Level ................................. 23 Figure 4-11: FWLB with Fail First Trip - SG Secondary Side Inventory ........................... 24 Figure 4-12: FWLB with Fail First Trip - SG % Operating Range ............................................. 25 Figure 4-13: FWLB with Fail First Trip - SG Pressure .................... ................................. 26 Figure 4-14: FWLB with Fail First Trip - EFW Flow ..................... ................................... 27 Figure 4-15: FWLB with Fail First Trip - Integrated MSSV Flow .......................................... 28 Page 5
ANP-3052 Revision 0 Nomenclature Acronym Definition CR-3 Crystal River Unit 3 DSS Diverse Scram System EFIC Emergency Feedwater Initiation and Control EFW Emergency Feedwater EPU Extended Power Uprate FWLB Feedwater Line Break LAR Licensing Amendment Request LOOP Loss of Offsite Power MFW Main Feedwater MSSV Main Steam Safety Valves NRC Nuclear Regulatory Commission PORV Pilot Operated Relief Valve PSV Pressurizer Safety Valve RCP Reactor Coolant Pump RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RPS Reactor Protection System SG Steam Generator TSV Turbine Stop Valves Page 6
ANP-3052 Revision 0
1.0 INTRODUCTION
The Nuclear Regulatory Commission (NRC) Reactor Systems Branch staff requested an additional analysis be performed to support the review of the Crystal River Unit 3 (CR-3) extended power uprate (EPU) licensing amendment request (LAR). In particular, the staff requested an analysis of the feedwater line break (FWLB) transient assuming that the first safety grade reactor protection system (RPS) trip function fails to trip the reactor. This report documents the results of the requested analysis.
The analysis documented in this report assumes that the RPS high reactor coolant system (RCS) pressure trip function fails to trip the reactor. This analysis models the non-safety grade diverse scram system (DSS) trip, which inserts the regulating control rod banks upon reaching the DSS high RCS pressure trip setpoint. The peak RCS pressure during the FWLB transient is reported and compared to an acceptance criterion of 120% of the reactor coolant pressure boundary (RCPB) design pressure (1.20
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ANP-3052 Revision 0 2.0 ANALYTICAL METHODOLOGY The thermal-hydraulic analysis of the FWLB analyses at the CR-3 EPU power level with the first safety grade trip failed is performed using the RELAP5/MOD2-B&W computer program (Reference [1]). The code simulates RCS and secondary system operation. The reactor Core model is based on a point kinetics solution with reactivity feedback for control rod assembly insertion, fuel temperature changes, moderator temperature changes, and changes in boron concentration. The RCS model provides for heat transfer from the core, transport of the coolant to the steam generators (SG), and heat transfer to the steam generators. The secondary model includes a detailed depiction of the main steam system, including steam relief to the atmosphere through the main steam safety valves (MSSVs) and simulation of the turbine stop valves (TSVs). The secondary model also includes the delivery of feedwater, both main and emergency, to the steam generators.
The RELAP5/MOD2-B&W code has been approved by the NRC for use in non-LOCA safety analyses (Reference [2]). The analysis documented in this report is consistent with Reference [2] except that the first safety grade trip is not credited. Instead, this analysis credits the next available trip, which is the non-safety grade DSS trip on high RCS pressure.
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ANP-3052 Revision 0 3.0 ANALYSIS INPUTS The FWLB analysis with the first safety grade trip failed uses input that is consistent with the FWLB analysis without pressurizer spray performed to support Section 2.8.5.2.4 of the CR-3 EPU LAR, with a few notable exceptions. The changes include:
- 1. The RPS high RCS pressure trip function is disabled.
- 2. The DSS high RCS pressure trip function is modeled. The DSS trip setpoint is modeled as 2465 psia. The DSS high RCS pressure trip setpoint includes margin to bound possible changes in the containment pressure during a FWLB. The DSS trip delay time is modeled as 1.23 seconds.
Finally, the rod worth available to the DSS system is modeled as only 2.0 %Ak/k, since the DSS system only inserts the regulating control rod groups.
Table 3-1 summarizes all of the key input to the FWLB analysis with the first safety grade trip failed, including the changes highlighted above.
The peak RCS pressure is reached shortly after reactor trip before any active components such as emergency feedwater (EFW) are credited to mitigate the results. Therefore, there are no single failure assumptions that would result in a more limiting peak RCS pressure. However, the FWLB overpressure protection analysis with the first safety grade trip failed assumes the same single failure assumption as the FWLB analysis performed to support Section 2.8.5.2.4 of the CR-3 EPU LAR. The single failure assumption modeled is the failure of one train, of Emergency Feedwater Initiation and Control (EFIC) such that EFW flow is not initiated automatically in one train. Consequently, only one of the two EFW pumps is assumed available to provide flow to the SGs. This single failure assumption produces a conservative long-term transient response after EFW is assumed to reach the SGs.
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ANP-3052 Revision 0 Table 3-1: Input to Feedwater Line Break Analysis with First Trip Failed Parameter Value RCS Conditions Core Power, MWt 3014
- 1.004 = 3026.1 Decay Heat 1.0*ANS71 plus B&W Actinides Total Net Reactor Coolant Pump (RCP) Heat, MWt 16.4 Average RCS Temperature, °F 582 Initial Hot Leg Pressure, psia 2170 Total RCS Flow Rate, gpm 374,880 Pressurizer Initial Indicated Pressurizer Level, in 240 Pressurizer Spray Not Modeled Pressurizer Heaters Not Modeled Pilot Operated Relief Valve (PORV) Not Modeled Pressurizer Safety Valve (PSV) Setpoints, psig 2500 * (1 + 0.03) (open) 2500 * (1 - 0.04) (close)
Total PSV Rated Capacity, Ibm/hr 2
- 317,973 @ 2750 psig Secondary Side Initial Main Feedwater (MFW) Temperature, OF 460 OF Tube Plugging, % 5 Initial SG Level, %OR < 70 EFW Temperature, OF 120 EFW Minimum Required Flow, gpm 550 EFW Delay Time, sec 60 Turbine Trip Delay Time, s 0.0 TSV Stroke Time, s 0.2 Number of Main Steam Safety Valves (MSSVs) per SG 8 MSSV Capacity per SG 7 @ 845,759 Ibm/hr 1 @ 583,574 Ibm/hr MSSV Nominal Setpoints 2 @ 1050 psig 2 @ 1070 psig 2 @ 1090 psig 2 @ 1100 psig including small MSSV MSSV Setpoint Tolerance +3%
MSSV Accumulation +3%
MSSV Blowdown -5%
Core Kinetics Parameters Doppler Temperature Coefficient (Ak/k/°F) -1.30 E-5 Page 10
ANP-3052 Revision 0 Parameter Value Moderator Temperature Coefficient (Ak/°kIF) 0.0 E-4 Prompt Neutron Generation Time, (ps) 24.8 Effective Delayed Neutron Fraction 0.0070 DSS Insertable Rod Worth, %Ak/k 2.000 RPS High RCS Pressure Trip Assumed Failed DSS High RCS Pressure Trip Setpoint, psia 2465 DSS Trip Delay Time, s 1.23 Miscellaneous Offsite Power Available Single Failures One Train of EFIC Operator Actions None Page 11
ANP-3052 Revision 0 4.0 RESULTS I CONCLUSIONS The sequence of events for the FWLB accident with the first safety grade trip failed is listed in Table 4-1 and the calculated results are tabulated in Table 4-2. Plots that demonstrate the transient response following a FWLB are provided in Figures 4-1 through 4-15.
The transient progression is similar to the FWLB transient described in Section 2.8.5.2.4 of the CR-3 EPU LAR. The trip setpoint is reached at 9.025 seconds instead of 8.802 seconds because the DSS high RCS pressure setpoint is higher than the RPS high RCS pressure setpoint. In addition, because the delay time modeled for the DSS trip is longer than the delay time associated with the RPS, control rod insertion does not begin until 10.258 seconds compared to 9.415 seconds in the FWLB analysis in LAR Section 2.8.5.2.4. The longer time to reactor trip contributes to a higher peak RCS pressure. The peak RCS pressure is also higher for the analysis with the first safety grade trip failed because DSS only inserts the regulating control rod banks, and therefore has less inserted rod worth than the analysis modeling the RPS trip function.
The peak RCS pressure (2950.9 psia) occurred in the lower downcomer region of the reactor vessel and did not exceed 120% of the RCPB design pressure of 2500 psig (3000 psig).
As explained in Section 2.8.5.2.4 of the CR-3 EPU LAR, the FWLB event is sufficiently severe that the pressurizer fills., As a result, the PSVs begin to pass single-phase liquid. The PSVs of the type installed at CR-3 achieve satisfactory performance for fluid temperatures greater than -550'F. An additional check is performed to show that the PSV fluid inlet temperature remains greater than 600 OF to ensure that the PSVs operate as intended. Figure 4-4 demonstrates that at all times throughout the FWLB transient, the liquid temperature at the top of the pressurizer remains above 600 OF.
A sensitivity study was performed modeling the FWLB transient with the first safety grade trip failed and a loss of offsite power (LOOP). The LOOP is conservatively considered to occur coincident with the turbine trip that follows reactor trip. The case with a LOOP included the insertion of the safety control rod banks on LOOP. The sensitivity study determined that the peak RCS pressure from a LOOP case (2921.9 psia) is less limiting than the peak RCS pressure without a LOOP. Therefore, the FWLB evaluation using the DSS trip function and no LOOP is the bounding case.
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ANP-3052 Revision 0 Table 4-1: Sequence of Events for FWLB with First Trip Failed Parameter Time, sec Transient Initiated 0.0 MFW to both SGs Interrupted 1.OE-6 EFIC Actuated on Low SG-B Level 6.515 Peak Thermal Power Occurs 6.523 DSS High RCS Pressure Trip Setpoint Reached 9.025 Regulating Control Rod Groups Begin to Insert 10.258 Turbine Trip, TSVs Begin to Close 10.260 Initial PSV Lift -12 Peak RCS Pressure occurs 13.518 Affected SG depressurization complete -36 EFW to Unaffected SG Begins 66.520 Final PSV closure -282 Peak Tave occurs -404 Transient Analysis Ends 600 Table 4-2: Results for FWLB with First Trip Failed Parameter Value Peak RCS pressure (psia) 2950.9 Peak thermal power (%RTP) 100.49 Peak Tave (OF) 630.54 Page 13
Figure 4-1: FWLB with Fail First Trip RCS Pressure 3000
-- Hot Leg 1 (CV 110-4)
-- Hot Leg 2 (CV 210-4)
-A Core Exit (CV 352-01)
.... .< Top of PZR (CV 405-01)
-v Lower RV Downcomer (CV 324-04)
I--------------------
I -------------
cdl ..........
crn C0 0O 420 480 540 600 Time (s)
I
Figure 4-2: FWLB with Fail First Trip Reactor Power 3200 --- Total Reactor Power (CVAR 370)
-- Thermal Power (CVAR 379) 20------------------
24 I--------------------
0-.-------........................ ....
. . . . . . . -------- r-------
10- I--------- ---------------- --
800 0 60 120 180 240 300 360 420 480 540 600 Time (s)
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Figure 4-3: FWLB with Fail First Trip Reactivity 0.010 0.000
-0.010
-0.020 0
-0.030
-0.040
ý CDo
-0.050 S0 0 300 Time (s) 3
Figure 4-4: FWLB with Fail First Trip Primary System Temperatures 720
-- RCS Ave Temp (CVAR 900)
-o Loop 1 Hot Leg (CV 110-4)
Loop 2 Hot Leg (CV 210-4) 700......... - 7v Loop Loop IA 1 B ColdCold Leg Leg (CV 160-4)
(CV 180-4)
Loop 2A Cold Leg (CV 260-4)
-+ Loop 2B Cold Leg (CV 280-4) 680 .................. - x Top of Pressurizer Liquid Temperature (CV 405) ------------------
660 --- ..... ... ...........
S660 -- -- ----........
--- ...... ..............--..............--... ......................... I-------------------
cl) 60 -- - - - - -- -- -- - - - - -,-- - . . .. . . -- - - i . . ... ...-- m '- :k---
-- - - - -- -- - .^ -, -- --- - -- - - - - -
545 .......----.----- .---- .---. I..-- .. -....
520 0= 0W 0 60 120 180 240 300 360 420 480 540 600 0 to Time (s)
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First Figure 4-5: FWLB with -Fail First Trip Indicated Pressurizer Level Level 400 Pressurizer Indicated with.Fail FWLB 4-5:
Figure I- CVAR 409. I 380 ............ --------------
360 ---------------..........------------------ I............ ----------- .......... .............
340 ............ ........................ ...........------------
_m _m PPRRERR 2RRRMMnPRPMRRR ---- ----------
RPPRRPRRRRRSýý ý=mmm"m=
320 -------------------- ---- ------------ -- ---- ---------- .............. -----------------
................ i 300 ............ ---------------- ---------------- --------------------- ----------------- -------------- :---------------
- 280 260 ------------------ .............
240 ------------ ------------- ------------------ ----------- -------------- -------------_
220 ------------ ---------------- ------------------ -------------- --- ----------- -------------- -------------------
200 0 U 60 120 180 240 300 360 420 480 540 600 Time (s)
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Figure 4-6: FWLB with Fail First Trip Pressurizer Collapsed Liquid Level 44 IICVAR 325 I 42 40 -------------------------------------------------- ---------------------------- ----------------------------- ------------------
38 -----------------
36 :-------------------
34 ---------------------------------------------
------------------- I:--------
-o 32 -------------------------------------
-------------------------- ------------------- -------------- --------------- L--------------
30 ----------------
28 ................
------------- I .................. -------- ---------------- --------- ------------------------------- ------------------
26 ................ ------------------
I-------------------- - ---------------------------- ----------------------------- ------------------
24 oLA t"0 0 60 120 180 240 300 360 420 480 540 600 Time (s)
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Figure 4-7: FWLB with Fail First Trip Pressurizer Surge Line Flow 500 250 ------------------- --------------------------------------
-250 ...................------
C
-500
-750 .1.1.'..... --------------------------.....
-1000 I--------
-1250 ------------------- -------
-1500 S0 0 60 120 180 240 300 360 420 480 540 600 Time (s) t0
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Figure 4-8: FWLB with Fail First Trip RCS Volumetric Flow Rate 400000
- aCVAR 332R 350000 .................. ------- ---------------------------- ...................... -----------------------------
300000 .................. ------- ----------- ----------------
250000 ------------------ ------- ...........
I...................
0 200000 - --------------------- ------------
U 1500001 . ----------- ................ ---------------- .........................................
100000 1. ................. ----- --- ---
50000 ................---------------------------- --------------------------------------
ýo ý 0 0 0j 0 60 120 180 240 300 360 420 480 540 600 Time (s) 8
Figure 4-9: FWLB with Fail First Trip Pressurizer Safety Valve Flow with .FailFlow Valve First Trip 500 Safety FWLB Figure 4-9: Pressurizer mflowj-492000000 450 ............... .......................................------------------------------- ------------------
400 ................ -------------------- -------------------------------------------------- ..............
350 300
-1 0 250 ---------- ......... -----------------------------------
200 il . .......-.-..--.-.-.-...-- ............. I--------------------
150 1ý 100 --------------------.................. ----------------
50 .----- .............. ... --- ----------------------------------
CDo ý 0 0 0-120 180 240 360 420 480 540 Time (s)
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Figure 4-10: FWLB with Fail First Trip SG Secondary Side Liquid Level 12 - q 9 10 1 ---
I------------- ------------
8 -----------
au
-V--- .......- -----------------------------
6 ----------------- .........
Cn 4 --------------------------------
Ct2
----------------- B--------------- -----------
2 ----------- ................
0 - @ 0 9 0 9 @ @ e. 0 0 8 @ @ 0 @ @-."- v ow 0 60 120 180 240 3T0 360 600 CDQ Time (s)
I0
Figure 4-11: FWLB with Fail First Trip SG Secondary Side Inventory, Ibm 80000 I
--- SG-A (CVAR 607)
-o SG-B (CVAR 707) 70000 1------------- -----------
60000 1- ------------------
S 0 5 00 00. 1 . . . . . . .. . . . . . . . .. . . . . . . . . . .. . . . . . . . . -------------------
I ---------
------------------------- I ------------------
I-----------------------------------------
-o 40000 -----------------.. . ---------------- I.........-----------------------------
I ..................
c-I 0 30 000 -.................... I------------------------------------------------------------------------------
C/I 0Ct2 20000 --- --------- ---- ---- --------------- -----.....................
10000 ...... ------------------- ..................
p-o 0 0 00 180 240 300 360 420 480 540 600 C
Time (s) 11
Figure 4-12: FWLB with Fail First Trip SG % Operating Range 1oo
-- SG-A (CVAR 612)
-- i SG-B (CVAR 712) 90 ..................... -------------- v.......... ------- . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . ..................
80 --------------------- ---------- ........----------------- .......
704- .....................
60 -------------- : ---------------- ------------- .......
C 50 ............------------------ I--------------- -- ------- -----
0#
40 ------------ .......... I----------------- ------
30 -----------.................. I-----------------
20 1--I---. ............ ---------------- --------------- ............. I--------------- --------------
10 ---------- --------- - - - -- ------------------ ------------------ - ----------- ------------------- -------------
0 4 % s ~ if 4 l --- , MmfMW RRa99 3R SH1 " .H-- - --- Hea a e * -i~ 00 0 60 120 180 240 300 360 420 480 540 600 0.Q Time (s)
CD/
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Figure 4-13: FWLB with Fail First Trip SG Pressure, psia 1200
-- n Unaffected (CV648-1)
-* Affected (CV 748-1) 1100 .- "
1000 ----
100 0 --- .----- .
-----------.--------- --------- .------- ...........I .---------- -------.------------ -- .-- .---------------
90 080 -------- --- --------------------------------- - - - - - - - - - - - - - - - ----
50 0 ---.-------------- ..-------
400 ......................... . .... .... .......
...--- I..--
400 20 o -- . .
0 60 1.20 180 240 300 360 420 480 540 600 Time (s) 13
Figure 4-14: FWLB with Fail First Trip EFW Flow 100
--s To Unaffected SG (JUN 626)
--- To Affected SG (JUN 726) 90 -------------
80 70 .......... ----------------
I 60 ----------- ----- ---------------- ----- ----------------------------------------------- -----------------------
0 50 . ......... ----------------------
40 --------------- - -- ------------ ----- - ------------- -
30 ----------- ---------------------- .7 ------------------ ------------ -----------------------
20 ----------- ----------------- ... ------------------ --------.......... -------............ ---------------.......
10 ----------- --- - ----------- ... ... ---------- ---------- -------- ---------------------
01 -- j 0 0,0 0-- - 1 0 0 0 0 1ý 0 0-0 0 ý 0 - I.- - 0 0 0 0 0 0 0 0. 0 W A A R A @ - , - - - 0.0 0P 60 120 180 240 300 360 420 480 540 600 Time (s)
-.1 14
Figure 4-15: FWLB with Fail First Trip Integrated MSSV Flow 100000
-- Unaffected SG (OVAR 659)
,-- Affected SG (CVAR 660) 90000------------------------------------------------------------------- ------------------
7 0000 . . . . . . . .. . . . . . . . . .. .. --.-------------------
80 000 - - -- - - - - --
70000 ---- -
360000- -- . .. . . .. . . . .. . . .. . . .. . . .. . .--..
C
-Er 200050000--------------- ----- 1 -------------------------
10000
.* o. . o 0 60 120 180 240 300 360 420 480 540 600 0 Time (s)
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ANP-3052 Revision 0
5.0 REFERENCES
- 1. BAW-10164PA-06, "RELAP5/MOD2-B&W - An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis."
- 2. BAW-10193PA-00, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurizer Water Reactors."
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