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Category:Report
MONTHYEARML24047A2092024-02-22022 February 2024 Calendar Year 2023 Baseline Inspection Completion RA-22-0165, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 242022-06-0909 June 2022 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 24 RA-22-0134, Unit 1 Cycle 23 Mellla+ Eigenvalue Tracking Data2022-05-25025 May 2022 Unit 1 Cycle 23 Mellla+ Eigenvalue Tracking Data RA-21-0198, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 252021-06-21021 June 2021 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 25 RA-21-0176, Cycle 24 Mellla+ Eigenvalue Tracking Data2021-06-0707 June 2021 Cycle 24 Mellla+ Eigenvalue Tracking Data RA-20-0353, Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1)2021-02-24024 February 2021 Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1) RA-20-0347, 10 CFR 71.95 Report on the 3-60B Cask2020-11-16016 November 2020 10 CFR 71.95 Report on the 3-60B Cask RA-19-0479, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-12-31031 December 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-19-0411, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-10-23023 October 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-19-0338, Pressure and Temperature Limits Report for Unit Nos. 1 and 22019-08-15015 August 2019 Pressure and Temperature Limits Report for Unit Nos. 1 and 2 RA-19-0243, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-07-0202 July 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019 Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46 RA-19-0240, Response to Request for Additional Information Regarding Advanced Framatome Methodologies License Amendment Request2019-05-29029 May 2019 Response to Request for Additional Information Regarding Advanced Framatome Methodologies License Amendment Request RA-19-0139, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-03-18018 March 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-18-0100, Request for License Amendment Regarding Application of Advanced Framatome Methodologies2018-10-11011 October 2018 Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-18-0131, Technical Requirements Manual, Revision 662018-08-13013 August 2018 Technical Requirements Manual, Revision 66 ML18249A1592018-08-13013 August 2018 Technical Requirements Manual, Revision 73 RA-18-0024, Supplemental Information to Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report2018-05-29029 May 2018 Supplemental Information to Request for License Amendment to Revise the Technical Specifications to Relocate the Pressure-Temperature Limit Curves to a Pressure and Temperature Limits Report BSEP 18-0044, Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D2018-04-11011 April 2018 Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 ML16287A4222016-09-26026 September 2016 FS1-0028339 Revision 1.0, Brunswick, Unit 1, Cycle 21 and Unit 2 Cycle 23 MELLLA SLMCPR Analyses with SAFLIM30 Methodology. ML16223A7252016-08-17017 August 2016 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Phase 2 of Order EA-13-109 (Severe Accident Capable Hardened Events) CAC Nos. MF4467 and MF4468) ML18250A0022016-08-11011 August 2016 Technical Requirements Manual, Revision 70 ML18250A0032016-08-11011 August 2016 Technical Requirements Manual, Revision 63 ML16257A4112016-07-31031 July 2016 DUKE-0B21-1104-000(NP), Safety Analysis Report for Brunswick Steam Electric Plant, Units 1 and 2, Maximum Extended Load Line Limit Analysis Plus. ML16041A4352016-03-0101 March 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force ML16257A4062015-12-31031 December 2015 ANP-3106(NP), Revision 2, Brunswick Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for Atrium 10XM Fuel for Mellla+ Operation. ML15275A2902015-09-30030 September 2015 Areva, Inc., - ANP-3397NP, Revision 0, Brunswick, Unit 2, Atrium 11 Lead Test Assemblies Design & Licensing Summary Report. BSEP 16-0056, ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain.2015-07-31031 July 2015 ANP-3108(NP), Revision 1, Applicability of Areva BWR Methods to Brunswick Extended Power Flow Operating Domain. ML16257A4082015-07-31031 July 2015 ANP-3105(NP), Revision 1, Brunswick Units 1 and 2 LOCA Break Spectrum Analysis for Atrium 10XM Fuel for Mellla+ Operation. BSEP 15-0004, Enclosure 3 - Areva Operability Assessment. (CR 2014-7395)2015-01-13013 January 2015 Enclosure 3 - Areva Operability Assessment. (CR #2014-7395) BSEP 14-0131, Expedited Seismic Evaluation Process Report in Response to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident2014-12-18018 December 2014 Expedited Seismic Evaluation Process Report in Response to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident ML14297A2662014-10-24024 October 2014 Record of Review, Brunswick Steam Electric Plant, Units 1 and 2, LAR Attachment U- Table U-1 Internal Events PRA Peer Review- Facts and Observations (F&Os), 10/24/14 BSEP 14-0101, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors2014-09-0404 September 2014 Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors BSEP 14-0093, Report of 10 CFR 50.59 Evaluations and Commitment Changes2014-08-14014 August 2014 Report of 10 CFR 50.59 Evaluations and Commitment Changes ML14176A9612014-06-24024 June 2014 Submittal of Non-Proprietary BWROG Technical Product, BWROGTP-11-006 - ECCS Containment Walkdown Procedure, Rev 1 (January 2011), as Formally Requested During the Public Meeting Held on April 30, 2014 BSEP 14-0028, Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi A2014-03-31031 March 2014 Seismic Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accid ML13220A0902013-11-22022 November 2013 Interim Staff Evaluation Regarding Overall Integrated Plan in Response to Order EA-12-049 - Mitigation Strategies ML13317A5922013-11-20020 November 2013 Mega-Tech Services, LLC Technical Evaluation Report Regarding the Overall Integrated Plan for Brunswick Steam Electric Plant, Units 1 and 2, TAC Nos.: MF0975 and MF0976 ML13277A0412013-09-19019 September 2013 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition ML16257A4072013-05-31031 May 2013 ANP-3013(NP), Revision 0, Brunswick Unit 1 Cycle 19 Fuel Cycle Design Mellla+ Operating Domain. BSEP 13-0030, Overall Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2013-02-28028 February 2013 Overall Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events BSEP 13-0002, Technical Data Record, 12-9197120-000, Brunswick SAFLIM3D LAR Support - Calculation Error Impact2013-01-14014 January 2013 Technical Data Record, 12-9197120-000, Brunswick SAFLIM3D LAR Support - Calculation Error Impact ML13031A0112013-01-11011 January 2013 Engineering Information Record, Document No. 51-9196989-000, Supplemental Information for Brunswick SAFLIM3D Submittal: Impact of Assemblies Outside Existing Channel Bow Fast Fluence Gradient Database BSEP 12-0127, Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2012-11-27027 November 2012 Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident BSEP 12-0133, Document No. 51-9195058-000, Areva Input to Process Energy: Follow-Up to SAFLIM3D LAR Request for Additional Information 3, Enclosure 3 to BSEP 12-01332012-11-27027 November 2012 Document No. 51-9195058-000, Areva Input to Process Energy: Follow-Up to SAFLIM3D LAR Request for Additional Information 3, Enclosure 3 to BSEP 12-0133 BSEP 12-0128, Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation2012-11-14014 November 2012 Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation ML12321A3192012-11-0101 November 2012 1000771.402NP, Revision 2, Life Extension for Core Plate Plugs at Brunswick Nuclear Plant Unit 2. ML12065A3802012-03-26026 March 2012 Request to Reinitiate Section 7 Consultation for Atlantic Sturgeon at Brunswick Steam Electric Plant, Units 1 and 2 BSEP 12-0031, Areva Document ANP-3086(NP), Brunswick Units 1 and 2 SLMCPR Operability Assessment Critical Power Correlation for Atrium 10XM Fuel - Improved K-factor Model, Enclosure 20 to BSEP 12-00312012-02-29029 February 2012 Areva Document ANP-3086(NP), Brunswick Units 1 and 2 SLMCPR Operability Assessment Critical Power Correlation for Atrium 10XM Fuel - Improved K-factor Model, Enclosure 20 to BSEP 12-0031 2024-02-22
[Table view] Category:Miscellaneous
MONTHYEARML24047A2092024-02-22022 February 2024 Calendar Year 2023 Baseline Inspection Completion RA-22-0165, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 242022-06-0909 June 2022 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 24 RA-21-0198, Inservice Inspection Program Owner'S Activity Report for Refueling Outage 252021-06-21021 June 2021 Inservice Inspection Program Owner'S Activity Report for Refueling Outage 25 RA-20-0353, Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1)2021-02-24024 February 2021 Proposed Alternative to ASME Boiler & Pressure Vessel Code Section XI Requirements for Repair/Replacement of Service Water (SW) System Buried Piping in Accordance with 10 CFR 50.55a(z)(1) RA-19-0479, Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies2019-12-31031 December 2019 Supplement to Request for License Amendment Regarding Application of Advanced Framatome Methodologies RA-19-0223, Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.462019-05-30030 May 2019 Annual Report of Changes to or Errors in Emergency Core Cooling System Evaluation Models Pursuant to 10CFR 50.46 BSEP 18-0044, Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D2018-04-11011 April 2018 Application of Dissimilar Metal Weld Full Structural Overlay - Reactor Pressure Vessel Nozzles N4A and N4D ML17024A0362016-12-31031 December 2016 Operating Data Report for 2016 ML16223A7252016-08-17017 August 2016 Interim Staff Evaluation Relating to Overall Integrated Plan in Response to Phase 2 of Order EA-13-109 (Severe Accident Capable Hardened Events) CAC Nos. MF4467 and MF4468) ML16041A4352016-03-0101 March 2016 Staff Assessment of Information Provided Pursuant to Title 10 of the Code of Federal Regulations Part 50 Section 50.54(f) Seismic Hazard Reevaluations for Recommendation 2.1 of the Near-Term Task Force BSEP 15-0004, Enclosure 3 - Areva Operability Assessment. (CR 2014-7395)2015-01-13013 January 2015 Enclosure 3 - Areva Operability Assessment. (CR #2014-7395) BSEP 14-0131, Expedited Seismic Evaluation Process Report in Response to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident2014-12-18018 December 2014 Expedited Seismic Evaluation Process Report in Response to 10 CFR 50.54(f) Request for Information Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-lchi Accident ML14297A2662014-10-24024 October 2014 Record of Review, Brunswick Steam Electric Plant, Units 1 and 2, LAR Attachment U- Table U-1 Internal Events PRA Peer Review- Facts and Observations (F&Os), 10/24/14 BSEP 14-0101, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors2014-09-0404 September 2014 Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors BSEP 14-0093, Report of 10 CFR 50.59 Evaluations and Commitment Changes2014-08-14014 August 2014 Report of 10 CFR 50.59 Evaluations and Commitment Changes ML13277A0412013-09-19019 September 2013 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition BSEP 13-0030, Overall Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events2013-02-28028 February 2013 Overall Integrated Plan in Response to March 12, 2012, Commission Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events BSEP 12-0127, Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident2012-11-27027 November 2012 Recommendation 2.3 Seismic Walkdown of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident BSEP 12-0128, Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation2012-11-14014 November 2012 Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation ML1127000692011-09-26026 September 2011 Enclosure 2, Mfn 10-245 R4, Description of the Evaluation and Surveillance Recommendations for BWR/2-5 Plants BSEP 14-0007, 1707-01-F03, Rev. 3, Operability Assessment (CR2013-8241)2011-09-0808 September 2011 1707-01-F03, Rev. 3, Operability Assessment (CR#2013-8241) BSEP 11-0038, Renewed Facility Operating License and Cycle 20 Core Operating Limits Report2011-04-0808 April 2011 Renewed Facility Operating License and Cycle 20 Core Operating Limits Report ML1108712502011-03-24024 March 2011 BWR Vessel and Internals Inspection Summaries for Spring 2010 Outages BSEP 10-0097, Report of 10 CFR 50.59 Evaluations and Commitment Changes2010-08-18018 August 2010 Report of 10 CFR 50.59 Evaluations and Commitment Changes ML1019304172010-05-0606 May 2010 Tritium Database Report ML1012704392010-05-0505 May 2010 Y020100187 - List of Historical Leaks and Spills at U.S. Commercial Nuclear Power Plants ML1019305532009-12-14014 December 2009 NEI 07-07 NEI Groundwater Protection Initiative NEI Peer Assessment Report for Brunswick BSEP 09-0022, Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation2009-02-23023 February 2009 Special Report - Technical Requirements Manual Section 3.4, Accident Monitoring Instrumentation BSEP 08-0121, Submittal of Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors2008-09-24024 September 2008 Submittal of Annual 10 CFR 50.46 Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML0711000062006-12-0808 December 2006 Operational Safety Review Team (Osart) Mission and Follow-up Visit Report BSEP 06-0095, Report of 10 CFR 50.59 Evaluation and Commitment Changes2006-08-24024 August 2006 Report of 10 CFR 50.59 Evaluation and Commitment Changes BSEP 06-0075, Groundwater Questionnaire2006-07-28028 July 2006 Groundwater Questionnaire ML0614404872006-05-18018 May 2006 Submittal of 10-O Report for Quarterly Period Ending March 31, 2006 BSEP 05-0137, Core Operating Limits Report, Revision 2, Cycle 152005-11-14014 November 2005 Core Operating Limits Report, Revision 2, Cycle 15 BSEP 05-0110, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors2005-08-31031 August 2005 Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors ML0515903332005-05-31031 May 2005 Submittal of 10-Q Report ML0535300692005-05-0909 May 2005 Operational Safety Review Team (Osart) Mission to Brunswick Nuclear Plant 9-25 May 2005 BSEP 05-0018, Occupational Radiation Exposure Report for 20042005-04-20020 April 2005 Occupational Radiation Exposure Report for 2004 ML0434301332004-11-30030 November 2004 Submital of 10-Q Report ML0414602462004-05-19019 May 2004 Q Report for Period Ending 03/31/2004 ML0324716322003-08-22022 August 2003 Submittal of 10-Q Report ML0318909362003-07-10010 July 2003 Relaxation of the Order, Exercising Enforcement Discretion, and Extension of the Time to Submit an Answer or Request a Hearing Regarding Order EA-03-038, Fitness-for-Duty Enhancements for Nuclear Security Force Personnel for Brunswick, Crys ML0317000272003-06-0606 June 2003 Submittal of 10-O Report ML19204A1621998-11-18018 November 1998 Enclosure 1 - Staff Evaluation Report of Individual Plant Examination of External Events (IPEEE) Submittal on Brunswick Steam Electric Plant (Bsep), Units 1 and 2 2024-02-22
[Table view] |
Text
BSEP 14-0007 Enclosure 3 Non-Proprietary Version of Operability Assessment (CR #2013-8241) 1707-01-F03 (Rev. 003, 09/08/2011)
A OPERABILITY ASSESSMENT (CR #2013-8241)
AREVA Issue
Description:
CR2011-2274 identifies an issue with the approved ACE correlation for the ATRIUM T M 1OXM* fuel design with regard to the calculation of K-factor within the ACE correlation.
[This issue was discussed with the USNRC during the review of a licensing amendment request to add the ATRIUM 1 OXM version of the ACE correlation to the list of approved COLR references in the Brunswick Technical Specifications.
As part of the approval of this licensing amendment (Reference 1), the USNRC imposed a license condition that requires follow-on operability assessments.
The specifics of this license condition were modified in the Reference 2 acceptance of the SAFLIM3D methodology to the list of approved COLR references in the Brunswick Technical Specifications.
The revised license condition is: Safety Limit Minimum Critical Power Ratio (SLMCPR), setpoint, and core operating limit values determined using the ANP-10298PA, ACE/ATRIUM 10XM Critical Power Correlation (i.e., TS 5.6.5.b.21), shall be evaluated to verify the values determined using the NRC-approved method remain applicable and the core operating limits include margin sufficient to bound the effects of the K-factor calculation issue described in AREVA Operability Assessment CR 2011-2274, Revision 1. SLMCPR shall be evaluated with methods described in AREVA Document ANP-3086(P), Revision 0, "Brunswick Unit 1 and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for ATRIUM IOXM Fuel -Improved K-factor Model. " Setpoint and core operating limit values shall be evaluated with methods described in AREVA Operability Assessment CR 2011-2274, Revision 1. The results of the evaluation shall be documented and submitted to the NRC, for review, at least 60 days prior to startup of each operating cycle.This is the operability assessment for Brunswick Unit 1 Cycle 20 and as such addresses impacts on information provided in Reference 3.
References:
- 1. Brunswick Steam Electric Plant, Units 1 and 2 -Issuance of Amendments regarding Addition of Analytical Methodology Topical Report to Technical Specification 5.6.5 (TAC Nos. ME3856 and ME3857), Accession No.ML111010234, April 8, 2011.2. Brunswick Steam Electric Plant, Units 1 and 2 -Issuance of Amendments Regarding Addition of Analytical Methodology Topical Reports to Technical Specification 5.6.5 and Revision to Minimum Critical Power Ratio Safety Limit (TAC Nos. ME8135 and ME8136), March 1, 2013. (NRC Ascension Number ML13037A551)
- 3. ANP-3263(P)
Revision 0, Brunswick Unit I Cycle 20 Reload Safety Analysis, AREVA NP, December 2013.4. ANP-3086(P)
Revision 0, Brunswick Unit I and Unit 2 SLMCPR Operability Assessment Critical Power Correlation for-A TRIUM IOXM Fuel -Improved K-factor Model, AREVA NP, February 2012.Affected Projects Customer, Reload or Customer Reactor (if applicable)
Project Identifier BRK1-20 Duke -Progress Energy Carolinas Brunswick Unit 1 ATRIUM is a trademark of AREVA NP.Ref: 1707-01 Page 1 of 10 AREVA NP Inc.
1707-01-F03 (Rev. 003, 09/08/2011)
Assessment:
The USNRC approved ACE correlation was used in Cycle 20 core design and licensing analyses.
It will also be used in the POWERPLEX-Ill core monitoring system to ensure Technical Specifications compliance of the fuel operating limits during operation.
This assessment addresses the licensing and subsequent monitoring of the Cycle 20 core in regard to the potential impacts of the ACE correlation issue described above.I These are the same methods previously utilized in the Operability Assessment CR2011-2274 Revision 1 and their use is therefore consistent with the license condition imposed by Reference 1 and the license condition in Reference 2.The evaluation for the SLMCPR calculation was performed using methods as described in document ANP-3086(P) Revision 0 (Reference 4), as required by the Reference 2 licensing condition.
Methods Used for Operability Evaluation In the licensed methodology, a K-factor is determined by []For the setpoint and core operating limit evaluation a version of the ACE correlation was constructed in which the K-factor []For the SLMCPR evaluation, [utilized.] as described in ANP-3086(P)
Revision 0 was Ref: 1707-01 AREVA NP Inc.Page 2 of 10 1707-01-F03 (Rev. 003, 09/08/2011)
Justification of the Evaluation Tool I Justification for the critical power correlation used in the SLMCPR evaluation is provided in Reference 4.Monitorina During Cycle 20 Operation The Brunswick Unit 1 Cycle 20 core is composed of the following fuel types: Fresh Once-Burnt Twice-Burnt Fuel Design ATRIUM 10XM ATRIUM 1OXM ATRIUM-10 No. Bundles 222 234 104 CPR Correlation ACE (10XM)ACE (10XM)SPCB As noted earlier, the concern identified in the condition report affects the K-factor calculation within the ACE correlation.
This only affects the ATRIUM 1OXM fuel in the Cycle 20 core since the co-resident ATRIUM-10 fuel utilizes a different CPR correlation.
Once Cycle 20 begins operation, compliance to the operating limits will be performed using the POWERPLEXO-III core monitoring system which contains the approved version of the ACE correlation.
The nominal design step-through depletion was recalculated with the modified version [] of the ACE correlation in order to evaluate potential non-conservatisms in monitoring during Cycle 20 operation.
Ref: 1707-01 AREVA NP Inc.Page 3 of 10 1707-01-F03 (Rev. 003, 09/08/2011)
I Figure 1 Impact on Core Limiting MFLCPR during Cycle 20 Figure 1 shows that there is no significant non-conservative impact to the limiting CPR margin throughout most of the cycle. []Figure 2 Change in ATRIUM 10XM CPR at Rated Conditions It is also useful to quantify the impact of the modified correlation on the calculated CPR for the ATRIUM 10XM fuel design (on an absolute AMCPR basis). For the same rated power cases, the maximum change in the calculated CPR was determined to be [ ]Ref: 1707-01 AREVA NP Inc.Page 4 of 10 1707-01-F03 (Rev. 003, 09/08/2011)
This evaluation was then expanded to look at a number of off-rated cases to determine if the rated power AMCPR results are representative for operation at reduced power conditions.
The off-rated cases were performed for a subset of the cycle exposure points presented in Figure 2. The results of this evaluation are summarized in Figure 3.The primary results of this off-rated evaluation include: Figure 3 Change in ATRIUM 1OXM CPR at Rated and Off-Rated Conditions The off-rated evaluation described above was performed for power levels at and above [I Ref: 1707-01 AREVA NP Inc.Page 5 of 10 1707-01-F03 (Rev. 003. 09/08/2011')
lmpact on the Safety Limit MCPR (SLMCPR)An evaluation of the impact on the SLMCPR was performed consistent with the Reference 2 licensing condition.
The critical power correlation described in Reference 4 was used in the assessment.
The results using the Reference 4 critical power correlation support a 0.01 lower SLMCPR for both TLO and SLO than the calculations with the ACE correlation described in ANP-10298PA.
Overall, the results from Reference 3 and the operability assessment support the TLO SLMCPR of 1.08 and SLO SLMCPR of 1.11 presented in the Brunswick Unit 1 Cycle 20 reload report.It should be noted MCPR is monitored relative to the OLMCPR and not directly to the SLMCPR. Therefore, there is no need in this SLMCPR evaluation to [I Impact on CPR Operating Limits (OLMCPR)The licensing analyses result in OLMCPRs that are a combination of power-dependent (MCPRp) and flow-dependent (MCPRf) limits. These are calculated from a series of quasi-steady-state and transient pressurization analyses.
The quasi-steady-state events that have the potential to contribute to the MCPRp limits are the Control Rod Withdrawal Error (CRWE) and the Loss of Feedwater Heating (LFWH).The MCPRf limits are based on the quasi-steady-state flow run-up analysis.
[The Cycle 20 flow run-up results were reviewed to determine the margin to the MCPRf limits presented in Reference
- 3. This review confirmed that for MCPRf limits above 1.35 (the lowest rated power MCPR limit), the margin to the limit is more than adequate to compensate for the K-factor [ ] in setting the MCPRf limits and the steady-state CPR monitoring concern.The MCPRp limits are a combination of the results from the pressurization analyses, CRWE, and potentially LFWH.Pressurization Transient Impacts. This K-factor issue [] the later cycle pressurization transient analysis results are only minimally affected by this K-factor issue. This was verified with analyses using the modified version of the ACE Ref: 1707-01 Page 6 of 10 AREVA NP Inc.
1707-01-F03 (Rev. 003, 09/08/2011) correlation.
Pressurization transient analyses are also performed
[The results of the analyses at all power levels show that in all cases (including the equipment out of service scenarios) the change in ACPR was such that there remains sufficient margin to the MCPRp limits. The margin is also large enough to account for the MCPR monitoring concern discussed above.CRWE Impacts. CRWE analyses have been performed using both the as-approved version of the ACE correlation
[ ] The CRWE results reported in Table 5.10 of Reference 3 are the most limiting from both sets of calculations.
The CRWE event is limiting for some conditions and sets the MCPRp limit for nominal scram speed with equipment in-service.
For this reason, []The RBM operability requirements specified in Table 5.11 of Reference 3 [] are valid for both the as-approved and modified version of the ACE correlation.
LFWH Impacts. The LFWH event is non-limiting with significant margin to the Cycle 20 MCPRp limits.This was confirmed for Cycle 20 with the performance of cycle-specific calculations using the modified ACE correlation.
This calculation verified that the generic LFWH methodology continues to bound the cycle specific results including the impact of the ACE K-factor [ ] .Adequate margin exists at all power levels to offset the potential monitoring impact of the K-factor [ ].Summary of MCPR impacts: The Reference 3 MCPRf and MCPRp limits have sufficient margin to account for the K-factor [ ] on the Cycle 20 quasi-steady-state and pressurization transient analysis results. This evaluation confirms [ ] for the use of the as-approved ACE correlation in the POWERPLEX-III core monitoring system.Impact on Instrumentation Setpoints The licensing analyses for Cycle 20 support CPR-based instrumentation setpoints for both the Rod Block Monitor (RBM) and for the Oscillation Power Range Monitor (OPRM).The RBM setpoints are based upon the CRWE event which in turn has been analyzed based on the analytical high power trip setpoints listed in Table 5.10 of Reference
- 3. Since it was confirmed that the MCPR, limits continue to bound the CRWE ACPR results, the use of the Table 5.10 setpoints continues to be supported for Cycle 20 operation.
The OPRM setpoints are based upon a combination of the cycle-specific DIVOM (Delta over Initial CPR Versus Oscillation Magnitude), the plant-specific HCOM (Hot Channel Oscillation Magnitude), and cycle-specific 2PT (2 recirculation pump trip) results. The cycle-specific DIVOM and 2PT results are the two components of this calculation that are potentially impacted by a CPR correlation error. Two sets of operating limits are reported for each OPRM setpoint, OLMCPR(SS) and OLMCPR(2PT).
The DIVOM can impact both OLMCPRs and the 2PT results can only impact the OLMCPR(2PT) results.DIVOM Impacts: The discussion in the DIVOM approved topical report (BAW-10255PA Revision 2)indicates that the DIVOM analysis is [3 Therefore, there is no DIVOM impact on the OLMCPR(SS) and OLMCPR(2PT) results.2PT Impacts: The recirculation pump trip cases provide a ratio of the CPR before and after the 2PT trip. An increase in this ratio will result in an increase in the required MCPR prior to the pump trip, that is OLMCPR(2PT).
This evaluation considers both the impact on the selection of the OPRM setpoint Ref: 1707-01 Page 7 of 10 AREVA NP Inc.
1707-01-F03 (Rev. 003, 09/08/2011) and on whether operation with the selected setpoint retains adequate CPR margin to compensate for monitoring the core with the unmodified ACE correlation.
The OLMCPR(SS) and OLMCPR(2PT) values provided in Table 4.3 of Reference 3 []. The use of these setpoints remains supported with the unmodified ACE correlation in the POWERPLEX-Il1 core monitoring system.Imoact on Fuel Loadina Error (FLE)The FLE (misorientation and mislocation) is an infrequent event that is analyzed to assure the off-site dose criteria defined in Section 15.4.7 of NUREG-0800 is not exceeded.
[I Impact on Loss-of-Coolant Accident (LOCA)A MCPR limit is assumed in the LOCA analyses to apply an upper limit on the hot channel power. A MCPR of 1.34 was used in the Brunswick ATRIUM 1OXM LOCA analysis and is based on a conservatively low K-factor which results in a higher hot assembly power. []. Therefore, the results of the Brunswick ATRIUM 1OXM LOCA analysis remain applicable for Brunswick Unit 1 Cycle 20.Ref: 1707-01 AREVA NP Inc.Page 8 of 10 1707-01-F03 (Rev. 003, 09/08/2011)
Conclusion This evaluation used a combination of calculations and first principal arguments to address the impacts of the ACE correlation concern on planned Brunswick Unit 1 Cycle 20 licensing and operation.
The calculations were performed using the modified version of the ACE correlation, [I.The fuel in the Cycle 20 core that is impacted by the ACE K-factor [ ] is the fresh and once-burnt ATRIUM 1OXM fuel type. The co-resident twice-burnt ATRIUM-10 fuel design is licensed and will be monitored with the SPCB correlation and is therefore not impacted.This operability assessment has evaluated the potential impact on all CPR related limits or analyses associated with the limits reported in the Reference 3 Reload Safety Analysis report. This evaluation has determined the SLMCPR and corresponding MCPR operating limits remain applicable to Brunswick Unit 1 Cycle 20 operation.
The potential impact on the core monitoring system was conservatively addressed and was also found to be bounded by both the MCPRp and MCPRf values.The MCPRp limits were set based in part upon the use of the 111% RBM setpoint and this remains supported with the current MCPRp values. The RBM operability limits were also confirmed to be applicable for Cycle 20 operation.
The OPRM setpoints were evaluated using the modified ACE correlation and confirmed to remain applicable for Cycle 20 operation.
The fuel loading error infrequent event was evaluated and it was confirmed that the required offsite dose criteria continues to be met.There is no impact on the LOCA analysis.In conclusion, no changes are required in the operating limits or instrument setpoints supplied in the Reference 3 Reload Safety Analysis report. These limits may be used to support operation and monitoring with the current POWERPLEX-I11 core monitoring system.Ref: 1707-01 AREVA NP Inc.Page 9 of 10
-1707-01 -F03 (Rev. 003, 09/08/2011)
APPROVALS:-Y ________________
- 1' -'I* *y. -,a2--/1-20 13 Approved: Date: I +Issue Evaluator
-E. E. Riley (Neutronic Design and Analysis)Approved:
/jJ. Date: /_ ((1/3 Issue Evaluator
-D. R. Tinkler (TIH Applications)
Approved:
7 Y c c Date: Peer Review -R. E. Fowles (Neutronic Design and Analysis)Approved:
Date:3 Peer Review -D. D. Crockett (T/H Applications)
Approved:
.Date: Peer Review -M. T. Bunker (T/H Codes and Methods)Approved:
Date: /2 3 Issue Owner -E. E. Riley (Neutronic Design and Analysis)Ref: 1707-01 AREVA NP Inc.Page 10 of 10