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Category:Letter type:SBK
MONTHYEARSBK-L-23053, Radiological Emergency Plan (Ssrep), Revision 802023-06-22022 June 2023 Radiological Emergency Plan (Ssrep), Revision 80 SBK-L-23031, 2022 Annual Radioactive Effluent Release Report2023-04-28028 April 2023 2022 Annual Radioactive Effluent Release Report SBK-L-23044, 2022 Annual Radiological Environmental Operating Report2023-04-28028 April 2023 2022 Annual Radiological Environmental Operating Report SBK-L-23033, 2022 Annual Environmental Operating Report2023-04-11011 April 2023 2022 Annual Environmental Operating Report SBK-L-22093, 75 Day and 150 Day Submittal Letters2022-12-0101 December 2022 75 Day and 150 Day Submittal Letters SBK-L-22113, Radiological Emergency Plan (Ssrep), Revision 792022-11-16016 November 2022 Radiological Emergency Plan (Ssrep), Revision 79 SBK-L-22114, Independent Spent Fuel Storage Installation (Isfsi), Cask Registration2022-11-0909 November 2022 Independent Spent Fuel Storage Installation (Isfsi), Cask Registration SBK-L-22112, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2022-11-0303 November 2022 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration SBK-L-22094, Relief Request 4RA-22-001, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)2022-10-26026 October 2022 Relief Request 4RA-22-001, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) SBK-L-22105, Supplement to 2021 Annual Radioactive Effluent Release Report2022-10-24024 October 2022 Supplement to 2021 Annual Radioactive Effluent Release Report SBK-L-22103, 5 to the Physical Security Plan, Training and Qualification Plan and Safeguards Contingency Plan2022-10-12012 October 2022 5 to the Physical Security Plan, Training and Qualification Plan and Safeguards Contingency Plan SBK-L-22102, Independent Spent Fuel Storage Installation (ISFSI) Cask Registration2022-10-0606 October 2022 Independent Spent Fuel Storage Installation (ISFSI) Cask Registration SBK-L-22080, Submittal of 10 CFR 72.48 Report of Changes, Tests and Experiments for August 2020 to July 20222022-08-12012 August 2022 Submittal of 10 CFR 72.48 Report of Changes, Tests and Experiments for August 2020 to July 2022 SBK-L-22054, SBK-L-22054 - License Amendment Request 22-01, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections2022-06-0909 June 2022 SBK-L-22054 - License Amendment Request 22-01, Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam Generator Tube Inspections SBK-L-22044, 2021 Annual Radioactive Effluent Release Report2022-04-29029 April 2022 2021 Annual Radioactive Effluent Release Report SBK-L-22036, Steam Generator Tube Inspection Report2022-04-25025 April 2022 Steam Generator Tube Inspection Report SBK-L-22040, Radiological Emergency Plan (Ssrep), Revision 7822022-04-20020 April 2022 Radiological Emergency Plan (Ssrep), Revision 782 SBK-L-22024, Response to Request for Additional Information (Raj) Regarding License Amendment Request 21-01, Revise 120-Volt AC Vital Instrument Panel Requirements2022-03-0707 March 2022 Response to Request for Additional Information (Raj) Regarding License Amendment Request 21-01, Revise 120-Volt AC Vital Instrument Panel Requirements SBK-L-22003, Inservice Inspection Examination Report2022-01-24024 January 2022 Inservice Inspection Examination Report SBK-L-21126, Fourth 10 Year Interval - Inservice Inspection (ISI) Program Plan2022-01-0404 January 2022 Fourth 10 Year Interval - Inservice Inspection (ISI) Program Plan SBK-L-21135, Submittal of Changes to the Seabrook Station Technical Specification Bases2021-12-29029 December 2021 Submittal of Changes to the Seabrook Station Technical Specification Bases SBK-L-21108, Core Operating Limits Report for Cycle 222021-11-0404 November 2021 Core Operating Limits Report for Cycle 22 SBK-L-21106, Transmittal of WCAP-18607-NP, Analysis of Capsule X from the NextEra Energy Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program2021-09-30030 September 2021 Transmittal of WCAP-18607-NP, Analysis of Capsule X from the NextEra Energy Seabrook Unit 1 Reactor Vessel Radiation Surveillance Program SBK-L-21105, Withdrawal of Emergency License Amendment Request 21-02, Service Water System I Ultimate Heat Sink One-Time Allowed Outage Time (AOT) Extension2021-09-27027 September 2021 Withdrawal of Emergency License Amendment Request 21-02, Service Water System I Ultimate Heat Sink One-Time Allowed Outage Time (AOT) Extension SBK-L-21103, Request for Exercise of Enforcement Discretion2021-09-25025 September 2021 Request for Exercise of Enforcement Discretion SBK-L-21104, Emergency License Amendment Request 21-02, Service Water System I Ultimate Heat Sink Onetime Allowed Outage Time (AOT) Extension2021-09-25025 September 2021 Emergency License Amendment Request 21-02, Service Water System I Ultimate Heat Sink Onetime Allowed Outage Time (AOT) Extension SBK-L-21098, Supplement to License Amendment Request 21-01, Revise 120-Volt AC Vital Instrument Panel Requirements2021-09-22022 September 2021 Supplement to License Amendment Request 21-01, Revise 120-Volt AC Vital Instrument Panel Requirements SBK-L-21083, NextEra Energy Seabrook, LLC - Relief Request 3IR-18 and 3IR-19 for End of Third 10-Year ISI Interval for Examinations with Limited Coverage2021-08-18018 August 2021 NextEra Energy Seabrook, LLC - Relief Request 3IR-18 and 3IR-19 for End of Third 10-Year ISI Interval for Examinations with Limited Coverage SBK-L-21067, License Amendment Request 21-01, Revise 120-Volt AC Vital Instrument Panel Requirements2021-07-21021 July 2021 License Amendment Request 21-01, Revise 120-Volt AC Vital Instrument Panel Requirements SBK-L-21049, Revision to Updated Final Response to NRC Generic Letter 2004-022021-06-24024 June 2021 Revision to Updated Final Response to NRC Generic Letter 2004-02 SBK-L-21056, Response to Request for Additional Information Regarding Seabrook Steam Generator Tube Inspection Report Review References:2021-05-20020 May 2021 Response to Request for Additional Information Regarding Seabrook Steam Generator Tube Inspection Report Review References: SBK-L-21045, Submittal of 2020 Annual Radioactive Effluent Release Report2021-04-30030 April 2021 Submittal of 2020 Annual Radioactive Effluent Release Report SBK-L-21044, 2020 Annual Radiological Environmental Operating Report2021-04-30030 April 2021 2020 Annual Radiological Environmental Operating Report SBK-L-21042, Response to RIS 2021-01 Preparation and Scheduling of Operator Licensing Examinations2021-04-22022 April 2021 Response to RIS 2021-01 Preparation and Scheduling of Operator Licensing Examinations SBK-L-21033, Response to Request for Additional Information (RAI) Regarding License Amendment Request (LAR) 20-02, Resolve Non-Conservative Heat Flux Hot Channel Factor (Fo(Z)) Requirements2021-03-24024 March 2021 Response to Request for Additional Information (RAI) Regarding License Amendment Request (LAR) 20-02, Resolve Non-Conservative Heat Flux Hot Channel Factor (Fo(Z)) Requirements SBK-L-21023, Radiological Emergency Plan (Ssrep), Revision 772021-03-0404 March 2021 Radiological Emergency Plan (Ssrep), Revision 77 SBK-L-20154, Submittal of Changes to the Seabrook Station Technical Specification Bases2020-12-22022 December 2020 Submittal of Changes to the Seabrook Station Technical Specification Bases SBK-L-20149, Response to Request for Additional Information in Respect to One-Time Exemption from 10 CFR 73, Appendix B, Section VI, Subsection C.3.(I)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic2020-12-0808 December 2020 Response to Request for Additional Information in Respect to One-Time Exemption from 10 CFR 73, Appendix B, Section VI, Subsection C.3.(I)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic SBK-L-20146, One-Time Exemption Request for the Biennial Emergency Preparedness Exercise Requirements of 10 CFR 50, Appendix E. IV.F.2.b and IV.F.2.c2020-12-0303 December 2020 One-Time Exemption Request for the Biennial Emergency Preparedness Exercise Requirements of 10 CFR 50, Appendix E. IV.F.2.b and IV.F.2.c SBK-L-20144, Request for a One-Time Exemption from 10 CFR 73, Appendix B, Section VI Subsection C.3.(1)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic2020-11-23023 November 2020 Request for a One-Time Exemption from 10 CFR 73, Appendix B, Section VI Subsection C.3.(1)(1) Regarding Annual Force-on-Force (FOF) Exercises, Due to Covid 19 Pandemic SBK-L-20143, Withdrawal of Exemption Request Supporting Updated Final Response to NRC Generic Letter 004-022020-11-19019 November 2020 Withdrawal of Exemption Request Supporting Updated Final Response to NRC Generic Letter 004-02 SBK-L-20132, Fourth Interval Inservice Testing (IST) Program2020-10-21021 October 2020 Fourth Interval Inservice Testing (IST) Program SBK-L-20127, Generator Tube Inspection Report2020-10-20020 October 2020 Generator Tube Inspection Report SBK-L-20117, Response to Request for Additional Information Regarding Technical Specification 3/4.8.1 for a One-Time Extension of Allowed Outage Time License Amendment Request2020-09-23023 September 2020 Response to Request for Additional Information Regarding Technical Specification 3/4.8.1 for a One-Time Extension of Allowed Outage Time License Amendment Request SBK-L-20103, Inservice Inspection Examination Report2020-07-27027 July 2020 Inservice Inspection Examination Report SBK-L-20092, Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-411 and TSTF-4182020-07-13013 July 2020 Response to Request for Additional Information Regarding License Amendment Requests to Adopt TSTF-411 and TSTF-418 SBK-L-20068, License Amendment Request 20-01, One-Time Change to the Seabrook Technical Specifications A.C. Sources - Operating2020-07-13013 July 2020 License Amendment Request 20-01, One-Time Change to the Seabrook Technical Specifications A.C. Sources - Operating SBK-L-20085, Supplement to Exemption Request for Access Authorization and Fitness for Duty Requirements Due to COVID-19 Pandemic2020-06-16016 June 2020 Supplement to Exemption Request for Access Authorization and Fitness for Duty Requirements Due to COVID-19 Pandemic SBK-L-20077, Core Operating Limits Report for Cycle 212020-06-0101 June 2020 Core Operating Limits Report for Cycle 21 SBK-L-20057, Response to Request for Additional Information Regarding Degraded Voltage Time Delay Setpoint License Amendment Request2020-05-20020 May 2020 Response to Request for Additional Information Regarding Degraded Voltage Time Delay Setpoint License Amendment Request 2023-06-22
[Table view] Category:Licensee 30-Day Written Event Report
MONTHYEARL-2019-058, CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report2019-03-25025 March 2019 CFR 50.46 - Emergency Core Cooling System LBLOCA 30-Day Report L-2018-063, CFR 50.46 Annual Reporting and 30-day Notification of Changes or Errors in Emergency Core Cooling System Models or Applications2018-03-26026 March 2018 CFR 50.46 Annual Reporting and 30-day Notification of Changes or Errors in Emergency Core Cooling System Models or Applications L-2017-102, Submittal of 10 CFR 50.46 - Emergency Core Cooling 30-Day Report2017-06-0202 June 2017 Submittal of 10 CFR 50.46 - Emergency Core Cooling 30-Day Report L-2016-124, 10 CFR 50.46 30-Day Report for Turkey Point, Seabrook, and Point Beach2016-06-17017 June 2016 10 CFR 50.46 30-Day Report for Turkey Point, Seabrook, and Point Beach SBK-L-14073, Best Estimate Large Break Loss of Coolant Accident 10 CFR 50.46 30-Day Report2014-04-16016 April 2014 Best Estimate Large Break Loss of Coolant Accident 10 CFR 50.46 30-Day Report SBK-L-12264, Nuclear Fuel Pellet Thermal Conductivity Degradation Impact on Current Seabrook Be LBLOCA Analysis Using the 1996 Cqd Methodology 10 CFR 50.46 30-Day Report2012-12-21021 December 2012 Nuclear Fuel Pellet Thermal Conductivity Degradation Impact on Current Seabrook Be LBLOCA Analysis Using the 1996 Cqd Methodology 10 CFR 50.46 30-Day Report SBK-L-11141, Submittal of Special Report Regarding Containment Enclosure Building Structural Integrity2011-07-11011 July 2011 Submittal of Special Report Regarding Containment Enclosure Building Structural Integrity 2019-03-25
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,#ýEA ROK December 21, 2012 Docket No. 50-443 SBK-L-12264 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001 Seabrook Station Nuclear Fuel Pellet Thermal Conductivity Degradation Impact On Current Seabrook BE LBLOCA Analysis Using the 1996 CQD Methodology 10 CFR 50.46 30-Day Report
Reference:
Letter from M. O'Keefe (NextEra) to U.S. Nuclear Regulatory Commission (SBK-L-12126, Docket No. 50-443), "Seabrook Station, Annual Reporting of Changes to, or Errors in, Emergency Core Cooling System Models or Applications,"
July 16, 2012.
In accordance with the requirements of 10 CFR 50.46(a)(3)(ii), NextEra Energy Seabrook, LLC (NextEra) submits a tabulation of the current Large Break and Small Break LOCA PCT margin utilization tables applicable to Seabrook Station. NextEra Energy Seabrook letter SBK-L-12126, documented the 2011 Annual Report of Peak Clad Temperature (PCT) for the Small Break Loss of Coolant Accident (SBLOCA) and the Large Break Loss of Coolant Accident (LBLOCA) for the Seabrook Station. The reported LBLOCA PCT and cumulative change were 1789 °F and 5 °F, respectively.
10 CFR 50.46(a)(3)(ii) requires that changes to the LBLOCA Evaluation Model (EM) and SBLOCA EM PCT exceeding 50 OF be reported to the NRC within 30 days. This 30-day report documents the results of the evaluation performed by Westinghouse Electric Company, LLC to address the effect of nuclear fuel pellet thermal conductivity degradation (TCD) on the Seabrook Station LBLOCA Code Qualified Document (CQD) EM.
Enclosure 1 summarizes the assessment made to the Seabrook Station LBLOCA CQD EM to account for TCD effects. Taýle 2 in Enclosure 1 provides the new LBLOCA PCT rackup for Seabrook Station that accounts for the TCD effects and power peaking burndown. Enclosure 2 provides supplemental information on TCD effects. The new LBLOCA predicted values for the PCT and the cumulative change are 1914 OF and 130 °F, respectively.
NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 A002.,.
U. S. Nuclear Regulatory Commission SBK-L-12264/ Page 2 10 CFR 50.46(a)(3)(ii) also requires that a schedule for reanalysis be provided or compliance with the requirements of the regulation be shown. Continued compliance with 10 CFR 50.46 requirements is demonstrated by the total estimated new PCT value of 1914 OF remaining well below the limit of 2200 OF. NextEra is continuing to monitor industry initiatives to incorporate the affects of TCD on approved EMs and will provide further updates in accordance with 10CFR50.46 as necessary.
Should you have any questions regarding this report, please contact me at (603) 773-7745.
Sincerely, NextEra Energy Seabrook, LLC Micha-i O'Keefe r Licensing Manager cc: NRC Region I Administrator J. Lamb, NRC Project Manager, Project Directorate 1-2 NRC Senior Resident Inspector
ENCLOSURE 1 TO SBK-L-12264
Background
Fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown were not explicitly considered in the Seabrook Large Break Loss-of-Coolant Accident (LBLOCA) Analysis of Record (AOR). Nuclear Regulatory Commission (NRC)
Information Notice 2011-21 (Reference 1) notified addressees of recent information obtained concerning the impact of irradiation on fuel thermal conductivity and its potential to cause significantly higher predicted peak clad temperature (PCT) results in realistic emergency core cooling system (ECCS) evaluation models. This evaluation provides an estimated effect of TCD and peaking factor burndown on the PCT calculation for the ECCS at Seabrook. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451 (Reference 2).
Affected Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect A quantitative evaluation using a method similar to that described in Reference 3 was performed to assess the PCT effect of TCD and peaking factor burndown with other considerations of burnup on the Seabrook LBLOCA analysis and concluded that the estimated PCT impact is a penalty of 125*F for Reflood 1 and a penalty of 180°F for Reflood 2 for 10 CFR 50.46 reporting purposes. With these new penalties, Reflood 1 PCT remains limiting. The peaking factor burndown included in the evaluation is provided in Table 1. NextEra and its vendor, Westinghouse Electric Company LLC, utilize processes which ensure that the LOCA analysis input values conservatively bound the as-operated plant values for those parameters.
Table 1: Peaking Factors Assumed in the Evaluation of TCD Rod Burnup FdH (1),(2) FQ Transient (1) FQ Steady-State (MWdIMTU) 0 1.65 2.50 2.00 30,000 1.65 2.50 2.00 60,000 1.40 2.00 1.60 62,000 1.40 2.00 1.60 (1) Includes uncertainties.
(2) Hot assembly average power follows the same burndown, since it is a function of FdH.
References
- 1. NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation,"
December 13, 2011. (NRC ADAMS # ML113430785)
- 2. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October 1992.
- 3. LTR-NRC-12-27, Letter from J. A. Gresham (Westinghouse) to NRC, 'Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.
Table 2 Seabrook LBLOCA PCT Rackup Peak Clad Cumulative Temperature Change LBLOCA 2011 10 CFR 50.46 Annual Report (Reference 1) 1789 OF 5 OF TCD Impact 125 0F 125 OF Reference
- 1. Letter from M. O'Keefe to the USNRC, "Seabrook Station, Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications," SBK-L-12126, Docket No.
50-443, July 16, 2012.
ENCLOSURE 2 TO SBK-L-12264 ENCLOSURE 2 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND POWER PEAKING BURNDOWN SUPPLEMENTAL INFORMATION 1.0 Background The Nuclear Regulatory Commission (NRC) approved 1996 Westinghouse Best Estimate Large Break Loss-of-Coolant Accident (LOCA) Evaluation Model (Reference 1) is based on the PAD 3.4 fuel performance code (Reference 2). Upon NRC approval of PAD 4.0 (Reference 3), its usage was extended to the 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model, as reported to the NRC in Reference 4. The Seabrook Large-Break LOCA (LBLOCA) analysis utilized fuel rod design input from PAD 4.0. PAD 3.4 and 4.0 were licensed without explicitly considering fuel pellet thermal conductivity degradation (TCD) with burnup. Explicit modeling of fuel pellet TCD in the fuel performance code leads to changes in the fuel rod design parameters beyond' beginning-of-life which are input to the LBLOCA analysis. The effects of explicitly modeling fuel pellet TCD on the Seabrook LBLOCA.analysis (Reference 5) have been considered. The modeling of fuel pellet TCD is a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-1 3451 (Reference 6).
Fuel performance data that accounts for fuel pellet TCD (using an unlicensed model) was used as input to the Seabrook evaluation. The new PAD fuel performance data was generated with a representative model that includes explicit modeling of fuel pellet TCD. Therefore, the evaluations performed consider the fuel pellet TCD effects cited in NRC Information Notice 2011-21 (Reference 7).
2.0 Large Break LOCA Input Parameters and Assumptions The evaluation of fuel TCD and peaking factor burndown considered the following input parameter changes to the LBLOCA analysis:
- Fuel rod design data with PAD 4.0 + TCD
" Peaking factor burndown shown in Table 1 Table 1: Peaking Factors Assumed in the Evaluation of TCD Rod Burnup FdH (1),(2) FQ Transient (1) FQ Steady State (MWd/MTU) 0 1.65 2.50 2.00 30,000 1.65 2.50 2.00 60,000 1.40 2.00 1.60 62,000 1.40 2.00 1.60 (1) Includes uncertainties.
(2) Hot assembly average power follows the same burndown, since it is a function of FdH.
d ENCLOSURE2 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND POWER PEAKING BURNDOWN SUPPLEMENTAL INFORMATION 3.0 Large Break LOCA Description of Evaluation A quantitative evaluation using a method similar to that described in Reference 8 was performed to assess the peak clad temperature (PCT) effect of TCD and peaking factor burndown with other considerations of burnup on the Seabrook LBLOCA analysis. However, the Seabrook TCD evaluation utilizes a different burnup selection, makes no changes to design inputs beyond peaking factor burndown, and considers both the Reference Transient as well as the Limiting Split case from the Analysis-of-Record (AOR).
In addition, a quantitative evaluation was performed for oxidation.
4.0 Large Break LOCA Results The results of the evaluation are presented in Table 2.
Table 2: Fuel Pellet TCD Evaluation Results Evaluation Reflood 1 PCT (OF) Reflood 2 PCT (*F)
Starting Rackup PCT 1789 1729 TCD (including burndown) 125 180 Final New Rackup PCT 1914 1909 With these new evaluations, acceptance criterion 10 CFR 50.46(b)(1) is met as follows:
= Calculated Peak Clad Temperature (PCT) = 1914 OF < 2200 OF A quantitative evaluation was performed for oxidation, and new MLO and CWO results are presented below. As such, acceptance criteria 10 CFR 50.46(b)(2) and (b)(3) are met as follows:
- Calculated Maximum Local Oxidation (MLO) = 7.8% < 17%
- Calculated Core Wide Oxidation (CWO) = 0.34% < 1%
Based on the PCT and oxidation results herein, and the existing grid crush calculations, it was determined that acceptance criterion 10 CFR 50.46(b)(4) "Coolable Geometry" continues to be met.
Based on the demonstration of coolable geometry and an evaluation of TCD on long term core cooling, it was concluded that acceptance criterion 10 CFR 50.46(b)(5) "Long Term Core Cooling" continues to be met with respect to this evaluation.
As a result, all of the 10 CFR 50.46 acceptance criteria are met for the Seabrook LBLOCA analysis when considering the changes for TCD evaluated herein.
ENCLOSURE2 EVALUATION OF FUEL PELLET THERMAL CONDUCTIVITY DEGRADATION AND POWER PEAKING BURNDOWN SUPPLEMENTAL INFORMATION 5.0 References
- 1. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998.
- 2. WCAP-10851-P-A, "Improved Fuel Performance Models for Westinghouse Fuel Rod Design and Safety Evaluations," August 1988.
- 3. WCAP-15063-P-A with Errata, Revision 1, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," July 2000.
- 4. LTR-NRC-01-6, "10 CFR 50.46 Annual Notification and Reporting for 2000," March 13, 2001.
- 5. WCAP-16254-P, "Best-Estimate Analysis of the Large Break Loss of Coolant Accident for the Seabrook Station," May 2004.
- 6. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October 1992.
- 7. NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011. (NRC ADAMS # ML113430785)
- 8. LTR-NRC-12-27, "Westinghouse Input Supporting Licensee Response to NRC 10CFR50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," March 7, 2012.