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Category:Letter type:L
MONTHYEARL-2024-132, 2024 Population Update Analysis2024-08-13013 August 2024 2024 Population Update Analysis L-2024-129, Relief Request (RR) 14. Limited Coverage Exams Due to Impractical Inservice Inspection Requirements - Fourth Ten-Year Inservice Inspection Program Interval2024-08-0707 August 2024 Relief Request (RR) 14. Limited Coverage Exams Due to Impractical Inservice Inspection Requirements - Fourth Ten-Year Inservice Inspection Program Interval L-2024-121, Subsequent License Renewal Commitment 30 Revision2024-07-30030 July 2024 Subsequent License Renewal Commitment 30 Revision L-2024-123, Submittal of In-Service Inspection Program Owners Activity Report (OAR-1)2024-07-29029 July 2024 Submittal of In-Service Inspection Program Owners Activity Report (OAR-1) L-2024-125, Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes2024-07-24024 July 2024 Notice of Intent to Provide Supplemental Information to License Amendment Request to Adopt Common Emergency Plan with Site-Specific Annexes L-2024-110, Environmental Protection Plan Report, Unusual or Important Environmental Event - Manatee in Intake2024-07-10010 July 2024 Environmental Protection Plan Report, Unusual or Important Environmental Event - Manatee in Intake L-2024-114, Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal2024-07-10010 July 2024 Quality Assurance Topical Report (FPL-1 Revision 31 Annual Submittal L-2024-109, Schedule for Subsequent License Renewal Environmental Review2024-07-0303 July 2024 Schedule for Subsequent License Renewal Environmental Review L-2024-104, Response to Request for Additional Information, St. Luce Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1 Extension of Inspection Interval for Reactor Pressure Vessel Welds from 102024-06-26026 June 2024 Response to Request for Additional Information, St. Luce Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1 Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 L-2024-097, Technical Specification Special Report2024-06-20020 June 2024 Technical Specification Special Report L-2024-102, Official Service List Update2024-06-19019 June 2024 Official Service List Update L-2024-090, Revised Steam Generator Tube Inspection Reports2024-06-0404 June 2024 Revised Steam Generator Tube Inspection Reports L-2024-075, Notification of Improved Standard Technical Specifications (ITS) Implementation2024-05-13013 May 2024 Notification of Improved Standard Technical Specifications (ITS) Implementation L-2024-053, License Amendment Request L-2024-053, Updated Spent Fuel Pool Criticality Analysis2024-04-30030 April 2024 License Amendment Request L-2024-053, Updated Spent Fuel Pool Criticality Analysis L-2024-071, Cycle 27 Core Operating Limits Report2024-04-29029 April 2024 Cycle 27 Core Operating Limits Report L-2024-070, Cycle 32 Core Operating Limits Report2024-04-29029 April 2024 Cycle 32 Core Operating Limits Report L-2024-056, Annual Radiological Environmental Operating Report for Calendar Year 20232024-04-17017 April 2024 Annual Radiological Environmental Operating Report for Calendar Year 2023 L-2024-064, Florida Power & Light Company - 10 CFR 50.46 - Emergency Core Cooling System SBLOCA 30-Day Report2024-04-17017 April 2024 Florida Power & Light Company - 10 CFR 50.46 - Emergency Core Cooling System SBLOCA 30-Day Report L-2024-054, 2023 Annual Environmental Operating Report2024-04-0909 April 2024 2023 Annual Environmental Operating Report L-2024-047, Proposed Use of a Subsequent ASME Code Edition and Addenda2024-03-28028 March 2024 Proposed Use of a Subsequent ASME Code Edition and Addenda L-2024-045, Report of 10 CFR 72.48 Plant Changes2024-03-27027 March 2024 Report of 10 CFR 72.48 Plant Changes L-2024-011, and Point Beach, Units 1 and 2 - 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications2024-03-13013 March 2024 and Point Beach, Units 1 and 2 - 10 CFR 50.46 Annual Reporting of Changes to, or Errors in Emergency Core Cooling System Models or Applications L-2024-023, Unusual or Important Environmental Event - Turtle Mortality2024-03-0606 March 2024 Unusual or Important Environmental Event - Turtle Mortality L-2024-015, 2023 Annual Radioactive Effluent Release Report2024-02-29029 February 2024 2023 Annual Radioactive Effluent Release Report L-2024-029, 2023 Annual Operating Report2024-02-28028 February 2024 2023 Annual Operating Report L-2024-026, Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules2024-02-27027 February 2024 Revised Reactor Vessel Materials Surveillance Capsule Withdrawal Schedules L-2024-010, Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3)2024-01-25025 January 2024 Point Units 3 and 4, Seabrook, Duane Arnold, and Point Beach Units 1 and 2, Nuclear Property Insurance - 10 CFR 50.54(w)(3) L-2024-004, Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 to 20 Years2024-01-18018 January 2024 Relief Request (RR) 7, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) Extension of Inspection Interval for Reactor Pressure Vessel Welds from 10 to 20 Years L-2024-002, Withdrawal of Proposed Alternative to American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump2024-01-0808 January 2024 Withdrawal of Proposed Alternative to American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump L-2023-173, Quality Assurance Topical Report (FPL-1) Revision 30 Update2023-12-15015 December 2023 Quality Assurance Topical Report (FPL-1) Revision 30 Update L-2023-179, Unusual or Important Environmental Event - Turtle Mortality2023-12-14014 December 2023 Unusual or Important Environmental Event - Turtle Mortality L-2023-168, License Amendment Request Supplement to Revision 2 for the Technical Specifications Conversion to NUREG-1432 Revision 52023-12-12012 December 2023 License Amendment Request Supplement to Revision 2 for the Technical Specifications Conversion to NUREG-1432 Revision 5 L-2023-155, Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-06542023-11-28028 November 2023 Supplement to Response to Request for Additional Information, Revised NextEra Common Emergency Plan, and Revised Site-Specific Emergency Plan Annexes Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, L-2023-162, Response to 50.69 2nd Round of Rals2023-11-21021 November 2023 Response to 50.69 2nd Round of Rals L-2023-131, Subsequent License Renewal Application - Second Annual Update2023-09-28028 September 2023 Subsequent License Renewal Application - Second Annual Update L-2023-136, Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-26026 September 2023 Supplement to License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2023-122, Corrections to the 2022 Annual Radiological Environmental Operating Report2023-09-20020 September 2023 Corrections to the 2022 Annual Radiological Environmental Operating Report L-2023-127, Correction to the 2022 Annual Radioactive Effluent Release Report2023-09-18018 September 2023 Correction to the 2022 Annual Radioactive Effluent Release Report L-2023-113, Correction to the 2020 Annual Radiological Environmental Operating Report2023-09-14014 September 2023 Correction to the 2020 Annual Radiological Environmental Operating Report L-2023-118, Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors2023-09-11011 September 2023 Response to Request for Additional Information Regarding License Amendment Request to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors L-2023-108, Report of 10 CFR 50.59 Plant Changes2023-09-11011 September 2023 Report of 10 CFR 50.59 Plant Changes L-2023-107, Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.42023-09-0606 September 2023 Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.j.4 L-2023-112, Corrections to the 2021 Annual Radioactive Effluent Release Report2023-09-0606 September 2023 Corrections to the 2021 Annual Radioactive Effluent Release Report L-2023-114, Proposed Turkey Point Units 6 and 7; Seabrook Station; Point Beach Units 1 and 2 - Official Service List Update2023-08-17017 August 2023 Proposed Turkey Point Units 6 and 7; Seabrook Station; Point Beach Units 1 and 2 - Official Service List Update L-2023-098, and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 22023-08-0707 August 2023 and Point Beach Units 1 and 2 - Response to Request for Additional Information Regarding License Amendment Request for Common Emergency Plan Consistent with NUREG-0654, Revision 2 L-2023-105, Preparation and Scheduling of Operator Licensing Examinations2023-08-0303 August 2023 Preparation and Scheduling of Operator Licensing Examinations L-2023-102, Relief Request PSL2-15-RR-01, Proposed Alternative to ASME Section XI Code Examination Requirements for Reactor Vessel Bottom Area and Piping in Covered Trenches2023-07-26026 July 2023 Relief Request PSL2-15-RR-01, Proposed Alternative to ASME Section XI Code Examination Requirements for Reactor Vessel Bottom Area and Piping in Covered Trenches L-2023-099, Pump Relief Request 10 (PR-10), One-Time Request for an Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump2023-07-26026 July 2023 Pump Relief Request 10 (PR-10), One-Time Request for an Alternative to the American Society of Mechanical Engineers (ASME) Operation and Maintenance (OM) Code for the Auxiliary Feedwater (AFW) 2C Pump L-2023-097, Subsequent License Renewal Application Revision 1 - Supplement 62023-07-13013 July 2023 Subsequent License Renewal Application Revision 1 - Supplement 6 L-2023-076, In-Service Inspection Program Owner'S Activity Report (OAR-1)2023-07-11011 July 2023 In-Service Inspection Program Owner'S Activity Report (OAR-1) 2024-08-07
[Table view] Category:Licensee 30-Day Written Event Report
MONTHYEARL-2019-051, Unusual or Important Environmental Event - Turtle Mortalities - February 4, 2019 and February 11, 20192019-02-27027 February 2019 Unusual or Important Environmental Event - Turtle Mortalities - February 4, 2019 and February 11, 2019 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-2018-005, Submittal of Emergency Core Cooling 30-Day Report2018-01-12012 January 2018 Submittal of Emergency Core Cooling 30-Day Report L-2014-248, Technical Specification Special Report Radiation Monitor Inoperable Greater than 72 Hours2014-07-28028 July 2014 Technical Specification Special Report Radiation Monitor Inoperable Greater than 72 Hours L-2014-134, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Change Report2014-05-12012 May 2014 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Change Report L-2014-013, Submittal of Technical Specification Special Report2014-01-13013 January 2014 Submittal of Technical Specification Special Report L-2011-345, Technical Specification Special Report, Radiation Monitor Inoperable Greater than 72 Hours2013-12-23023 December 2013 Technical Specification Special Report, Radiation Monitor Inoperable Greater than 72 Hours L-2013-291, Submittal of Technical Specification Special Report Radiation Monitor Inoperable Greater than 72 Hours2013-10-0101 October 2013 Submittal of Technical Specification Special Report Radiation Monitor Inoperable Greater than 72 Hours L-2013-217, Revised Technical Specification Special Report, Failure of Channel a Reactor Vessel Level Monitoring System (Rvlms) on June 4, 20132013-07-0303 July 2013 Revised Technical Specification Special Report, Failure of Channel a Reactor Vessel Level Monitoring System (Rvlms) on June 4, 2013 L-2013-215, Technical Specification Special Report - Failure of Channel a Reactor Vessel Level Monitoring System (Rvlms)2013-06-0404 June 2013 Technical Specification Special Report - Failure of Channel a Reactor Vessel Level Monitoring System (Rvlms) L-2010-139, LER, St. Lucie, Unit 1, Submittal of Technical Specification Special Report, Inoperable Containment Sump Wide Range Level Channel B2010-06-28028 June 2010 LER, St. Lucie, Unit 1, Submittal of Technical Specification Special Report, Inoperable Containment Sump Wide Range Level Channel B L-2009-289, Submittal of Technical Specification Special Report Reactor Vessel Level Monitoring System a Channel Level Probe (Hjtc) Out of Service2009-12-14014 December 2009 Submittal of Technical Specification Special Report Reactor Vessel Level Monitoring System a Channel Level Probe (Hjtc) Out of Service L-2009-250, Date of Event: October 6, 2009, Technical Specification Special Report, Sodium Hypochlorite Spill Greater than 100 Pounds2009-11-0505 November 2009 Date of Event: October 6, 2009, Technical Specification Special Report, Sodium Hypochlorite Spill Greater than 100 Pounds L-2009-097, LER for St. Lucie, Unit 2 Regarding Technical Specification Special Report, Radiation Monitor Inoperable Greater than 72 Hours2009-04-20020 April 2009 LER for St. Lucie, Unit 2 Regarding Technical Specification Special Report, Radiation Monitor Inoperable Greater than 72 Hours L-2007-166, Technical Specification Special Report, Radiation Monitoring Inoperable Greater than 72 Hours2007-10-18018 October 2007 Technical Specification Special Report, Radiation Monitoring Inoperable Greater than 72 Hours L-2007-100, Technical Specification Special Report, Pressurizer Power Operate Relief Valve (PORV)2007-06-13013 June 2007 Technical Specification Special Report, Pressurizer Power Operate Relief Valve (PORV) L-2004-291, Technical Specification Special Report on Inoperable Containment Sump Wide Range Level Channel B2004-12-22022 December 2004 Technical Specification Special Report on Inoperable Containment Sump Wide Range Level Channel B L-2004-052, Technical Specification Special Report Regarding Radwaste Building Exhaust System Plant Ventilation Exhaust Monitor - Out-of-Service2004-02-23023 February 2004 Technical Specification Special Report Regarding Radwaste Building Exhaust System Plant Ventilation Exhaust Monitor - Out-of-Service L-2003-196, Special Report Re Inoperable Loose Parts Monitoring Channel, Per Requirements of Updated Final Safety Analyses Report Section 13.7.4.1 Action a2003-07-23023 July 2003 Special Report Re Inoperable Loose Parts Monitoring Channel, Per Requirements of Updated Final Safety Analyses Report Section 13.7.4.1 Action a ML18113A2701977-02-25025 February 1977 Forwards Non-Routine Radioactive Effluent Report ML18113A2861977-01-20020 January 1977 Forwards Non-Routine Environmental Operating Report Titled Impingement of Aquatic Organisms ML18088A9381976-07-29029 July 1976 Letter Attaching Reportable Occurrence 335-B-76-03 St. Lucie Unit 1 Date of Occurrence: July 1, 1976 Volume Control Tank. L-76-247, Letter Attaching Reportable Occurrence No. 335-B-76-02 Failure to Determine Condenser Delta T.1976-07-0707 July 1976 Letter Attaching Reportable Occurrence No. 335-B-76-02 Failure to Determine Condenser Delta T. 2019-02-27
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Text
FPL JAN 1 2 2018 L-2018-005 10 CFR 50.46(a)(3)(ii)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Florida Power & Light Company St. Lucie Units 1 and 2
. Docket Nos. 50-335, 50-389 10 CFR 50.46 -Emergency Core Cooling 30-Day Report In accordance with 10 CFR 50.46(a)(3)(ii), Florida Power & Light Company (FPL) is submitting this 30-day report of an error affecting the St. Lucie Nuclear Plant (PSL), Units 1 and 2 for the emergency core cooling system (ECCS) analysis performed by AREVA Incorporated.
FPL was informed by AREVA of an error affecting the oxidation calculation in the small break loss of coolant accident (SBLOCA) analyses ofrecord. Enclosure 1 to this letter contains a description and impact of the error for Unit 1, and Enclosure 2 is provided for Unit 2.
Please contact Larry Nicholson at (561) 304-6224 should you have any questions regarding this submittal.
.:ini~
Larry ~holson Director, Nuclear Licensing and Regulatory Compliance Enclosures (2)
Florida Power & Light Company 700 Universe Boulevard, Juno Beach, FL 33408
St. Lucie Units 1 and 2 L-2018-005 Docket No. 50-335, 50-389 Enclosure 10 CFR 50.46 30-Day Report Page 1 of 4 Enclosure 1 St. Lucie Unit 1 10 CPR 50.46 30-Day Report Emergency core cooling system (ECCS) small break loss of coolant accident (SBLOCA) analysis for St.
Lucie Unit 1 is performed by AREVA Inc. The following 30-Day report, pertaining to the application of the AREVA SBLOCA evaluation model, is provided pursuant to 10 CPR 50.46(a)(3)(ii). A summary of the model changes/errors and the estimated impact on the peak cladding temperature tPCT) for SBLOCA is provided in Table 1.
1.0 SBLOCA Model Changes o S-RELAP5 Calculations of Oxidation Due to High Temperature Metal-Water Reaction AREVA determined that there is an error in the oxidation model related to high temperature metal-water reaction. In a LOCA event, the cladding can swell (and potentially rupture) due to high difference in pressure between the fuel and the system which causes the clad to thin. The clad radius increases, while the thickness decreases. It was discovered that the S-RELAP5 oxidation calculations used cold cladding dimensions and therefore did not fully account for the swelling phenomena. The error can lead to an under-prediction of the oxidation and heat from the metal-water reaction.
- The PCT impact of the oxidation error is estimated to be +2 °P for St. Lucie Unit 1.
Previous SBLOCA PCT changes are documented in Reference 2.1. Table 1 summarizes the estimated impact of the changes/errors on the St. Lucie Unit 1 SBLOCA PCT, along with the cumulative effect of absolute PCT changes.
The cumulative impact for the SBLOCA evaluation model PCT is 84 °P. The final PCT including all changes/errors, is 1852 °P, which maintains significant margin to the limit of2200 °F.
2.0 References 2.1 L-2017-157, "St. Lucie Unit 1 Docket No. 50-335 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CPR 50.46 30-Day Report," September 13, 2017.
(MLl 7258A037)
St. Lucie Units 1 and 2 L-2018-005 Docket No. 50-335, 50-389 Enclosure 10 CFR 50.46 30-Day Report Page 2 of 4 Enclosure 1 Table 1 St. Lucie Unit 1 SBLOCA Margin Summary Sheet - 30 Day Report Plant Name: St. Lucie Unit 1 Utility Name: NextEra Energy Evaluation Model: EMF-2328(P)(A) Rev. 0 as supplemented by ANP-3000(P), Revision 0.
Evaluation Model PCT: 1828°P Net PCT Absolute PCT Effect Effect Prior 10 CPR 50.46 Changes or Error Corrections -
A +22 °P 82 °P up to Year 2016 Prior 10 CPR 50.46 Changes or Error Corrections - 0 op 0 op B
Reported in Year 2017 C 10 CPR 50.46 Changes in Year 2017 Since Item B Error in oxidation calculations due to high +2 op 2 op temperature metal-water reaction D Sum of 10 CPR 50.46 Changes +24 °P 84 °P The sum of the PCTfrom the most recent analysis using an acceptable evaluation model and the estimates of PCT 1852 °F < 2200 °F impact for changes and errors identified since this analysis
St. Lucie Units 1 and 2 L-2018-005 Docket No. 50-335, 50-389 Enclosure 10 CFR 50.46 30-Day Report Page 3 of4 Enclosure 2 St. Lucie Unit 2 10 CFR 50.46 30-Day Report Emergency core cooling system (ECCS) small break loss of coolant accident (SBLOCA) analysis for St.
Lucie Unit 2 is performed by AREVA Inc. The following 30-Day report, pertaining to the application of the AREVA SBLOCA evaluation model, is provided pursuant to 10 CFR 50.46(a)(3)(ii). A summary of the model changes/errors and the estimated impact on the peak cladding temperature (PCT) for SBLOCA is provided in Table 1.
1.0 SBLOCA Model Changes o M5 LOCA Swelling and Rupture Model (SRM) Update AREVA determined that certain aspects of the SRM model would be impacted when considering additional M5 cladding rupture test data since the model's approval by the NRC as part of M5 licensing. Following the same approach as the original model, an updated M5 SRM was developed to take into account the updated test database.
- The PCT impact of the updated model is estimated to be O °F for St. Lucie Unit 2.
o S-RELAP5 Calculations of Oxidation Due to High Temperature Metal-Water Reaction AREVA determined that there is an error in the oxidation model related to high temperature metal-water reaction. In a LOCA event, the cladding can swell (and potentially rupture) due to high difference in, pressure between the fuel and the system which causes the clad to thin. The clad radius increases, while the thickness decreases. It was discovered that the S-RELAP5 oxidation calculations used cold cladding dimensions and therefore did not fully account for the swelling phenomena. The error can lead to an under-prediction of the oxidation and heat from the metal-water reaction.
- The PCT impact of the oxidation error is estimated to be +57 °F for St. Lucie Unit 2.
o Change in Analysis HPSI Flow Assumption In order to offset some of the impact of the aforementioned error, a change is made to the HPSI flow to credit some of the available conservatism in the HPSI flow, while still maintaining adequate margin to the pump test criteria.
- The PCT impact of the HPSI flow change is estimated to be -336 °F for St. Lucie Unit 2.
Previous SBLOCA PCT changes are documented in Reference 2.1. Table 1 summarizes the estimated impact of the changes/errors on the St. Lucie Unit 2 SBLOCA PCT, along with the cumulative effect of absolute PCT changes.
The cumulative impact for the SBLOCA evaluation model PCT is 393 °F. The final PCT including all changes/errors, is 1778 °F, which maintains significant margin to the limit of 2200 °F.
2.0 References 2.1 . L-2017-056, "St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors 10 CFR 50.46 Annual Report," March 27, 2017. (MLl 7086A321).
St. Lucie Units 1 and 2 L-2018-005 Docket No. 50-335, 50-389 Enclosure 10 CFR 50.46 30-Day Report Page 4 of 4 Enclosure 2 Table 1 St. Lucie Unit 2 SBLOCA Margin Summary Sheet- 30 Day Report Plant Name: St. Lucie Unit 2 Utility Name: NextEra Energy Evaluation Model: EMP-2328(P)(A) Rev. 0 Evaluation Model PCT: 2057°P Net PCT Absolute PCT Effect Effect Prior 10 CPR 50.46 Changes or Error Corrections - 0 op 0 op A
up to Year 2016 Prior 10 CPR 50.46 Changes or Error Corrections-B None None Reported in Year 2017 C 10 CPR 50.46 Changes in Year 2017 Since Item B SRM Model Update due to new rupture test data 0 op 0 op Error in oxidation calculations due to high
+57 °P 57 °P temperature metal-water reaction Change in analysis HPSI flow assumption -336 °P 336 °P D Sum of 10 CPR 50.46 Changes -279°P 393 °P The sum of the PCTfrom the most recent analysis using an acceptable evaluation model and the estimates ofPCT 1778 °F < 2200 °F impact for changes and errors identified since this analysis