NL-17-116, Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report

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Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report
ML17261A191
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 09/11/2017
From: Vitale A
Entergy Nuclear Northeast
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-116
Download: ML17261A191 (18)
v • d • e


Text

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Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Tel (914) 254-6700 Anthony J Vitale Site Vice President NL-17-116 September 11, 2017 U.S. Nuclear Regulatory Commiss.ion ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

2016 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report Indian Point Unit Nos. 2 and 3 Docket Nos. 50-247 & 50-286 License Nos. DPR-26 and DPR-64

REFERENCE:

1) Entergy letter to NRC, "2015 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report," dated July 11, 2016 (NL 079)

Dear Sir or Madam:

This letter provides the annual 1 O CFR 50.46 report for calendar year 2016 for Indian Point Units 2 (IP2) and 3 (IP3). The prior report for calendar year 2015 was provided in Reference 1.

There are no new peak cladding temperature (PCT) adjustments required for the IP2 or the IP3 small and large break loss-of-coolant accident (LOCA) analyses of record (AOR) for calendar year 2016. Although not required by regulation, this letter reports the small break LOCA PCT for IP2 as 1028 °F and IP3 as 1543 °F, and the large break LOCA PCT for IP2 as 2119 °F and IP3 as 2046 °F, respectively.

Four separate changes, error corrections or enhancements resulting in an estimated PCT impact of 0 °F were reported by Westinghouse Electric Company, LLC (Westinghouse),

concerning the Indian Point Units 2 and 3 large break loss of coolant (LBLOCA) evaluation models as follows:

1) General Code Maintenance (Indian Point Units 2 and 3)
2) Error in Oxidation Calculations (Indian Point Units 2 and 3)
3) Error in Use of ASME Steam Tables (Indian Point Unit 2)
4) Evaluation of Increased Containment Metal due to Control Rod Drive Shafts and Water Shields (Indian Point Unit 2)

These changes are further described in Enclosure 1.

NL-17-116 Docket Nos. 50-247 and 50-286 Page 2 of 2 The limiting PCT results for the IP2 and IP3 small and large break LOCAs continue to meet the criteria of 10 CFR 50.46, paragraph (b) for calendar year 2016.

There are no new commitments being made in this submittal. If you have any questions or require additional information, please contact Mr. Robert W. Walpole, Regulatory Assurance Manager at (914) 254-6710.

Sincerely, AJV/rl

Enclosure:

1. Westinghouse Letter LTR-LIS-17-36, "10 CFR 50.46 Annual Notification and Reporting for 2016" cc: Mr. Daniel H. Dorman, Regional Administrator, NRC Region I Mr. Richard V. Guzman, NRR Senior Project Manager Ms. Bridget Frymire, New York State Department of Public Service Ms. Alicia Barton, President and CEO NYSERDA NRC Resident Inspector's Office

ENCLOSURE 1 TO NL-17-116 Westinghouse Letter LTR-LIS-17-36 10 CFR 50.46 Annual Notification and Reporting for 2016 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286

Westinghouse Non-Proprietary Class 3

@Westinghouse Westinghouse Electric Company Engineering Center of Excellence 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Directtel: (412)374-5598 e-mail: mcmillh@westinghouse.com Ourref: LTR-LIS-17-36 February 8, 2017 Indian Point Units 2 and 3 10 CFR 50.46 Annual Notification and Reporting for 2016

Dear Sir or Madam:

This is a notification of 10 CFR 50.46 reporting information pertaining to the Westinghouse Electric Company Evaluation Models/analyses. As committed to in WCAP-13451, Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting, Westinghouse is providing an Annual Report for Emergency Core Cooling System (ECCS)

Evaluation Model changes and errors for the 2016 model year. All necessary standardized reporting pages for any changes and errors for the Evaluation Models utilized for your plant(s) are enclosed, consistent with the commitment following the NUPIC audit in early 1999. Peak Clad Temperature (PCT) sheets are enclosed. All necessary revisions for any non-zero, non-discretionary PCT changes have been included. Non-discretionary PCJ impacts of 0°F will generally not be presented on the PCT sheet. The Evaluation Model changes and errors (except any plant-specific errors in the application of the model) will be provided to the NRC via Westinghouse letter.

This information is for your use in making a determination relative to the reporting requireme1'.ts of 10 CFR 50.46.

The information that is provided in this letter was prepared in accordance with Westinghouse's Quality Management System (QMS). Please contact your LOCA plant cognizant engineer (PCE), Carmen Teolis (412-374-2202), if there are any questions concerning this information.

Author: (Electronically Approved)* Verified: (Electronically Approved)*

Heather McMillen Carmen D. Teolis LOCA Integrated Services II LOCA Integrated Services I Approved: (Electronically Approved)*

Amy J. Colussy Manager, LOCA Integrated Services I

Attachment:

10 CFR 50.46 Reporting Text and PCT Summary Sheets (14 Pages)

  • Electronically approved records are authenticated in the electronic document management system.

© 2017 Westinghouse Electric Company LLC All Rights Resen*ed

Attachment to LTR-LIS-17-36 February 8, 2017 Page 1 of 14 GENERAL CODE MAINTENANCE

Background

Various changes have been made to enhance the usability of codes and to streamline future analyses.

Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding. These changes represent Discretionary Changes that will be implemented on a forward-fit basis in accordance with Section 4.1.1 ofWCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best-Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

Attachment to LTR-LIS-17-36 February 8, 2017 Page 2of14 ERROR IN OXIDATION CALCULATIONS

Background

A closely-related group of errors were discovered in the WCOBRA/TRAC software program. The errors are related to the calculation of high temperature oxidation within a realistic large break loss-of-coolant accident (LOCA) calculation. This issue has been evaluated to estimate the impact on the Automated Statistical Treatment of Uncertainty Method (ASTRUM) and the Best-Estimate (BE) Large-Break Loss-of-~oolant Accident (LBLOCA) licensing-basis analysis results. The resolution of this issue represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best-Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Ev:aluation Model Using ASTRUM Estimated Effect It was determined that correcting the high temperature oxidation calculation in WCOBRA/TRAC is estimated to have a negligible impact on the BE LBLOCA peak cladding temperature (PCT) analysis results, leading to an estimated PCT impact of 0°F for IO CFR 50.46 reporting purposes.

I .

Attachment to LTR-LIS-17-36 February 8, 2017 Page 3of14 ERROR IN USE OF ASME STEAM TABLES

Background

The American Society of Mechanical Engineers (ASME) steam tables are used to calculate the steady-state upper head liquid temperature as a function of the pressure and specific enthalpy in the ASTRUM software program. The steam table applicable to steam/gas is used to determine the upper head fluid temperature. However, the water in the upper head is in the subcooled liquid state during normal operation (and the steady-state calculation). Therefore, the steam table applicable to liquid should be used to determine the upper head fluid temperature. This issue has been evaluated to estimate the impact on Automated Statistical Treatment of Uncertainty Method (ASTRUM) Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) analysis results. The resolution of this issue represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.

Affected Evaluation Model(s) 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect It was determined that the temperatures calculated by the ASME steam tables applicable to the steam/gas side and the liquid side are very similar within the typical upper head pressure and liquid specific enthalpy ranges. Therefore, this error was evaluated to have a negligible impact on the ASTRUM BE LBLOCA analysis results, leading to an estimate~ PCT impact of 0°F for 10 CFR 50.46 reporting purposes.

Attachment to L TR-LIS-17-36 February 8, 2017 Page 4of14 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Plant Name: Indian Point Unit 2 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Analysis Information EM: ASTRUM (2004) Analysis Date: 2/15/2005 Limiting Break Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: 15x15 Upgraded SGTP(o/o): 10 Notes: See note (b) for current peaking factor and SGTP limitations.

Clad Temp {°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1962 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

. HOTSPOT Fuel Relocation Error 0 3 2 . Evaluation of Pellet Thermal Conductivity Degradation and Peaking 209 4,5 (a)

Factor Burndown 3 . Revised Heat Transfer Multiplier Distributions -32 6 4 . Error in Burst Strain Application 38 7 B. PLANNED PLANT MODIFICATION EVALUATIONS

. Bent Fuel Assembly Alignment Pins 5 2 . Changes to Containment Sump Strainer Evaluation 0 2 3 . Evaluation of Design Input Changes with Respect to Plant Operation -63 4,5 (a, b) 4 . Evaluation oflncreased Containment Metal due to Control Rod Drive 0 8 (c)

Shafts and Water Shields C. 2016 ECCS MODEL ASSESSMENTS 1 . None 0 D.OTHER*

1 . None 0 LICENSING BASIS PCT+ PCT ASSESSMENTS PCT= 2119 It is recommended that the licensee determine if these PCT allocations should be considered with respect to 10 CFR 50.46 reporting requirements.

References 1 . WCAP-16405-P, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for Indian Point Unit 2 Nuclear Plant Using the ASTRUM Methodology," May 2005.

2 . LTR-LIS-06-299, "Evaluation of Sump Strainer Modification on Indian Point Unit 2 (IPP) Best Estimate Large Break LOCA Analyses and Transmittal of Revised PCT Sheets," May 2006.

3 . LTR-LIS-07-379, "IO CFR 50.46 Reporting Text for HOTSPOT Fuel Relocation Error for Indian Point 2," June 2007.

4 . 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),"

March2012.

5 . NF-ECH-12-23, "Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown Including Design Input Changes for Indian Point Unit 2," May 2012.

Attachment to LTR-LIS-17-36 February 8, 2017 Page 5of14 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Plant Name: Indian Point Unit 2 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 6 . LTR-LIS-13-350, "Indian Point Units 2 and 3 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions,"

July 2013 .

7 . LTR-LIS-14-34, "Indian Point Units 2 and 3 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction,"

January 2014 .

8 . IPP-16-20, "Evaluation of Additional Containment Metal on the Indian Point Unit 2 (!PP) Best-Estimate Large Break LOCA (BE LBLOCA) Analysis," March 2016.

Notes:

(a) These assessments are coupled via an evaluation ofbumup effects which include thermal conductivity degradation, peaking factor burndown and design input changes.

(b) Design input changes were a reduction in FQ(tr) from 2.5 to 2.3; FQ(ss) from 2.0 to 1.8, F~H from 1.7 to 1.65 and a corresponding reduction in Pbar-HA, and maximum steam generator tube plugging from 10% to 5%. These peaking factor limits and steam generator tube plugging limit supersede the values cited for the analysis-of-record.

(c) The evaluation of the added control rod drive shafts and water shields takes credit for margin in several post-analysis-of-record evaluations.

Attachment to LTR-LIS-17-36 February 8, 2017 Page 6of14 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Indian Point Unit 2 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Analysis Information EM: NOTRUMP Analysis Date: 1/11/2004 Limiting Break Size: 3 inch FQ: 2.5 FdH: 1.7 Fuel: 15xl5 Upgraded SGTP(%): IO Notes: None Clad Temp {°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1028 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . None 0 C. 2016 ECCS MODEL ASSESSMENTS I . None 0 D.OTHER*

I . None 0 LICENSING BASIS PCT+ PCT ASSESSMENTS PCT= 1028 It is recommended that the licensee determine if these PCT allocations should be considered with respect to I 0 CFR 50.46 reporting requirements.

References I . WCAP-16157-P, "Indian Point Nuclear Generating Unit No. 2 Stretch Power Uprate NSSS and BOP Licensing Report,"

January 2004.

Notes:

None

Attachment to LTR-LIS-17-36 February 8, 2017 Page 7of14 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Composite Analysis Information EM: CQD (1996) Analysis Date: 1/23/2004 Limiting Break Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: OFAw/IFMs SGTP (%): 10 Notes: Analysis also supports 15x15 Upgraded Fuel (Reference 1). See note (c) for current peaking factor Clad Temp {°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1944 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

. Revised Blowdown Heatup Uncertainty Distribution 5 3 2 . HOTSPOT Fuel Relocation Error 0 6 (a) 3 . Evaluation of Pellet Thermal Conductivity Degradation and Peaking 185 6 (b)

Factor Burndown 4 . Revised Heat Transfer Multiplier Distributions .35 7 5 . Error in Burst Strain Application 25 8 B. PLANNED PLANT MODIFICATION EVALUATIONS

. Bent Fuel Assembly Alignment Pins 5 2 2 . Evaluation of Increased Containment Sump Strainer Metal and 80°F 12 4 Initial Containment/Accumulator Temperature 3 . Thimble Plug Removal 0 5 4 . Design Input Changes with Respect to Plant Operation .95 6 (b, c)

C. 2016 ECCS MODEL ASSESSMENTS 1 . None 0 D.OTHER*

1 . None 0 LICENSING BASIS PCT+ PCT ASSESSMENTS PCT= 2046 It is recommended that the licensee determine if these PCT allocations should be considered with respect to 10 CFR 50.46 reporting requirements.

References 1 . WCAP* 16178-P, "Best-Estimate Analysis of the Large Break Loss of Coolant Accident for Indian Point Unit 3 Nuclear Plant Stretch Power Uprate," March 2004.

2 . IPP-04-126,\"Revised PCT Rackups for 10 CFR 50.46 Annual Notification and Reporting for 2003," October 2004.

3 . INT-05-15, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.

4 . INT-08-14, Revision I, "Evaluation of Containment Sump Strainer Modifications, Reduced Containment Initial Temperature and Accumulator Water Temperature on Indian Point Unit 3 (INT) Best Estimate Large Break LOCA Analysis," February 2010.

5 . INT-10-15, "10 CFR 50.46 Report for Thimble Plug Removal," December 2010.

Attachment to LTR-LIS-17-36 February 8, 2017 Page 8of14 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Composite 6 . NF-ECH-12-24, "Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Bumdown Including Design Input Changes for Indian Point Unit 3," May 2012 .

7 . LTR-LIS-13-350, "Indian Point Units 2 and 3 I 0 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions,"

July 2013 .

8 . LTR-LIS-14-34, "Indian Point Units 2 and 3 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction,"

January 2014.

Notes:

(a) The magnitude of this assessment was changed from prior versions of the PCT rackup sheet due to a recalculation in Reference 6.

(b) These assessments are coupled via an evaluation of bumup effects which include thermal conductivity degradation, peaking factor bumdown and design input changes.

(c) Design input changes were a reduction in FQ(tr) from 2.5 to 2.3, FQ(ss) from 2.0 to 1.8, Fi'1H from 1.7 to 1.65, and a corresponding reduction in Pbar-HA. These peaking factor limits supersede the values cited for the analysis-of-record.

Attachment to LTR-LIS-17-36 February 8, 2017 Page 9of14 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Reflood 1 Analysis Information EM: CQD (1996) Analysis Date: 1/23/2004 Limiting Break Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: OFAw/IFMs SGTP(%): IO Notes: Analysis also supports 15x15 Upgraded Fuel (Reference 1). See note (c) for current peaking factor Clad Temp {°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1904 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

. Revised Blowdown Heatup Uncertainty Distribution 5 3 2 . HOTSPOT Fuel Relocation Error 0 6 (a) 3 . Evaluation of Pellet Thermal Conductivity Degradation and Peaking 205 6 (b)

Factor Burndown 4 . Revised Heat Transfer Multiplier Distributions 5 7 5 . Error in Burst Strain Application 20 8 B. PLANNED PLANT MODIFICATION EVALUATIONS

. Bent Fuel Assembly Alignment Pins 5 2 2 . Evaluation oflncreased Containment Sump Strainer Metal and 80°F 3 4 Initial Containment/Accumulator Temperature 3 . Thimble Plug Removal 0 5 4 . Design Input Changes with Respect to Plant Operation -130 6 (b, c)

C. 2016 ECCS MODEL ASSESSMENTS l . None 0 D.OTHER*

l . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 2017

  • It is recommended that the licensee determine if these PCT allocations should be considered with respect to 10 CFR 50.46 reporting requirements.

References l . WCAP-16178-P, "Best-Estimate Analysis of the Large Break Loss of Coolant Accident for Indian Point Unit 3 Nuclear Plant Stretch Power Uprate," March 2004.

2 . IPP-04-126, "Revised PCT Rackups for 10 CFR 50.46 Annual Notification and Reporting for 2003," October 2004.

3 . INT-05-15, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.

4 . INT-08-14, Revision 1, "Evaluation of Containment Sump Strainer Modifications, Reduced Containment Initial Temperature and Accumulator Water Temperature on Indian Point Unit 3 (INT) Best Estimate Large Break LOCA Analysis," February 2010.

5 . INT-10-15, "10 CFR 50.46 Report for Thimble Plug Removal," December 2010.

Attachment to L TR-LIS-17-36 February 8, 2017 Page 10of14 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Reflood 1 6 . NF-ECH-12-24, "Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Bumdown Including Design Input Changes for Indian Point Unit 3," May 2012 .

7 . LTR-LIS-13-350, "Indian Point Units 2 and 3 I 0 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions,"

July 2013 .

8 . LTR-LIS-14-34, "Indian Point Units 2 and 3 IO CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction,"

January 2014.

Notes:

(a) The magnitude of this assessment was changed from prior versions of the PCT rackup sheet due to a recalculation in Reference 6.

(b) These assessments are coupled via an evaluation ofbumup effects which include thermal conductivity degradation, peaking factor bumdown and design input changes.

(c) Design input changes were a reduction in FQ(tr) from 2.5 to 2.3, FQ(ss) from 2.0 to 1.8, Fti.H from 1.7 to 1.65, and a corresponding reduction in Pbar-HA. These peaking factor limits supersede the values cited for the analysis-of-record.

Attachment to LTR-LIS-17-36 February 8, 2017 Page 11of14 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Reflood 2 Analysis Information EM: CQD (1996) Analysis Date: 1/23/2004 Limiting Break Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: OFA w/IFMs SGTP (%): 10 Notes: Analysis also supports 15x15 Upgraded Fuel (Reference 1). See note (c) for current peaking factor Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1944 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

. Revised Blowdown Heatup Uncertainty Distribution 5 3 2 . HOTSPOT Fuel Relocation Error 0 6 (a) 3 . Evaluation of Pellet Thermal Conductivity Degradation and Peaking 185 6 (b)

Factor Burndown 4 . Revised Heat Transfer Multiplier Distributions -35 7 5 . Error in Burst Strain Application 25 8 B. PLANNED PLANT MODIFICATION EVALUATIONS

. Bent Fuel Assembly Alignment Pins 5 2 2 . Evaluation oflncreased Containment Sump Strainer Metal and 80°F 12 4 Initial Containment/Accumulator Temperature 3 . Thimble Plug Removal 0 5 4 . Design Input Changes with Respect to Plant Operation -95 6 (b, c)

C. 2016 ECCS MODEL ASSESSMENTS 1 . None 0 D.OTHER*

1 . None 0 LICENSING BASIS PCT+ PCT ASSESSMENTS PCT= 2046 It is recommended that the licensee determine if these PCT allocations should be considered with respect to 10 CFR 50.46 reporting requirements.

References 1 . WCAP-16178-P, "Best-Estimate Analysis of the Large Break Loss of Coolant Accident for Indian Point Unit 3 Nuclear Plant Stretch Power Uprate," March 2004.

2 . IPP-04-126, "Revised PCT Rackups for 10 CFR 50.46 Annual Notification and Reporting for 2003," October 2004.

3 . INT-05-15, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.

4 . INT-08-14, Revision 1, "Evaluation of Containment Sump Strainer Modifications, Reduced Containment Initial Temperature and Accumulator Water Temperature on Indian Point Unit 3 (INT) Best Estimate Large Break LOCA Analysis," February 2010.

5 . INT-10-15, "10 CFR 50.46 Report for Thimble Plug Removal," December 2010.

Attachment to L TR-LIS-17-36 February 8, 2017 Page 12of14 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Reflood 2 6 . NF-ECH-12-24, "Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Bumdown Including Design Input Changes for Indian Point Unit 3," May 2012 .

7 . LTR-LIS-13-350, "Indian Point Units 2 and 3 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions,"

July 2013 .

8 . LTR-LIS-14-34, "Indian Point Units 2 and 3 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction,"

January 2014.

Notes:

(a) The magnitude of this assessment was changed from prior versions of the PCT rackup sheet due to a recalculation in Reference 6.

(b) These assessments are coupled via an evaluation of bumup effects which include thermal conductivity degradation, peaking factor bumdown and design input changes.

(c) Design input changes were a reduction in FQ(tr) from 2.5 to 2.3, FQ(ss) from 2.0 to 1.8, F8.H from 1.7 to 1.65, and a corresponding reduction in Pbar-HA. These peaking factor limits supersede the values cited for the analysis-of-record.

Attachment to LTR-LIS-17-36 February 8, 2017 Page 13of14 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/8/2017 Analysis Information EM: NOTRUMP Analysis Date: 3/3/2004 Limiting Break Size: 3 Inch FQ: 2.5 FdH: 1.7 Fuel: 15x15 Upgraded SGTP (%): 10 Notes: None Clad Temp {°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1543 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . Thimble Plug Removal 0 4 C. 2016 ECCS MODEL ASSESSMENTS I . None 0 D.OTHER*

I . AFW Flow Mismatch 0 2,3 LICENSING BASIS PCT+ PCT ASSESSMENTS PCT= 1543

  • It is recommended that the licensee determine if these PCT allocations should be considered with respect to 10 CFR 50.46 reporting requirements.

References I . WCAP-16212-NP (Class 3), "Indian Point Nuclear Generating Unit No. 3 Stretch Power Uprate NSSS and BOP Licensing Report," June 2004.

2 . LTR-LIS-09-340, "Indian Point Unit 3 Auxiliary Feedwater (AFW) Flow Mismatch SBLOCA Evaluation," May 2009.

3 . LTR-LIS-09-343, "10 CFR 50.46 Reporting for Indian Point Unit 3 Auxiliary Feedwater (AFW) Flow Mismatch SBLOCA Evaluation," May 2009.

4 . INT-10-15, "IO CFR 50.46 Report for Thimble Plug Removal," December 2010.

Notes:

None

/

Attachment to LTR-LIS-17-36 February 8, 2017 Page 14of14 10 CFR 50.46 Reporting SharePoint Site Check:

EMs applicable to Indian Point:

Best Estimate Large Break - CQD (1996)

Realistic Large Break - ASTRUM (2004)

Appendix K Small Break - NOTRUMP 2016 Issues Transmittal Letter Issue Descri tion None None

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