NL-18-082, Submittal of 2017 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report

From kanterella
Jump to navigation Jump to search
Submittal of 2017 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report
ML18323A081
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 11/13/2018
From: Vitale A
Entergy Nuclear Northeast
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-18-082
Download: ML18323A081 (21)
v • d • e


Text

  • Entergx
  • ='=-

Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, NY 10511-0249 Tel (914) 254-6700 Anthony J Vitale Site Vice President 10 CFR 50.46 NL-18-082 November 13, 2018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

2017 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 N RC, "2016 Annual 10 CFR 50.46 Emergency Core Cooling System Evaluation Changes Report," dated September 11, 2017 (NL-17-116)

Dear Sir or Madam:

This letter provides the annual 1 O CFR 50.46 report for calendar year 2017 for Indian Point Units 2 (IP2) and 3 (IP3). The prior report for calendar year 2016 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 2017. 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.

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

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

1) General Code Maintenance (Indian Point Units 2 and 3)
2) Vessel Average Temperature Uncertainty (Indian Point Units 2 and 3)
3) Error in the Upper Plenum Fluid Volume Calculation (Indian Point Units 2 and 3)
4) Inconsistent Application of Numerical Ramp Applied to the Entrained Liquid/ Vapor lnterfacial Drag Coefficient (Indian Point Units 2 and 3)
5) Inappropriate Resetting of Transverse Liquid Mass Flow (Indian Point Units 2 and 3)
6) Steady-State Fuel Temperature Calibration Method (Indian Point Unit 2)

These changes are further described in Enclosure 1.

NL-18-082 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)(1) for calendar year 2017.

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/mm/aye

Enclosure:

1. Westinghouse Letter LTR-LIS-18-19, "10 CFR 50.46 Annual Notification and Reporting for 2017", dated February 15, 2018.

cc: Mr. David Lew, 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-18-082 Westinghouse Letter LTR-LIS-18-19 10 CFR 50.46 Annual Notification and Reporting for 2017 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 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-5598 e-mail: mcmillh@westinghouse.com Ourref: L1R-LIS-18-19 February 15, 2018 Indian Point Units 2 and 3 10 CFR 50.46 Annual Notification and Reporting for 2017

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 2017 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) summary sheets are enclosed. All necessary revisions for any non-zero, non-discretionary PCT changes have been included. Changes with estimated PCT impacts of 0°F may not be presented on the PCT summary 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 requirements 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. (17 Pages)

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

© 2018 Westinghouse Electric Company LLC All Rights Reserved

Attachment to LTR-LIS-18-19 February 15, 2018 Page 1 of 17 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 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated Peak Cladding Temperature (PCT) impact of 0°F.

~~ --

Attachment to LTR-LIS-18-19 February 15, 2018 Page 2 of 17 VESSEL AVERAGE TEMPERATURE UNCERTAINTY

Background

A hysteresis issue was identified for plants with Weed Resistance Temperature Detectors (RTDs) supplied to Westinghouse, which resulted in an additional uncertainty of +0.1 °F bias (indicated higher than actual) that applies to the Reactor Coolant System (RCS) average temperature accident analysis initial condition uncertainty. This discrepancy has been evaluated for impact on existing Large and Small Break Loss-of-Coolant Accident (LOCA) analysis results, and its resolution represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP 1996 Westinghouse Best-Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect This issue was evaluated as having a negligible impact on existing Large and Small Break LOCA analysis results, leading to an estimated PCT impact of 0°F.

Attachment to LTR-LIS-18-19 February 15, 2018 Page 3 of 17 ERROR IN THE UPPER PLENUM FLUID VOLUME CALCULATION

Background

An error was found in the fluid volume calculation in the upper plenum where the support column outer diameter was being used instead of the inner diameter. The correction of this error lead to a reduction in the upper plenum fluid volume used in the Appendix K Large Break LOCA and Small Break LOCA analyses. The corrected values represent a less than 1 % change in the total RCS fluid volume and will be incorporated on a forward-fit basis, based on the evaluated impact on the current licensing basis analysis results. These changes represent a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The differences in the upper plenum fluid volume are relatively minor and have been evaluated to have a negligible effect on large and small break LOCA analysis results, leading to an estimated PCT impact of 0°F.

Attachment to LTR-LIS-18-19 February 15, 2018 Page 4 of 17 INCONSISTENT APPLICATION OF NUMERICAL RAMP APPLIED TO THE ENTRAINED LIQUID/VAPOR INTERFACIAL DRAG COEFFICIENT

Background

A numerical ramp which was used to account for the disappearance of the entrained liquid phase was applied to the entrained liquid / vapor interfacial drag coefficient. The numerical ramp was applied such that the interfacial drag coefficient used in the solution of the entrained liquid and vapor momentum equations was not consistent. WCOBRA/TRAC was updated to apply the numerical ramp prior to usage of the interfacial drag coefficient in the momentum equations, such that a consistent interfacial drag coefficient was used in the entrained liquid and vapor momentum equations.

This item 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 Evaluation Model Using ASTRUM Estimated Effect Based on the code validation results, the impact of correcting the error is estimated to have a 0°F impact on PCT.

Attachment to LTR-LIS-18-19 February 15, 2018 Page 5 of 17 INAPPROPRIATE RESETTING OF TRANSVERSE LIQUID MASS FLOW

Background

In the WCOBRA/TRAC routine which evaluates the mass and energy residual error of the time step solution, the transverse liquid mass flow is reset as the liquid phase disappears. The routine is updated to remove the resetting of the transverse liquid mass flow since the routine is to only evaluate the residual error based on the time step solution values.

This item 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 Evaluation Model Using ASTRUM Estimated Effect Based on the code validation results and limited applicability of the logic removed, correcting the error is estimated to have a 0°F impact on PCT.

Attachment to LTR-LIS-18-19 February 15, 2018 Page 6 of 17 STEADY-STATE FUEL TEMPERATURE CALIBRATION METHOD

Background

In the Automated Statistical Treatment of Uncertainty Method (ASTRUM) Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) Evaluation Model (EM), the steady-state fuel pellet temperature calibration method involves solving for the hot gap width (AGFACT) to calibrate the fuel temperature for each fuel rod. In some infrequent situations, small non-conservatisms can occur in the calibration process such that the resulting fuel pellet temperature will be slightly lower than intended and outside the acceptable range defined by Table 12-6 of WCAP-16009-P/NP-A [l].

This issue has been evaluated to estimate the impact on ASTRUM BE LBLOCA 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) 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A review of licensing basis analyses concluded that the potential non-conservatisms in the fuel pellet temperature calibration did not occur for the limiting analysis cases. Therefore, an estimated PCT impact of 0°F is assigned for 10 CFR 50.46 reporting purposes.

Reference(s)

1) WCAP-16009-P/NP-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," January 2005.

Attachment to LTR-LIS-18-19 February 15, 2018 Page 7 of 17 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: 2/1/2018 Analysis Information EM: ASTRUM (2004) Analysis Date: 2/15/2005 Limiting Break Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: 15x15 Upgraded SGTP (%): 10 Notes: See note (b) for current peaking factor and SGTP limitations.

Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1962 1 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 Bumdown 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 of Increased Containment Metal due to Control Rod Drive 0 8 (c)

Shafts and Water Shields C. 2017 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, "10 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(1) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary),"

March 2012.

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

Attachment to LTR-LIS-18-19 February 15, 2018 Page 8 of 17 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: 2/1/2018 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 CPR 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 (IPP) 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, FAfI 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-18-19 February 15, 2018 Page 9 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Indian Point Unit 2 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 Analysis Information EM: NOTRUMP Analysis Date: 1/11/2004 Limiting Break Size: 3 inch FQ: 2.5 FdH: 1.7 Fuel: I5x15 Upgraded SGTP (%): 10 Notes: None Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1028 1 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS 1 . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS 1 . None 0 C. 2017 ECCS MODEL ASSESSMENTS 1 . None 0 D.OTHER*

1 . 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 10 CFR 50.46 reporting requirements.

References 1 . 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-18-19 February 15, 2018 Page 10 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 Composite Analysis Information EM: CQD (1996) Analysis Date: 1/23/2004 Limiting Break Size: Guillotine FQ: 2.5 FdH: 1.7 Fuel: OFA w/lFMs SGTP (%): 10 Notes: Analysis also supports 15xl5 Upgraded Fuel (Reference I). See note (c) for current peaking factor Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1944 1 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. 2017 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 IO CPR 50.46 reporting requirements.

References I . 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 CPR 50.46 Annual Notification and Reporting for 2003," October 2004.

3 . INT-05-15, "10 CPR 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 CPR 50.46 Report for Thimble Plug Removal," December 2010.

Attachment to LTR-LIS-18-19 February 15, 2018 Page 11 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 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 IO CPR 50.46 Report for Revised Heat Transfer Multiplier Distributions," July 2013 .

8 . LTR-LIS-14-34, "Indian Point Units 2 and 3 IO CPR 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, FL'iH 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-18-19 February 15, 2018 Page 12 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 Reflood 1 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 1904 1 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 of Increased 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. 2017 ECCS MODEL ASSESSMENTS I . None 0 D. OTHER*

I . 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 CPR 50.46 reporting requirements.

References I . 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 CPR 50.46 Annual Notification and Reporting for 2003," October 2004.

3 . INT-05-15, "10 CPR 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, "IO CPR 50.46 Report for Thimble Plug Removal," December 2010.

Attachment to LTR-LIS-18-19 February 15, 2018 Page 13 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 Reflood 1 6 . NF-ECH-12-24, "Information Regarding the Evaluation of Pue! Pellet Thermal Conductivity Degradation and Peaking Factor Burndown Including Design Input Changes for Indian Point Unit 3," May 2012 .

7 . LTR-LIS-13-350, "Indian Point Units 2 and 3 10 CPR 50.46 Report for Revised Heat Transfer Multiplier Distributions," July 2013.

8 . LTR-LIS-14-34, "Indian Point Units 2 and 3 10 CPR 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 burn up effects which include thermal conductivity degradation, peaking factor burndown 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, FllH 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-18-19 February 15, 2018 Page 14 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 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 15xl5 Upgraded Fuel (Reference 1). See note (c) for current peaking factor Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1944 1 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. 2017 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-18-19 February 15, 2018 Page 15 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 Reflood 2 6 . NF-ECH-12-24, "Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown 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 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 of burn up effects which include thermal conductivity degradation, peaking factor burndown 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, FAH 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-18-19 February 15, 2018 Page 16 of 17 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Indian Point Unit 3 Utility Name: Entergy Nuclear Northeast Revision Date: 2/1/2018 Analysis Information EM: NOTRUMP Analysis Date: 3/3/2004 Limiting Break Size: 3 Inch FQ: 2.5 FdH: 1.7 Fuel: 15xl5 Upgraded SGTP (%): 10 Notes: None Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1543 1 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . Thimble Plug Removal 0 4 C. 2017 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 . L TR-LIS-09-343, "IO CFR 50.46 Reporting for Indian Point Unit 3 Auxiliary Feed water (AFW) Flow Mismatch SBLOCA Evaluation," May 2009.

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

Notes:

None

Attachment to LTR-LIS-18-19 February 15, 2018 Page 17 of 17 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 2017 Issues Transmittal Letter Issue Descri tion None None

Retrieved from "https://kanterella.com/w/index.php?title=NL-18-082,_Submittal_of_2017_Annual_10_CFR_50.46_Emergency_Core_Cooling_System_Evaluation_Changes_Report&oldid=2398190"