ML19360A018
| ML19360A018 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 02/13/2019 |
| From: | Entergy Nuclear Operations, Westinghouse |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19360A025 | List: |
| References | |
| L TR-LIS-19-14, NL-19-094 | |
| Download: ML19360A018 (15) | |
Text
Enclosure NL-19-094 Westinghouse Letter L TR-LIS-19-14 Indian Point Units 2 and 3 10 CFR 50.46 Annual Notification and Reporting for 2018 (14 pages)
Westinghouse Non-Proprietary Class 3 Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-5598 e-mail: mcmillh@westinghouse.com Ourref: LJR-LIS-19-14 February 13, 2019 Indian Point Units 2 and 3 10 CFR 50.46 Annual Notification and Reporting for 2018 I
D,;iar Sir or Madam:
This is a notification of 10 CFR 50.46 reporting information pertaining to the Westinghouse Electric Company Evaluation Models/analyses. As c~mmitted 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 2018 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) have been 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 (PCB), Carmen Teolis (412-374-2202), if there are any questions concerning this information.
Lead Author:
(Electronically Approved)*
Heather McMillen LOCA Integrated Services II Scope: All, except for PCT Summary sheets in the attachment Verified:
(Electronically Approved)*
Carmen D. Teolis LOCA Integrated Services I Co-Author:
(Electronically Approved)*
Sean T. Kinnas Containment and Radiological Analysis Scope: PCT Summary sheets in the attachment Approved:
(Electronically Approved)*
Amy J. Colussy Manager, LOCA Integrated Services I
Attachment:
10 CFR 50.46 Reporting Text and PCT Summary Sheets (13 Pages)
- Electronically approved records are authenticated in the electronic document management system.
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 U02 FUEL PELLET HEAT CAPACITY
Background
Page 1 of 13 A typographical error was discovered in the implementation of the U02 fuel pellet heat capacity as described by Equation C-4 of WCAP-8301 [1] for fuel rod heat-up calculations within the Appendix K Large Break and Small Break LOCA evaluation models. The erroneous formulation results in an over-prediction of heat capacity that increases with fuel temperature. The corrected formulation results in a maximum decrease in heat capacity on the order of approximately 1.2% for existing analyses of record.
This represents a Non-Discretionary Change in accordance with Section 4.1.2 ofWCAP-13451.
Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect I
The small over-prediction in U02 fuel pellet heat capacity has been evaluated to have a negligible effect on existing large and small break LOCA analysis results due to the small magnitude of the change, leading to an estimated PCT impact of 0°F.
Reference
Attachment to LTR-LIS-19-14 VAPOR TEMPERATURE RESETTING
Background
Page 2 of 13 In the WCOBRA/TRAC and WCOBRA/TRAC-TF2 codes, when the vapor temperature is greater than the wall temperature, and several other conditions are met, the vapor temperature is reset to the saturation temperature for heat transfer calculations. It was discovered that this vapor temperature resetting logic results in an inconsistency between the conduction solution and the hydraulic solution, such that energy is not conserved between the two solutions. The correction of this error represents a Non-Discretionary Change in the Evaluation Model as described in Section 4.1.2 of WCAP-13451.
Mfected Evaluation Model(s).
1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect I
Engineering judgement supported by :sensitivity calculations showed that correcting this error; had minimal impact on LOCA transient calculations, leading to an estimated peak cladding temperature impact of 0°F.
Attachment to LTR-LIS-19-14 LOCA Peak Cladding Temperature (PCT) Summary Plant Name:
INDIAN POINT 2 Utility Name:
Entergy Nuclear Northeast EM:
NOTRUMP AOR
Description:
Appendix K Small Break Summary Sheet Status: Current PCT (0F) 1028 Reference#
ANALYSIS-OF-RECORD 1
AOR + ASSESSMENTS PCT=
1028.0 °F REFERENCES Page 3 of 13 Note#
1 WCAP-16157-P, "Indian Point Nuclear Generating Unit No. 2 Stretch Power Uprate NSSS and BOP Licensing Report," January 2004.
NOTES:
(a) None Version: INDIAN POINT 2 IPP_LOCA-50.46_IPP_Base_Appendix._K_SBLOCA-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 LOCA Peak Cladding Temperature (PCT) Summary Plant Name:
INDIAN POINT 2 Utility Name:
Entergy Nuclear Northeast EM:
ASTRUM (2004)
Description:
Best Estimate Large Break Summary Sheet Status: Current PCT (0F) 1962 Reference#
ANALYSIS-OF-RECORD 1
Page 4 of 13 Note#
Delta PCT Reporting ASSESSMENTS*
(OAF)
Reference#
Note#
Year**
- 1.
HOTSPOT Fuel Relocation 0
3 Error
- 2.
Evaluation of Pellet Thermal Conductivity Degradation and 209 4,5 (a)
Peaking Factor Burndown
- 3.
Revised Heat Transfer 6
Multiplier Distributions
-32
- 4.
Error in Burst Strain Application 38 7
- 5.
Bent Fuel Assembly Alignment Pins 5
1
- 6.
Changes to Containment Sump Strainer Evaluation 0
2
- 7.
Evaluation of Design Input Changes with Respect to Plant
-63 4,5 (a,b,c)
Operation 8.. Evaluation oflncreased Containment Metal due to Control Rod Drive Shafts and 0
8 (d)
Water Shields AOR + ASSESSMENTS PCT= 2119.0 °F The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.
The "Reporting Year" refers to the annual reporting year in which this assessment was included.
Version: 1NDIANPOINT2 IPP_LOCA-50.46_IPP_Base_ASTRUM-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved 2007 2012 2013 2013 2005 2006 2012 2016
Attachment to LTR-LIS-19-14 Page 5 of 13 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(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Nori-Proprietary)," March 2012.
5 NF-ECH-12-23, "Information Regarding the Evaluation ofFuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown Includiµg Design Input Changes for Indian Point Unit 2," May 2012.
I 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 (IPP) Best-Estimate Large Break LOCA (BE LBLOCA) Analysis," March 2016.
NOTES:
(a) These assessments are coupled via an evaluation ofburnup 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%.
( c) The peaking factor limits and steam generator tube plugging limit in note (b) supersede the values cited for the analysis-of-record.
( d) The evaluation of the added control rod drive shafts and water shields takes credit for margin in several post-analysis-of-record evaluations.
Version: INDIAN POINT 2 IPP _LOCA-50.46_IPP _Base_ASTRUM-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 Page 6 of 13 LOCA Peak Cladding Temperature (PCT) Summary Plant Name:
INDIAN POINT 3 Utility Name:
Entergy Nuclear Northeast EM:
NO TRUMP AOR
Description:
Appendix K Small Break Summary Sheet Status: Current Reference#
Note#
ANALYSIS-OF-RECORD PCT (
0F) 1543 1
Delta PCT Reporting ASSESSMENTS*
(OAF)
Reference#
Note#
Year**
- 1.
Thimble Plug Removal 0
4 2010
- 2.
AFW Flow Mismatch 0
2,3 2009 AOR + ASSESSMENTS PCT=
1543.0 °F The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.
- The "Reporting Year" refers to the annual reporting year in which this assessment was included.
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, "10 CFR 50.46 Report for Thimble Plug Removal," December 2010.
NOTES:
(a) None Version: INDIAN POINT 3 INT_LOCA-50.46_1NT__Base_Appendix_K_SBLOCA-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
-\\
L Attachment to LTR-LIS-19-14 LOCA Peak Cladding Temperature (PCT) Summary Plant Name:
INDIAN POINT 3 Utility Name:
Entergy Nuclear Northeast EM:
CQD (1996)
Description:
Best Estimate Large Break - Composite Summary Sheet Status: Current ANALYSIS-OF-RECORD ASSESSMENTS*
- 1.
Revised Blowdown Heatup Uncertainty Distribution
- 2.
HOTSPOT Fuel Relocation Error
- 3.
Evaluation of Pellet Thermal Conductivity Degradation and Peaking Factor Burndown
- 4.
Revised Heat Transfer Multiplier Distributions
- 5.
Error in Burst Strain Application
- 6.
Bent Fuel Assembly Alignment Pins
- 7.
Evaluation of Increased Containment Sump Strainer Metal and 80°F Initial Containment/ Accumulator Temperature
- 8.
Thimble Plug Removal
- 9.
Design Input Changes with Respect to Plant Operation AOR + ASSESSMENTS PCT {°F) 1944 Delta PCT (OAF) 5 0
185
-35 25 5
12 0
-95 PCT=
Reference#
1 Reference#
3 6
6 7
8 2
4 5
6 2046.0 °F Page 7 of 13 No.te#
Reporting Note#
Year**
2004 (a) 2012 2012 (b) 2013 2013 2004 2009 2010 (b,c) 2012 The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.
- The "Reporting Year" refers to the annual reporting year in which this assessment was included.
Version: INDIAN POINT 3 INT_LOCA-50.46_INT_Base_CQD_Composite-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 Page 8 of 13 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.
6 NF-EdH-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 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 ofburnup 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, FAfI 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.
(
Version: INDIAN POINT 3 INT_LOCA-50.46_INT_Base_CQD_Composite -1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 LOCA Peak Cladding Temperature (PCT) Summary Plant Name:
INDIAN POINT 3 Utility Name:
Entergy Nuclear Northeast EM:
CQD (1996}
Description:
Best Estimate Large Break - Reflood 1 Summary Sheet Status: Current ANALYSIS-OF-RECORD ASSESSMENTS*
- 1.
Revised Blowdown Heatup Uncertainty Distribution
- 2.
HOTSPOT Fuel Relocation Error
- 3.
Evaluation of Pellet Thermal Conductivity Degradation and Peaking Factor Burndown
- 4.
Revised Heat Transfer Multiplier Distributions
- 5.
Error in Burst Strain Application
- 6.
Bent Fuel Assembly Alignment Pins
- 7.
Evaluation oflncreased Containment Sump Strainer Metal and 80°F Initial Containment/ Accumulator Temperature
- 8.
Thimble Plug Removal
- 9.
Design Input Changes with Respect to Plant Operation AOR + ASSESSMENTS PCT (°F) 1904 Delta PCT
(
0AF) 5 0
205 5
20 5
3 0
-130 Re_ference #
1 Reference#
3 6
6 7
8 2
4 5
6 PCT= 2017.0 °F Note#
Note#
(a)
(b)
(b,c)
Page 9 of 13 Reporting Year**
2004 2012 2012 2013 2013 2004 2009 2010 2012 The lice~see should determine the reportability of these assessments pursuant to 10 CFR 50.46.
- The "Reporting Year" refers to the annual reporting year in which this assessment was included.
Version: INDIAN POINT 3 INT_LOCA-50.46_1NT_Base_CQD_Refloodl-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 Page 10 of 13 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.
6 NF-ECH-12-24, "Information Regarding the Evaluation of Fuel Pellet Thermal Conductivity Degradation
~d 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 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 ofburnup 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 L8, FdH 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.
Version: INDIAN POINT 3 INT_LOCA-50.46_INT_Base_CQD_Retloodl-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19:-14 LOCA Peak Cladding Temperature (PCT) Summary Plant Name:
INDIAN POINT 3 Utility Name:
Entergy Nuclear Northeast EM:
CQD (1996)
Description:
Best Estimate Large Break - Reflood 2 Summary Sheet Status: Current ANALYSIS-OF-RECORD ASSESSMENTS*
- 1.
Revised Blowdown Heatup Uncertainty Distribution
- 2.
HOTSPOT Fuel Relocation Error
- 3.
Evaluation of Pellet Thermal Conductivity Degradation and Peaking Factor Burndown
- 4.
Revised Heat Transfer Multiplier Distributions
- 5.
Error in Burst Strain Application
- 6.
Bent Fuel Assembly A11gnment Pins
- 7.
Evaluation of Increased Containment Sump Strainer Metal and 80°F Initial Containment/ Accumulator Temperature
- 8.
Thimble Plug Removal
- 9.
Design Input Changes with Respect to Plant Operation AOR + ASSESSMENTS PCT (°F) 1944 Delta PCT
(°AF) 5 0
185
-35 25 5
12 0
-95 PCT=
Reference#
1 Reference#
3 6
6 7
8 2
4 5
6 2046.0 °F Page 11 of 13 Note#
Reporting Note#
Year**
2004 (a) 2012 2012 (b) 2013 2013 2004 2009 2010 (b,c) 2012 The licensee should determine the reportability of these assessments pursuant to 10 CFR 50.46.
- The "Reporting Year" refers to the annual reporting year in which this assessment was included.
Version: INDIAN POINT 3 INT_LOCA-50.46_1NT_Base_CQD_Reflood2-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 Page 12 of 13 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.
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 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 ofburnup 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, FAfI from 1.7 to 1.65, and a corresponding r:eduction in Pbar-HA. These peaking factor limits supersede the values cited for the analysis-of-record.
Version: INDIAN POINT 3 INT_LOCA-50.46_INT_Base_CQD_Reflood2-1.1 V.V
© 2019 Westinghouse Electric Company LLC All Rights Reserved
Attachment to LTR-LIS-19-14 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 2018 Issues Transmittal Letter Issue Descri tion None None
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Page 13 of 13