Text

CFR 50.46 30-Day Report of Changes & Errors to Calculated Peak Cladding Temperature
	ML14064A431
Person / Time
Site:	Watts Bar
Issue date:	02/28/2014
From:	James Shea; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CNL-14-035
	Download: ML14064A431 (26)
	v • d • e

L44 140228 001 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-14-035 February 28, 2014 10 CFR 50.46 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License No. NPF-90 Construction Permit No. CPPR-92 NRC Docket Nos. 50-390 and 50-391

Subject:

10 CFR 50.46 Day Report for Watts Bar, Units 1 and 2

Reference:

TVA Letter to NRC, 10 CFR 50.46 Day and Annual Report for 2012, dated April 25, 2013 [ML13120A005]

The purpose of this letter is to provide a 30-day report of changes and errors to the calculated peak cladding temperature (PCT) for the Watts Bar Nuclear Plant (WBN), Units 1 and 2, Emergency Core Cooling System (ECCS) evaluation model. This report is required in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, paragraph (a)(3)(ii).

The PCT changes and errors identified for WBN, Units 1 and 2, in the referenced report submitted on April 25, 2013, when expressed as the cumulative sums of the absolute magnitudes, exceed 50 degrees Fahrenheit (°F). In accordance with 10 CFR 50.46(a)(3)(ii), a holder of an operating license or construction permit is required to report changes and errors affecting an ECCS evaluation model to the NRC within 30 days when the cumulative sum of the absolute magnitudes of resulting PCT changes exceeds 50°F. The licensee is also required to include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with the 10 CFR 50.46 requirements.

On January 30, 2014, Westinghouse Electric Company (WEC) notified Tennessee Valley Authority (TVA) of changes and errors affecting the HOTSPOT computer code used in the WEC Best Estimate Large Break Loss Of Coolant Accident (BE LBLOCA) evaluation model and Realistic LBLOCA Evaluation Model using ASTRUM. The reported changes and errors are reportable to the NRC in a 30-day report for WBN, Units 1 and 2, in accordance with 10 CFR 50.46(a)(3)(ii).

U.S. Nuclear Regulatory Comm ission Page 2 February 28, 2014 As indicated in Enclosure 1, the current updated (net) licensing basis PCT for the WBN , Unit 1, LBLOCA analysis of record (AOR) is now estimated to be 1865°F, a change of -15°F from the previous (referenced) report. The updated (net) licensing basis PCT for the Small Break LOCA (SBLOCA) evaluation model is unchanged from the SBLOCA AOR baseline PCT, and remains at 1132°F.

As indicated in Enclosure 2, the current updated (net) licensing basis PCT for the WBN, Unit 2, LBLOCA AOR is now estimated to be 1711 °F, a change of -55°F from the previous (referenced) report. The updated (net) licensing basis PCT for the SBLOCA evaluation model is unchanged from the SBLOCA AOR baseline PCT, and remains at 1184°F.

The enclosed reports demonstrate that the calculated PCTs for the ECCS LBLOCA and SBLOCA AORs for WBN , Units 1 and 2, are below the limit of 2200°F. This provides the requisite demonstration of compliance with the 10 CFR 50.46 requirements . Therefore , no schedule for reanalysis. or other action to show compliance is required .

There are no regulatory commitments in this letter. Please direct questions concerning this report to Tom Hess at (423) 751 -3487.

Enclosures :

1. Watts Bar Nuclear Plant, Unit 1, 10 CFR 50.46 30-Day Report

2. Watts Bar Nuclear Plant, Unit 2, 10 CFR 50.46 30-Day Report

3. Additional Information on the Evaluation of Hotspot Burst Strain Error Corrections cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant Units 1 and 2

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 30-DAY REPORT In accordance with the reporting requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.46(a)(3)(ii), the following is a summary of the limiting design basis loss of coolant accident (LOCA) analysis results established using the current Watts Bar Nuclear Plant (WBN)

Emergency Core Cooling System (ECCS) evaluation models for Unit 1. This report describes the changes and errors affecting the calculated peak cladding temperatures (PCTs) since the last annual report.

The last 10 CFR 50.46 annual report for WBN Unit 1 was submitted to the United States Nuclear Regulatory Commission (NRC) on April 25, 2013 (Reference 1). As indicated in the previous report, Westinghouse Electric Company (WEC) WCAP-14839, Revision 1, Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant (Reference 2), is the current Best Estimate Large Break LOCA (BE LBLOCA) analysis of record (AOR) for Unit 1, with a baseline PCT value of 1892 °F. The baseline PCT for the previous report corresponds to the Reflood 2 (late reflood) case. For this 30-day report, both Reflood 1 (early reflood) and Reflood 2 results are reported. The Reflood 1 case has a baseline PCT value of 1656 °F. The current Small Break LOCA (SBLOCA) AOR for Unit 1 is WTV-RSG-06-015, LOCA & Non-LOCA Analysis Summary for Replacement Steam Generator for WBN Unit 1 (Reference 3), with a baseline PCT of 1132°F.

Tables 1 and 2 detail the accumulated PCT effects resulting from the changes and errors in the LBLOCA and SBLOCA analyses since each of the respective AORs (References 2 and 3) was established for WBN Unit 1. Changes and errors that were not previously identified in either the most recent 10 CFR 50.46 annual report or 30-day report are detailed in the Notes section following the tables.

As indicated in Table 1 for WBN Unit 1, the current updated (net) licensing basis PCT for the LBLOCA analysis is 1865°F. This is a 15°F decrease in PCT from the last annual report (Reference 1). Note that in the last 30 day special report (Reference 15), the early reflood (Reflood 1) case was the most limiting for PCT; however, with the most recently reported errors from WEC, the late reflood (Reflood

2) analysis case now results in a more limiting PCT. Both the early and late reflood cases are reported in Table 1. The WBN Unit 1 updated (net) licensing basis PCT value of 1132°F for the SBLOCA is unchanged from the SBLOCA AOR baseline PCT, as reflected in the previous reports.

In accordance with 10 CFR 50.46(a)(3)(ii), future changes affecting WBN Unit 1 will be considered significant for reporting purposes because the absolute magnitude of the accumulated changes and errors affecting the calculated PCT since the last LBLOCA re-analysis was performed (Reference 2) exceeds 50°F.

E1-1

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 30-DAY REPORT TABLE 1 (Sheet 1 of 3)

Summary of Changes to WBN Unit 1 PCT for LBLOCA Reflood 1 Reflood 2 PCT (°F) PCT PCT (°F) PCT Year Description (°F) (°F) Notes References 1998 BE LBLOCA AOR Baseline PCT 1656 --- 1892 --- --- 2 1999 Vessel Channel DX Error 56 56 -4 4 --- 7 2000 Increased Accumulator Room ---

4 4 4 4 7 Temperature Evaluation 2000 1.4% Uprate Evaluation 12 12 12 12 --- 7 2000 Accumulator Line/Pressurizer ---

-37 37 -131 131 7 Surge Line Data Evaluation 2000 MONTECF Decay Heat ---

4 4 4 4 8 Uncertainty Error 2001 WBN Specific LBLOCA Vessel ---

0 0 0 0 9 Geometry Input Errors 2003 Input Error Resulting in Incomplete ---

60 60 0 0 10 Solution Matrix 2003 Tavg Bias Error 8 8 8 8 --- 10 2004 Increased Stroke Time for ECCS ---

0 0 0 0 11 Valves 2004 Revised Blowdown Heatup ---

5 5 5 5 11 Uncertainty Distribution 2006 Replacement Steam Generators ---

-50 50 -10 10 12 (D3 to 68AXP) 2006 HOTSPOTTM Fuel Relocation Error 0 0 65 65 --- 12 E1-2

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 30-DAY REPORT TABLE 1 (Sheet 2 of 3)

Summary of Changes to WBN Unit 1 PCT for LBLOCA Reflood 1 Reflood 2 PCT PCT PCT PCT Year Description (°F) (°F) (°F) (°F) Notes References 2012 PMID/PBOT Violation Evaluation 20 20 20 20 --- 5, 6 2012 TCD and Peaking Factor 114 114 15 15 --- 5, 6 Burndown 2013 WCOBRA/TRACTM History File 0 0 0 0 --- 4 Dimension Error 2013 General Code Maintenance 0 0 0 0 --- 1 TM 2013 HOTSPOT Burst Temperature 0 0 0 0 --- 1 Calculation for ZIRLOTM Cladding 2013 HOTSPOTTM Iteration Algorithm for Calculation Initial Fuel Pellet 0 0 0 0 --- 1 Average Temperature 2013 WCOBRA/TRACTM Automated 0 0 0 0 --- 1 Restart Process Logic Error 2013 Rod Internal Pressure Calculation 0 0 0 0 --- 1 Error 2013 Elevations for Heat Slab 0 0 0 0 --- 15 Temperature Initialization 2013 Heat Transfer Model Error 0 0 0 0 --- 15 Corrections 2013 Correction to Heat Transfer Node 0 0 0 0 --- 15 Initialization 2013 Mass Conservation Error Fix 0 0 0 0 --- 15 2013 Correction to Split Channel 0 0 0 0 --- 15 Momentum Equation 2013 Heat Transfer Logic Correction 0 0 0 0 --- 15 for Rod Burst Calculation 2013 Changes to Vessel Superheated 0 0 0 0 --- 15 Steam Properties 2013 Update to Metal Density 0 0 0 0 --- 15 Reference Temperatures 2013 Decay Heat Model Error 0 0 0 0 --- 15 Corrections E1-3

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 30-DAY REPORT TABLE 1 (Sheet 3 of 3)

Summary of Changes to WBN Unit 1 PCT for LBLOCA Reflood 1 Reflood 2 PCT PCT PCT PCT Year Description (°F) (°F) (°F) (°F) Notes References 2013 Correction to the Pipe Exit 0 0 0 0 15 Pressure Drop Error 2013 WCOBRA/TRAC File Dimension 0 0 0 0 15 Error Correction 2013 Revised Heat Transfer Multiplier

-40 40 -85 85 15 Distributions 2014 HOTSPOT Burst Strain Error 20 20 70 70 1 ---

Updated (net) licensing basis PCT 1832 --- 1865 --- --- ---

AOR PCT + PCT Cumulative sum of PCT changes --- 430 --- 433 --- ---

PCT E1-4

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 30-DAY REPORT NOTE:

1) HOTSPOT Burst Strain Error An error in the application of the burst strain was discovered in HOTSPOT. The equation for the application of the burst strain is given as Equation 7-69 in WCAP-12945-P-A (Reference 14).

The outer radius of the cladding after burst occurs should be calculated based on the burst strain, and the inner radius of the cladding should be calculated based on the outer radius. In HOTSPOT, the burst strain is applied to the calculation of the cladding inner radius. The cladding outer radius is then calculated based on the inner radius. As such, the burst strain is incorrectly applied to the inner radius rather than the outer radius, which impacts the resulting cladding geometry at the burst elevation after burst occurs. Correction of the erroneous calculation results in thinner cladding at the burst node and more fuel relocating into the burst node, leading to an increase in the Peak Cladding Temperature (PCT) at the burst node. This issue has been evaluated to estimate the impact on existing 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 of WCAP-13451(Reference 13).

A case representative of Watts Bar Unit 1 was run using HOTSPOT versions which only differ in the burst strain application. Based on the change in the 95th percentile results, estimated PCT effects of 20°F for Reflood 1, and 70°F for Reflood 2 have been established for 10 CFR 50.46 reporting purposes for Watts Bar Unit 1.

E1-5

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 30-DAY REPORT TABLE 2 Summary of Changes to WBN Unit 1 PCT for SBLOCA Year Description SBLOCA SBLOCA Notes References PCT (°F) PCT

(°F) 2006 SBLOCA AOR Baseline PCT 1132 --- --- 3 2013 NOTRUMP-EMTM Evaluation of 0 0 --- 1 Fuel Pellet TCD Updated (net) licensing basis PCT 1132 --- --- ---

AOR PCT + PCT Cumulative sum of PCT changes --- 0 --- ---

PCT E1-6

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 30-DAY REPORT REFERENCES

1. Letter from TVA to NRC, 10 CFR 50.46 Day and Annual Report for 2012, dated April 25, 2013 [ML13120A005]

2. WCAP-14839, Revision 1, Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant, September 1998

3. WTV-RSG-06-015, LOCA & Non-LOCA Analysis Summary for Replacement Steam Generator, February 2006

4. Letter from TVA to NRC, 10 CFR 50.46 Day Report for Watts Bar Unit 1, dated March 19, 2013 [ML13046A002]

5. Letter from TVA to NRC, Supplement to 10 CFR 50.46 Day Special Report, dated February 13, 2013 [ML13046A002]

6. Letter from TVA to NRC, 10 CFR 50.46 Day Special Report, dated October 18, 2012

[ML12296A254]

7. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes Day Report and Annual Notification and Reporting for 2000, dated October 26, 2000 [ML003764646]

8. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes Day Report and Revised Annual Notification Report for 2000, dated September 7, 2001 [ML012570290]

9. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - Annual Notification and Reporting for 2001, dated April 3, 2002 [ML021070404]

10. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes Day Report and Revised Annual Notification and Reporting for 2003, dated April 19, 2004 [ML041130196]

11. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - Annual Notification and Reporting for 2004, dated April 19, 2005 [ML051120164]

12. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes Day Report and Annual Notification and Reporting for 2006, dated July 3, 2007 [ML071860388]

13. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting,"

October 1992

14. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, Code Qualification Document for Best Estimate LOCA Analysis, 1998

15. Letter from TVA to NRC, 10CFR50.46 - 30 day Report for Watts Bar, Unit 1, dated August 28, 2013 [ML13267A034]

E1-7

ENCLOSURE 2 WATTS BAR NUCLEAR PLANT, UNIT 2 10 CFR 50.46 30-DAY REPORT In accordance with the reporting requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.46(a)(3)(ii), the following is a summary of the limiting design basis loss of coolant accident (LOCA) analysis results established using the current Watts Bar Nuclear Plant (WBN)

Emergency Core Cooling System (ECCS) evaluation models for Unit 2. This report describes the changes and errors affecting the calculated peak cladding temperatures (PCTs) since the last annual report.

The last 10 CFR 50.46 annual report for WBN Unit 2 was submitted to the United States Nuclear Regulatory Commission (NRC) on April 25, 2013 (Reference 1). That report indicated that, Westinghouse Electric Company (WEC) WCAP-17093-P, Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Unit 2 Nuclear Plant Using the ASTRUM Methodology (Reference 2), as the current Best Estimate Large Break LOCA (BE LBLOCA) analysis of record (AOR) for Unit 2. However, since that report, reference 6 established WCAP-17093-P, Revision 1 (reference 3) to be the current AOR for unit 2, with a baseline PCT value of 1766 °F. The current Small Break LOCA (SBLOCA) AOR for Unit 2 is WBT-D-1460, Final Small Break LOCA Summary Report for WBN Unit 2 (Reference 4), with a baseline PCT of 1184°F.

Tables 1 and 2 detail the accumulated PCT effects resulting from the changes and errors in the LBLOCA and SBLOCA analyses since each of the respective AORs (References 3 and 4) was established for WBN Unit 2. Changes and errors that were not previously identified in either the most recent 10 CFR 50.46 annual report or 30-day report are detailed in the Notes section following the tables.

As indicated in Table 1 for WBN Unit 2, the current updated (net) licensing basis PCT for the LBLOCA analysis is 1711°F. This is a 55°F decrease in PCT from the last annual report (Reference 1). The WBN Unit 2 updated (net) licensing basis PCT value of 1184°F for the SBLOCA is unchanged from the SBLOCA AOR baseline PCT, as reflected in the previous reports.

In accordance with 10 CFR 50.46(a)(3)(ii), future changes affecting WBN Unit 2 will be considered significant for reporting purposes because the absolute magnitude of the accumulated changes and errors affecting the calculated PCT since the last LBLOCA re-analysis was performed (Reference 3) exceeds 50°F.

E2-1

ENCLOSURE 2 WATTS BAR NUCLEAR PLANT, UNIT 2 10 CFR 50.46 30-DAY REPORT TABLE 1 (Sheet 1 of 2)

Summary of Changes to WBN Unit 2 PCT for LBLOCA and SBLOCA LBLOCA SBLOCA PCT PCT PCT PCT Year Description (°F) (°F) (°F) (°F) Notes References 2013 BE LBLOCA AOR Baseline PCT 1766 --- --- --- --- 3 2010 SBLOCA AOR Baseline PCT -- - - -- - - 1184 -- - - -- - - 4 2013 Elevations for Heat Slab ---

0 0 0 0 5 Temperature Initialization 2013 Heat Transfer Model Error ---

0 0 0 0 5 Corrections 2013 Correction to Heat Transfer Node ---

0 0 0 0 5 Initialization 2013 Mass Conservation Error Fix 0 0 0 0 --- 5 2013 Correction to Split Channel ---

0 0 0 0 5 Momentum Equation 2013 Heat Transfer Logic Correction for ---

0 0 0 0 5 Rod Burst Calculation 2013 Changes to Vessel Superheated ---

0 0 0 0 5 Steam Properties 2013 Update to Metal Density ---

0 0 0 0 5 Reference Temperatures 2013 Decay Heat Model Error ---

0 0 0 0 5 Corrections 2013 Correction to the Pipe Exit ---

0 0 0 0 5 Pressure Drop Error 2013 WCOBRA/TRAC File Dimension ---

0 0 0 0 5 Error Correction 2013 Revised Heat Transfer Multiplier ---

-55 55 0 0 5 Distributions E2-2

ENCLOSURE 2 WATTS BAR NUCLEAR PLANT, UNIT 2 10 CFR 50.46 30-DAY REPORT TABLE 1 (Sheet 2 of 2)

Summary of Changes to WBN Unit 2 PCT for LBLOCA LBLOCA SBLOCA PCT PCT PCT PCT Year Description (°F) (°F) (°F) (°F) Notes References Initial Fuel Pellet Average 2013 Temperature Uncertainty 0 0 0 0 5 Calculation 2014 HOTSPOT Burst Strain Error 0 0 0 0 1 ---

Updated (net) licensing basis PCT 1711 --- 1184 --- --- ---

AOR PCT + PCT Cumulative sum of PCT changes --- 55 --- 0 --- ---

PCT E2-3

ENCLOSURE 2 WATTS BAR NUCLEAR PLANT, UNIT 2 10 CFR 50.46 30-DAY REPORT NOTES:

1) HOTSPOT Burst Strain Error An error in the application of the burst strain was discovered in HOTSPOT. The equation for the application of the burst strain is given as Equation 7-69 in WCAP-16009-P-A (Reference 8). The outer radius of the cladding after burst occurs should be calculated based on the burst strain, and the inner radius of the cladding should be calculated based on the outer radius. In HOTSPOT, the burst strain is applied to the calculation of the cladding inner radius. The cladding outer radius is then calculated based on the inner radius. As such, the burst strain is incorrectly applied to the inner radius rather than the outer radius, which impacts the resulting cladding geometry at the burst elevation after burst occurs. Correction of the erroneous calculation results in thinner cladding at the burst node and more fuel relocating into the burst node, leading to an increase in the Peak Cladding Temperature (PCT) at the burst node. This issue has been evaluated to estimate the impact on existing 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 of WCAP-13451(Reference 7).

The issue described above was evaluated by executing the most limiting plant-specific HOTSPOT runs with a HOTSPOT version that includes the correction of this error. This plant-specific sensitivity study resulted in an estimated PCT impact of 0°F for Watts Bar Unit 2.

E2-4

ENCLOSURE 2 WATTS BAR NUCLEAR PLANT, UNIT 2 10 CFR 50.46 30-DAY REPORT REFERENCES

1. Letter from TVA to NRC, 10 CFR 50.46 Day and Annual Report for 2012, dated April 25, 2013 [ML13120A005]

2. WCAP-17093-P, Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Unit 2 Nuclear Plant Using the ASTRUM Methodology, December 2009

3. WCAP-17093-P, Revision 1, Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Unit 2 Nuclear Plant Using the ASTRUM Methodology, June 2013

4. WBT-D-1460, Final Small Break LOCA Summary Report for WBN Unit 2, January 2010

5. Letter from TVA to NRC, Watts Bar Nuclear Plant, Unit 2 - Emergency Core Cooling system Evaluation Model Changes - 30 Day Report - 10 CFR 50.46 Notification, dated August 28, 2013.[ML13246A076]

6. Letter from TVA to NRC, Watts Bar Nuclear Plant Unit 2 - Fuel Thermal Conductivity Degradation, Response to Supplemental Safety Evaluation Report Open Item 61, August 6, 2013 [ML13225A024]

7. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October 1992

8. WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM), January 2005 E2-5

ENCLOSURE 3 WATTS BAR NUCLEAR PLANT, UNITs 1 and 2 10 CFR 50.46 30-DAY REPORT ADDITIONAL INFORMATION ON THE EVALUATION OF HOTSPOT BURST STRAIN ERROR CORRECTIONS

L94 140203 800 Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, PA 16066 USA WAT-D-12048 January 30, 2014 Mr. Mike Casner Site Engineering Director Tennessee Valley Authority Watts Bar Nuclear Plant P.O. Box 2000 EQB-1A Spring City, TN 37381 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNITS 1 & 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction

Dear Mr. Casner:

Please find attached the following document for your use:

LTR-LIS-14 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction dated January 29, 2014.

If you have any questions, please contact Carmen Teolis at 412-374-2202 or me at 423-697-5052.

Very truly yours, Ronald Kucharski Customer Projects Manager cc: John Pope TVA Kasey Decker TVA Chris Carey TVA Jim Lemons TVA Bob Clark TVA Linda Evans WEC Robin Selph WEC Gerri Thurman WEC Josh Borromeo WEC Amy Colussy WEC Carmen Teolis WEC Ben Solich WEC Westinghouse Non-Proprietary Class 3 Electronically Approved Records are Authenticated in the Electronic Document Management System

Westinghouse Non-Proprietary Class 3 Westinghouse Electric Company Engineering, Equipment and Major Projects 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-2202 Direct fax: (724) 720-0857 e-mail: teoliscd@westinghouse.com Our ref: LTR-LIS-14-38 January 29, 2014 Watts Bar Units 1 and 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction

Dear Sir or Madam:

documents the 10 CFR 50.46 report for the evaluations of the HOTSPOT burst strain error correction in the Watts Bar Units 1 and 2 Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) analyses. Attachment 1 also documents the updated Peak Cladding Temperature (PCT) rackup sheets for Watts Bar Units 1 and 2.

Please contact your LOCA Plant Cognizant Engineer if there are any questions concerning this information.

Author: (Electronically Approved)* Verifier: (Electronically Approved)*

Carmen D. Teolis Kaitlyn M. Musser LOCA Integrated Services I LOCA Integrated Services I Approved: (Electronically Approved)*

Amy J. Colussy, Manager LOCA Integrated Services I Attachment 1: Watts Bar Units 1 and 2 10 CFR 50.46 Report and Peak Cladding Temperature Rackup Sheets for the HOTSPOT Burst Strain Error Correction (9 Pages)

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

Westinghouse Non-Proprietary Class 3 to LTR-LIS-14-38 January 29, 2014 Page 1 of 9 Attachment 1 Watts Bar Units 1 and 2 10 CFR 50.46 Report and Peak Cladding Temperature Rackup Sheets for the HOTSPOT Burst Strain Error Correction (9 pages, including cover page)

Background

An error in the application of the burst strain was discovered in HOTSPOT. The equation for the application of the burst strain is given as Equation 7-69 in WCAP-16009-P-A and in WCAP-12945-P-A. The outer radius of the cladding after burst occurs should be calculated based on the burst strain, and the inner radius of the cladding should be calculated based on the outer radius. In HOTSPOT, the burst strain is applied to the calculation of the cladding inner radius. The cladding outer radius is then calculated based on the inner radius. As such, the burst strain is incorrectly applied to the inner radius rather than the outer radius, which impacts the resulting cladding geometry at the burst elevation after burst occurs. Correction of the erroneous calculation results in thinner cladding at the burst node and more fuel relocating into the burst node, leading to an increase in the Peak Cladding Temperature (PCT) at the burst node. This issue has been evaluated to estimate the impact on existing 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 of WCAP-13451.

Affected Evaluation Models 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect A case representative of Watts Bar Unit 1 was run using HOTSPOT versions which only differ in the burst strain application. Based on the change in the 95th percentile results, estimated PCT effects of 20°F for Reflood 1, and 70°F for Reflood 2 have been established for 10 CFR 50.46 reporting purposes for Watts Bar Unit 1.

The issue described above was evaluated by executing the most limiting plant-specific HOTSPOT runs with a HOTSPOT version that includes the correction of this error. This plant-specific sensitivity study resulted in an estimated PCT impact of 0°F for Watts Bar Unit 2.

to LTR-LIS-14-38 January 29, 2014 Page 3 of 9 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 12, RSG Utility Name: Tennessee Valley Authority Revision Date: 1/20/2014 Composite Analysis Information EM: CQD (1996) Analysis 9/1/1998 Limiting Break Guillotine FQ: 2.5 FdH: 1.65 Fuel: Vantage + SGTP (%): 12 Notes: Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1892 1,2 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS 1 . Vessel Channel DX Error -4 3 2 . MONTECF Decay Heat Uncertainty Error 4 6 3 . Input Error Resulting in Incomplete Solution Matrix 0 7 4 . Tavg Bias Error 8 7 5 . Revised Blowdown Heatup Uncertainty Distribution 5 8 6 . HOTSPOT Fuel Relocation Error 65 11 7 . Accumulator Line/Pressurizer Surge Line Data Evaluation -131 5 8 . Evaluation of Fuel Pellet Thermal Conductivity Degradation and 15 13 (a)

Peaking Factor Burndown B. PLANNED PLANT MODIFICATION EVALUATIONS 1 . Increased Accumulator Temperature Range Evaluation 4 5 2 . 1.4% Uprate Evaluation 12 5 3 . Increased Stroke Time for the ECCS Valves 0 9 4 . Replacement Steam Generators (D3 to 68AXP) -10 10 5 . PBOT/PMID Violation Evaluation 20 12 C. 2013 ECCS MODEL ASSESSMENTS 1 . Revised Heat Transfer Multiplier Distributions -85 14 2 . Error in Burst Strain Application 70 15 D. OTHER*

1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1865

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

Reference 1 . WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," September 1998.

to LTR-LIS-14-38 January 29, 2014 Page 4 of 9 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 12, RSG Utility Name: Tennessee Valley Authority Revision Date: 1/20/2014 Composite 2 . WAT-D-10499, "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 and 2, 10 CFR 50.46 Annual Notification and Reporting for 1997," February 27, 1998.

3 . WAT-D-10618,"Tennessee Valley Authority, Watts Bar Nuclear Plant Units 1 and 2, 10 CFR 50.46 Annual Notification and Reporting for 1998," March 5, 1999.

4 . WAT-D-10725,"Tennessee Valley Authority, Watts Bar Nuclear Plant Unit 1, 10 CFR 50.46 Annual Notification and Reporting for 1999," February 23, 2000.

5 . WAT-D-10840, "Tennessee Valley Authority, Watts Bar Nuclear Plant Unit 1, Final Deliverables for 1.4% Uprate Program," August 31, 2000.

6 . WAT-D-10904, "10 CFR 50.46 Annual Notification and Reporting for 2000," March 2001.

7 . WAT-D-11225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.

8 . WAT-D-11334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.

9 . WAT-D-11285, "Evaluation of Proposed Changes to the Stroke Time for the ECCS Valves," November 2004.

10 . WTV-RSG-06-015, "LOCA & Non-LOCA Analysis Summary for Replacement Steam Generator," February 2006.

11 . LTR-LIS-07-378, "10 CFR 50.46 Reporting Text for HOTSPOT Fuel Relocation Error and Revised PCT Rackup Sheets for Watts Bar Unit 1," June 2007.

12 . LTR-LIS-13-26, "10 CFR 50.46 Report for Watts Bar Unit 1 Cycle 12 RSAC PBOT/PMID Violation Evaluation and Removal of PAD 4.0 Benefit," January 2013.

13 . LTR-LIS-12-413, Watts Bar Units 1 and 2, 10 CFR 50.46 Notification and Reporting for Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown, September 20, 2012.

14 . LTR-LIS-13-354, "Watts Bar Units 1 and 2 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions," July 2013.

15 . LTR-LIS-14-38, "Watts Bar Units 1 and 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction,"

January 2014.

Notes:

(a) This evaluation credits peaking factor burndown, see Reference 13.

to LTR-LIS-14-38 January 29, 2014 Page 5 of 9 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 12, RSG Utility Name: Tennessee Valley Authority Revision Date: 1/20/2014 Reflood 1 Analysis Information EM: CQD (1996) Analysis 9/1/1998 Limiting Break Guillotine FQ: 2.5 FdH: 1.65 Fuel: Vantage + SGTP (%): 12 Notes: Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1656 1,2 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS 1 . Vessel Channel DX Error 56 3 2 . MONTECF Decay Heat Uncertainty Error 4 6 3 . Input Error Resulting in Incomplete Solution Matrix 60 7 4 . Tavg Bias Error 8 7 5 . Revised Blowdown Heatup Uncertainty Distribution 5 8 6 . HOTSPOT Fuel Relocation Error 0 11 7 . Accumulator Line/Pressurizer Surge Line Data Evaluation -37 5 8 . Evaluation of Fuel Pellet Thermal Conductivity Degradation and 114 13 (a)

Peaking Factor Burndown B. PLANNED PLANT MODIFICATION EVALUATIONS 1 . Increased Accumulator Temperature Range Evaluation 4 5 2 . 1.4% Uprate Evaluation 12 5 3 . Increased Stroke Time for the ECCS Valves 0 9 4 . Replacement Steam Generators (D3 to 68AXP) -50 10 5 . PBOT/PMID Violation Evaluation 20 12 C. 2013 ECCS MODEL ASSESSMENTS 1 . Revised Heat Transfer Multiplier Distributions -40 14 2 . Error in Burst Strain Application 20 15 D. OTHER*

1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1832

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

Reference 1 . WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," September 1998.

to LTR-LIS-14-38 January 29, 2014 Page 6 of 9 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 12, RSG Utility Name: Tennessee Valley Authority Revision Date: 1/20/2014 Reflood 1 2 . WAT-D-10499, "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 and 2, 10 CFR 50.46 Annual Notification and Reporting for 1997," February 27, 1998.