ML103610279

CFR 50.46 Annual Report for Model Year 2009

ML103610279
Person / Time
Site:	Watts Bar
Issue date:	12/21/2010
From:	Krich R Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML103610279 (24)

Text

IFVAI Tennessee Valley Authority 1101 Market Street, LP 3R Chattanooga, Tennessee 37402-2801 R. M. Krich Vice President Nuclear Licensing December 21, 2010 10 CFR 50.4 10 CFR 50.46 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390

Subject:

10 CFR 50.46 Annual Report for Model Year 2009

Reference:

TVA Letter to NRC, "Watts Bar Nuclear Plant Unit 1 -Emergency Core Cooling System Evaluation Model Changes -Annual Notification and Reporting," dated December 30, 2009 The purpose of this letter is to provide the annual report of changes or errors discovered in the emergency core cooling system (ECCS) evaluation model for Watts Bar Nuclear Plant, Unit 1. In accordance with 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems (ECCS) for Light-Water Nuclear Power Reactors," paragraph (a)(3)(ii), the enclosed report describes the nature and the estimated effect on the limiting ECCS analysis of changes or errors discovered since submittal of the reference letter.There are no regulatory commitments in this letter. Please direct questions concerning this issue to Michael Brandon at (423) 365-1824.Respectfully, R. M. Krich

Enclosure:

10 CFR 50.46 Annual Report cc (Enclosure):

NRC Regional Administrator-Region II NRC Senior Resident Inspector

-Watts Bar Nuclear Plant printed on recycled paper WATTS BAR NUCLEAR PLANT UNIT I 10 CFR 50.46 ANNUAL REPORT 1. 1985 WESTINGHOUSE SMALL BREAK LOCA EVALUATION MODEL WITH NOTRUMP There were no changes, error corrections, or enhancements to the 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP. Therefore, there is no associated reporting text related to the 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP for 2009.2. GENERAL CODE MAINTENANCE (Discretionary Change)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)1996Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The nature of these changes leads to an estimated PCT impact of 0°F.3. ERROR IN ASTRUM PROCESSING OF AVERAGE ROD BURNUP AND ROD INTERNAL PRESSURE (Non-Discretionary Change)Background An error was discovered in the processing of the burnup and rod internal pressure inputs for average core rods in ASTRUM analyses.

The correction of this error has been evaluated for impact on current licensing-basis analyses and will be incorporated into the ASTRUM method at a future time. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.Affected Evaluation Model(s)2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM E-1 of 23 Estimated Effect This error was evaluated to have a negligible impact on PCT, leading to an estimated impact of 0°F for 10 CFR 50.46 reporting purposes.4. DISCREPANCY IN METAL MASSES USED FROM DRAWINGS (Non-Discretionary Changes)Background Discrepancies were discovered in the use of lower support plate (LSP) metal masses from drawings.

The updated LSP metal masses have been evaluated for impact on current licensing-basis analysis results and will be incorporated on a forward-fit basis. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s)SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The lower support plate mass error is relatively minor and would be expected to have a negligible effect on the Large Break LOCA analysis results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.5. HOTSPOT GAP HEAT TRANSFER LOGIC (Non-Discretionary Change)Background The HOTSPOT code has been updated to incorporate the following changes to the gap heat transfer logic: (1) change the gap temperature from the pellet average temperature to the average of the pellet outer surface and cladding inner surface temperatures; (2) correct the calculation of the pellet surface emissivity to use a temperature in °R (as specified in Equation 7-28 of Reference

1) instead of 'F; and (3) revise the calculation of the gap radiation heat transfer coefficient to delete a term and temperature adder not shown in or suggested by Equation 7-28 of Reference

1. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s)1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM E-2 of 23 Estimated Effect Sample calculations showed a minimal impact on PCT, leading to an estimated effect of 0°F.Reference 1. WCAP-12945-P-A, Volume 1, Revision 2, "Code Qualification Document for Best Estimate LOCA Analysis, Volume I: Models and Correlations," March 1998.6. HOTSPOT STATISTICAL OUTPUT LOGIC (Non-Discretionary Change)Background The HOTSPOT code has been updated to incorporate the following changes to the statistical output logic for calculations using the Code Qualification Document methodology:

(1) revise one of the three methods for calculating the standard deviation of cladding temperature to correctly identify the bin containing the 97.5th percentile value; and (2) change the 50th, 95th and 97.5th percentile bin values from the lower end of the range to the upper end. These changes represent a closely-related group of Non- Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.Affected Evaluation Model(s)1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Best Estimate Large Break LOCA Evaluation Model, Application to PWRs with Upper Plenum Injection Estimated Effect Sample calculations suggested a minimal impact on the 95th percentile PCT, leading to an estimated effect of 0°F.7. WAT CYCLE 10 PMID VIOLATION Background The Watts Bar Unit 1 Cycle 10 reload core design resulted in several violations of the PMID limit used in the Large Break LOCA Analysis.

These violations were evaluated for Watts Bar, Unit 1 Cycle 10 operation.

This change represents a 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 Estimated Effect The impact of the PMID violation for Watts Bar Unit 1 Cycle 10 was determined via a plant-specific evaluation to be 20°F for Reflood 1 and Reflood 2.E-3 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Cycle 9, RSG Composite Limiting Break Size: Guillotine Analysis Information EM: CQD (1996) Analysis Date: 8/1/1998 FQ: 2.5 FdH: 1.65 Fuel: Vantage + SGTP (%): 12 Notes: Mixed Core -Vantage + / Performance

+ / RFA-2 LICENSING BASIS Analysis-Of-Record PCT PCT ASSESSMENTS (Delta PCT)A. PRIOR ECCS MODEL ASSESSMENTS

1. Vessel Channel DX Error 2. MONTECF Decay Heat Uncertainty Error 3. Input Error Resulting in Incomplete Solution Matrix 4. Tavg Bias Error 5 Revised Blowdown Heatup Uncertainty Distribution

6. HOTSPOT Fuel Relocation Error Clad Temp (*F)Ref. Notes 1892 1,2-4 4 0 8 5 65-131 4 12 0-10 20 3 6 7 7 8 11 4 5 5 9 10 12 B.1.2.3.4.5.6.PLANNED PLANT MODIFICATION EVALUATIONS Accumulator Line/Pressurizer Surge Line Data Evaluation Increased Accumulator Temperature Range Evaluation 1.4% Uprate Evaluation Increased Stroke Time for the ECCS Valves Replacement Steam Generators (D3 to 68AXP)PMID Violation Evaluation C. 2009 ECCS MODEL ASSESSMENTS

1. None D. OTHER*1. None LICENSING BASIS PCT + PCT ASSESSMENTS 0 0 PCT = 1865

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. 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.E-4 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Wafts Bar Unit 1 Cycle 9, RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Composite References (Continued):

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," February 2001.7. WAT-D-11225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.8. WAT-D-1 1334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.9. WAT-D-1 1285, "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-07-893, "10 CFR 50.46 Reporting Text for Watts Bar Unit 1 Cycle 9 RSAC PMID Violation Evaluation and Revised PCT Rackup Sheets," December 2007.Notes: None E-5 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Cycle 9, RSG Reflood 1 Analysis Information EM: CQD (1996) Analysis Date: 8/1/1998 FQ: 2.5 FdH: 1.65 Fuel: Vantage + SGTP (%): 12 Notes: Mixed Core -Vantage + / Performance

+ / RFA-2 Limiting Break Size: Guillotine Clad Temp (OF)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 2. MONTECF Decay Heat Uncertainty Error 3. Input Error Resulting in Incomplete Solution Matrix 4. Tavg Bias Error 5. Revised Blowdown Heatup Uncertainty Distribution

6. HOTSPOT Fuel Relocation Error B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Accumulator Line/Pressurizer Surge Line Data Evaluation

2. Increased Accumulator Temperature Range Evaluation

3. 1.4% Uprate Evaluation

4. Increased Stroke Time for the ECCS Valves 5. Replacement Steam Generators (D3 to 68AXP)6. PMID Violation Evaluation 56 4 60 8 5 0 3 6 7 7 8 11 4 5 5 9 10 12-37 4 12 0-50 20 C. 2009 ECCS MODEL ASSESSMENTS

1. None 0 0 D. OTHER*1. None LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1738

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. WAT-D-1 0499, "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 and 2, 10 CFR 50.46 Annual Notification and Reporting for 1997," February 27, 1998.E-6 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Cycle 9, RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Reflood 1 References (Continued):

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," February 2001.7. WAT-D-1 1225, "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-1 1285, "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-07-893, "10 CFR 50.46 Reporting Text for Watts Bar Unit 1 Cycle 9 RSAC PMID Violation Evaluation and Revised PCT Rackup Sheets," December 2007.Notes: None E-7 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Cycle 9, RSG Reflood 2 Analysis Information EM: CQD (1996) Analysis Date: 8/1/1998 FQ: 2.5 FdH: 1.65 Fuel: Vantage +Notes: Mixed Core -Vantage + / Performance

+ / RFA-2 Limiting Break Size: Guillotine SGTP (%): 12 Clad Temp (OF)LICENSING BASIS Analysis-Of-Record PCT 1892 PCT ASSESSMENTS (Delta PCT)A. PRIOR ECCS MODEL ASSESSMENTS

1. Vessel Channel DX Error 2. MONTECF Decay Heat Uncertainty Error 3. Input Error Resulting in Incomplete Solution Matrix 4. Tavg Bias Error 5 Revised Blowdown Heatup Uncertainty Distribution

6. HOTSPOT Fuel Relocation Error-4 4 0 8 5 65.131 4 12 0-10 20 Ref. Notes 1,2 3 6 7 7 8 11 4 5 5 9 10 12 B.1.2.3.4.5.6.PLANNED PLANT MODIFICATION EVALUATIONS Accumulator Line/Pressurizer Surge Line Data Evaluation Increased Accumulator Temperature Range Evaluation 1.4% Uprate Evaluation Increased Stroke Time for the ECCS Valves Replacement Steam Generators (D3 to 68AXP)PMID Violation Evaluation C. 2009 ECCS MODEL ASSESSMENTS

1. None D. OTHER*1. None LICENSING BASIS PCT + PCT ASSESSMENTS 0 0 PCT = 1865

References:

1. WCAP-1 4839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. 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.E-8 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Cycle 9, RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Reflood 2 References (Continued):

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," February 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-1 1285, "Evaluation of Proposed Changes to the Stroke Time for the ECCS Valves," November 2004.10. VVTV-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-07-893, "10 CFR 50.46 Reporting Text for Watts Bar Unit 1 Cycle 9 RSAC PMID Violation E valuation and Revised PCT Rackup Sheets," December 2007.Notes: None E-9 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Cycle 10, RSG Utility Name: Tennessee Valley Authority Revision Date: 02/04/2010 Composite Analysis Information EM: CQD (1996) Analysis Date: 8/1/1998 Limiting Break Size: 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 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Accumulator Line/Pressurizer Surge Line Data Evaluation

-131 4 2. Increased Accumulator Temperature Range Evaluation 4 5 3. 1.4% Uprate Evaluation 12 5 4. Increased Stroke Time for the ECCS Valves 0 9 5. Replacement Steam Generators (D3 to 68AXP) -10 10 6. PMID Violation Evaluation 20 12 C. 2009 ECCS MODEL ASSESSMENTS

1. None 0 D. OTHER*1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1865

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. 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.E-10 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Cycle 10, RSG Utility Name: Tennessee Valley Authority Revision Date: 02/04/2010 Composite References (Continued):

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," February 2001.7. WAT-D-1 1225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.8. WAT-D-1 1334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.9. WAT-D-1 1285, "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-10-118, "10 CFR 50.46 Reporting Text for Watts Bar Unit 1 Cycle 10 RSAC PMID Violation Evaluation and Revised PCT Rackup Sheets," February 2010.Notes: None E-11 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Utility Name: Tennessee Valley Authority Revision Date: 02/04/2010 Cycle 10, RSG Reflood 1 Analysis Information EM: CQD (1996) Analysis Date: 8/1/1998 FQ: 2.5 FdH: 1.65 Fuel: Vantage + SGTP (%): 12 Notes: Mixed Core -Vantage + / Performance

+ / RFA-2 LICENSING BASIS Analysis-Of-Record PCT PCT ASSESSMENTS (Delta PCT)A. PRIOR ECCS MODEL ASSESSMENTS

1. Vessel Channel DX Error 2. MONTECF Decay Heat Uncertainty Error 3. Input Error Resulting in Incomplete Solution Matrix 4. Tavg Bias Error 5. Revised Blowdown Heatup Uncertainty Distribution

6. HOTSPOT Fuel Relocation Error Limiting Break Size: Guillotine Clad Temp (OF)1656 56 4 60 8 5 0-37 4 12 0-50 20 Ref. Notes 1,2 3 6 7 7 8 11 4 5 5 9 10 12 B.1.2.3.4.5.6.C.1.D.1.PLANNED PLANT MODIFICATION EVALUATIONS Accumulator Line/Pressurizer Surge Line Data Evaluation Increased Accumulator Temperature Range Evaluation 1.4% Uprate Evaluation Increased Stroke Time for the ECCS Valves Replacement Steam Generators (D3 to 68AXP)PMID Violation Evaluation 2009 ECCS MODEL ASSESSMENTS None OTHER*None 0 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1738

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. WAT-D-1 0499, "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 and 2, 10 CFR 50.46 Annual Notification and Reporting for 1997," February 27, 1998.E-12 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Cycle 10, RSG Utility Name: Tennessee Valley Authority Revision Date: 02/04/2010 Reflood 1 References (Continued):

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," February 2001.7. WAT-D-11225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.8. WAT-D-1 1334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.9. WAT-D-1 1285, "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-10-118, "10 CFR 50.46 Reporting Text for Watts Bar Unit 1 Cycle 10 RSAC PMID Violation Evaluation and Revised PCT Rackup Sheets," February 2010.Notes: None E-13 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Cycle 10, RSG Utility Name: Tennessee Valley Authority Revision Date: 02/04/2010 Reflood 2 Analysis Information EM: CQD (1996) Analysis Date: 8/1/2009 Limiting Break Size: 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 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Accumulator Line/Pressurizer Surge Line Data Evaluation

-131 4 2. Increased Accumulator Temperature Range Evaluation 4 5 3. 1.4% Uprate Evaluation 12 5 4. Increased Stroke Time for the ECCS Valves 0 9 5. Replacement Steam Generators (D3 to 68AXP) -10 10 6. PMID Violation Evaluation 20 12 C. 2009 ECCS MODEL ASSESSMENTS

1. None 0 D. OTHER*1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1865

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. 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.E-14 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 Cycle 10, RSG Utility Name: Tennessee Valley Authority Revision Date: 02/04/2010 Reflood 2 References (Continued):

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-1 0904, "10 CFR 50.46 Annual Notification and Reporting for 2000," February 2001.7. WAT-D-1 1225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.8. WAT-D-1 1334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.9. WAT-D-1 1285, "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-10-1 18, "10 CFR 50.46 Reporting Text for Watts Bar Unit 1 Cycle 10 RSAC PMID Violation Evaluation and Revised PCT Rackup Sheets," February 2010.Notes: None E-15 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Composite Analysis Information EM: CQD (1996) Analysis Date: 8/1/1998 Limiting Break Size: 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 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Accumulator Line/Pressurizer Surge Line Data Evaluation

-131 4 2. Increased Accumulator Temperature Range Evaluation 4 5 3. 1.4% Uprate Evaluation 12 5 4. Increased Stroke Time for the ECCS Valves 0 9 5. Replacement Steam Generators (D3 to 68AXP) -10 10 C. 2009 ECCS MODEL ASSESSMENTS

1. None 0 D. OTHER*1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1845

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. WAT-D-1 0499, "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.E-16 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Composite References (Continued):

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," February 2001.7. WAT-D-1 1225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.8. WAT-D-1 1334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.9. WAT-D-1 1285, "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.Notes: None E-17 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Reflood 1 Analysis Information EM: CQD (1996) Analysis Date: 8/1/1998 Limiting Break Size: 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 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Accumulator Line/Pressurizer Surge Line Data Evaluation

-37 4 2. Increased Accumulator Temperature Range Evaluation 4 5 3. 1.4% Uprate Evaluation 12 5 4. Increased Stroke Time for the ECCS Valves 0 9 5. Replacement Steam Generators (D3 to 68AXP) -50 10 C. 2009 ECCS MODEL ASSESSMENTS

1. None 0 D. OTHER*1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT= 1718

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. WAT-D-1 0499, "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.E-18 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Reflood 1 References (continued)

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-1 0904, "10 CFR 50.46 Annual Notification and Reporting for 2000," February 2001.7. WAT-D-1 1225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.8. WAT-D-1 1334, "10 CFR.50.46 Annual Notification and Reporting for 2004," April 2005.9. WAT-D-1 1285, "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.Notes: None E-19 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Reflood 2 Analysis Information EM: CQD (1996) Analysis Date: 8/11/1998 Limiting Break Size: 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 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Accumulator Line/Pressurizer Surge Line Data Evaluation

-131 4 2. Increased Accumulator Temperature Range Evaluation 4 5 3. 1.4% Uprate Evaluation 12 5 4. Increased Stroke Time for the ECCS Valves 0 9 5. Replacement Steam Generators (D3 to 68AXP) -10 10 C. 2009 ECCS MODEL ASSESSMENTS

1. None 0 D. OTHER*1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1845

References:

1. WCAP-14839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," August 1998.2. WAT-D-1 0499, "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.E-20 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar, Unit 1 RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Reflood 2 References (Continued):

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," February 2001.7. WAT-D-11225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.8. WAT-D-1 1334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.9. WAT-D-1 1285, "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.Notes: None E-21 of 23 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Watts Bar, Unit 1 RSG Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Analysis Information EM: NOTRUMP Analysis Date: 5/17/2004 Limiting Break Size: 4 inch FQ: 2.5 FdH: 1.65 Fuel: RFA-2 SGTP (%): 12 Notes: Mixed Core -Vantage + / Performance

+ / RFA-2 Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1132 1 PCT ASSESSMENTS (Delta PCT)A. PRIOR ECCS MODEL ASSESSMENTS

1. None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Increased Stroke Time for the ECCS Valves 0 2 C. 2009 ECCS MODEL ASSESSMENTS

1. None 0 D. OTHER*1. Leaking SIS Relief Valve 120 3 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1252

References:

1. WTV-RSG-06-015, "LOCA & Non-LOCA Analysis Summary for Replacement Steam Generator," February 2006.2. WAT-D-1 1285, "Evaluation of Proposed Changes to the Stroke Time for the ECCS Valves," November 2004.3. WAT-D-1 1360, "Safety Injection Pump Discharge Relief Valve Leakage Evaluation," July 2005.Notes: None E-22 of 23 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Plant Name: Watts Bar, Unit 2 Future Utility Name: Tennessee Valley Authority Revision Date: 01/27/2010 Analysis Information EM: ASTRUM (2004) Analysis Date: 10/14/2009 FQ: 2.5 FdH: 1.65 Fuel: RFA-2 SGTP (%): 10 Notes: LICENSING BASIS Analysis-Of-Record PCT PCT ASSESSMENTS (Delta PCT)A. PRIOR ECCS MODEL ASSESSMENTS

1. None B.- PLANNED PLANT MODIFICATION EVALUATIONS

1. None C. 2009 ECCS MODEL ASSESSMENTS

1. None D. OTHER*1. None LICENSING BASIS PCT + PCT ASSESSMENTS Limiting Break Size: Split Clad Temp (OF)1552 0 0 0 0 PCT = 1552 Ref. Notes I

References:

1. WCAP-17093-P, Revision 0, "Best-Estimate Analysis of the Large-Break Loss-of-Coolant Accident for Watts Bar Unit 2 Nuclear Power Plant Using the ASTRUM Methodology," December 2009.Notes: None E-23 of 23