ML12117A261

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CFR 50.46 Annual Report for Model Year 2011
ML12117A261
Person / Time
Site: Watts Bar Tennessee Valley Authority icon.png
Issue date: 04/25/2012
From: James Shea
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML12117A261 (18)
v • d • e


Text

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 April 25, 2012 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 2011

Reference:

TVA letter to NRC dated April 27, 2011, "10 CFR 50.46 Annual Report for Model Year 2010" 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 Robert Clark, Senior Licensing Engineer at (423) 365-1818.

?RestfulIly,

Enclosure:

10 CFR 50.46 Annual Report cc page 2 401--vl) k-JOIý K(OL Printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 April 25, 2012 cc (Enclosure):

NRC Regional Administrator- Region I1 NRC Senior Resident Inspector - Watts Bar Nuclear Plant Division of Radiological Health - State of Tennessee

WATTS BAR NUCLEAR PLANT, UNIT 1 10 CFR 50.46 ANNUAL REPORT

1. LOTIC2 ERROR CORRECTIONS (Non-Discretionary Changes)

Background

Errors were discovered and corrected in the LOTIC2 code, which calculates the minimum containment pressure transient for a Large Break Loss-of-Coolant Accident (LOCA) analysis of a plant with an ice condenser containment design. The calculation of the initial specific internal energy of air in the dead ended compartment was corrected to use the appropriate reference temperature, and three calculations of the specific volume of air were corrected to remove a small, extraneous term that was apparently intended to prevent division by zero when there was no air in a given compartment. These changes represent Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.

Affected Evaluation Model(s) 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect Sensitivity calculations showed that correcting the errors had either no effect or a negligible effect on the lower compartment pressure transient, leading to an estimated Peak Cladding Temperature (PCT) impact of 00 F.

2. RADIATION HEAT TRANSFER LOGIC (Non-Discretionary Changes)

Background

Two errors were discovered in the calculation of the radiation heat transfer coefficient in the SBLOCTA computer code. First, existing diagnostics did not preclude non-physical negative or large (negative or positive) radiation heat transfer coefficients from being calculated.

These calculations occurred when the vapor temperature exceeded the cladding surface temperature or when the predicted temperature difference was less than 1 degree. Second, a temperature term incorrectly used degrees Fahrenheit instead of Rankine. These errors have been corrected in the SBLOCTA code and represent a closely related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-1 3451.

Affected Evaluation Model(s) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect A combination of SBLOCTA sensitivity calculations and engineering judgment led to an estimated PCT effect of 0°F for existing Small Break LOCA analysis results.

E-1 of 16

3. MAXIMUM FUEL ROD TIME STEP LOGIC (Non-Discretionary Changes)

Background

An error was discovered in the SBLOCTA code that allowed the fuel rod time step to exceed the specified maximum allowable time step. The time step logic has been corrected in the SBLOCTA code. This change 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 Estimated Effect A combination of SBLOCTA sensitivity calculations and engineering judgment led to an estimated PCT effect of 0°F for existing Small Break LOCA analysis results.

4. GENERAL CODE MAINTENANCE (Discretionary Change)

Background

Various changes have been made to enhance the usability of the codes and to streamline future analyses. Example 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) 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 1999 Westinghouse Realistic 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.

5. WAT CYCLE 11 PBOTIPMID VIOLATION (Discretionary Change)

Background

The Watts Bar Unit 1 Cycle 11 reload core design resulted in violations for the PBOT/PMID limit used in the Large Break LOCA Analysis. These violations were evaluated for Watts Bar Unit 1 Cycle 11 operation. This change represents a Discretionary Change in accordance with Section 4.1.2 of WCAP-1 3451.

E-2 of 16

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model Estimated Effect The impact of the PBOT/PMID violation for Watts Bar Unit 1 Cycle 11 was determined via a plant-specific evaluation to be 20'F for Reflood 1 and Reflood 2.

6. Reference
1. WAT-D-1 1948, "Tennessee Valley Authority Watts Bar Nuclear Plant Units 1 & 2 10 CFR 50.56 Annual Notification and Reporting for 2011," March 2012.

E-3 of 16

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: 2/23/2012 Reflood I 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
7. Accumulator Line/Pressurizer Surge Line Data Evaluation -37 5 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. PMID Violation Evaluation 20 12 C. 2010 ECCS MODEL ASSESSMENTS
1. None 0 D. OTHER
1. None 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," September 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-4 of 16

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: 2/23/2012 Reflood 1 References (Continued):

3. WAT-D-1 0618,"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-1 0725,"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-1 0840, "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-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-5 of 16

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: 2/23/2012 Reflood 2 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
7. Accumulator Line/Pressurizer Surge Line Data Evaluation -131 4 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. PMID Violation Evaluation 20 12.

C. 2010 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," September 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-6 of 16

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: 2/23/2012 Reflood 2 References (Continued):

3. WAT-D-1 0618,"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-1 0725,"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-1 0840, "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-10-118, "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-7 of 16

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: 2/23/2012 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
7. Accumulator Line/Pressurizer Surge Line Data Evaluation -131 4 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. PMID Violation Evaluation 20 12 C. 2010 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," September 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 16

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: 2/23/2012 Composite References (Continued):

3. WAT-D-1 0618,"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-1 0840, "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," March 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-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-9 of 16

Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 11, RSG Utility Name: Tennessee Valley Authority Revision Date: 2/23/2012 Reflood I 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 (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 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 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. PMID Violation Evaluation 20 12 C. 2010 ECCS MODEL ASSESSMENTS
1. None 0 D. OTHER
1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1738

References:

1. WCAP-1 4839, Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," September 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-10 of 16

Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit l Cycle 11, RSG Utility Name: Tennessee Valley Authority Revision Date: 2/23/2012 Reflood 1 References (Continued):

3. WAT-D-1 0618,"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-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-1 1-212, "10 CFR 50.46 Report for the Watts Bar Unit 1 RSCA PBOT/PMID Violation Evaluation and Removal of Temporary Condition Penalties," March 2011.

Notes:

None E-11 of 16

Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts BarUnit 1 Cycle 11, RSG Utility Name: Tennessee Valley Authority Revision Date: 2/23/2012 Reflood 2 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
7. Accumulator Line/Pressurizer Surge Line Data Evaluation -131 4 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/PMIT Violation Evaluation 20 12 C. 2011 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," September 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 16

Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unitl Cycle 11, RSG Utility Name: Tennessee Valley Authority Revision Date: 2/23/2012 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-1 0725,"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," March 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-11-212, "10 CFR 50.46 Report for the Watts Bar Unit 1 RSAC PBOT/PMID Violation Evaluation and Removal of Temporary Condition Penalties," March 2011.

Notes:

None E-13 of 16

Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit l Cycle 11, RSG Utility Name: Tennessee Valley Authority Revision Date: 2/23/2012 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 (OF) 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 4 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. PMID Violation Evaluation 20 12 C. 2010 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," September 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-14 of 16

Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit l Cycle 11, RSG Utility Name: Tennessee Valley Authority Revision Date: 2/23/2012 Com posite References (Continued):

3. WAT-D-1 0618,"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-1 0725,"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-1 0840, "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," March 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-1 1-212, "10 CFR 50.46 Report for Watts Bar Unit 1 RSAC PBOT/PMID Violation Evaluation and Removal of Temporary Condition Penalties," March 2011.

Notes:

None E-15 of 16

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: 2/23/2012 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 LICENSING BASIS Clad Temp (°F) Ref. Notes 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. 2011 ECCS MODEL ASSESSMENTS
1. None 0 D. OTHER
1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1132

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.

Notes:

None E-16 of 16

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