ML081980033

Emergency Core Cooling System Evaluation Model Changes - Annual Notification and Reporting

ML081980033
Person / Time
Site:	Watts Bar
Issue date:	07/02/2008
From:	Brandon M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML081980033 (24)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 July 2, 2008 10 CFR 50.46 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:

In the Matter of

)

Docket No. 50-390 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL CHANGES - ANNUAL NOTIFICATION AND REPORTING

References:

(1) Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - 30 Day Report and Annual Notification and Reporting for 2006, dated July 3, 2007.

(2) Westinghouse Letter LTR-LIS-08-66 to TVA dated January 31, 2008 -

10 CFR 50.46 Annual Notification and Reporting for 2007.

(3) Westinghouse Letter LTR-LIS-08-292 to TVA dated April 14, 2008 -

10 CFR 50.46 Report for CCFL Global Volume Error.

(4) Westinghouse Letter LTR-LIS-08-385 to TVA dated June 2, 2008 -

10 CFR 50.46 Report for Errors in Reactor Vessel Lower Plenum Surface Area Calculations.

This letter provides the annual update report required by 10 CFR 50.46. The enclosed information addresses changes or errors in the WBN ECCS evaluation model that affect calculation of peak cladding temperature (PCT). This report covers the period from WBN's last 10 CFR 50.46 annual report, which was submitted by letter in Reference 1, through June, 2008. WBN's ECCS evaluation model is contractually maintained by Westinghouse Electric Company, who provided the enclosed updates.

The changes to the model that have been made since our last update are described in. The PCT margin allocations resulting from the changes listed in Enclosure 1 are summarized in Enclosure 2. This update includes a 201F penalty for Cycle 9, the current cycle, only. Both Cycle 9 and general rackup sheets are provided in Enclosure 2

U.S. Nuclear Regulatory Commission Page 2

.JUL 0 2 2008 There are no regulatory commitments associated with this submittal. If you have any questions concerning this matter, please call me at (423) 365-1824.

,Sincerely, M. K. Brandon Manager, Site Licensing and Industry Affairs Enclosures cc (Enclosures):

NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 ATTN: Patrick D. Milano, Project Manager U.S. Nuclear Regulatory Commission Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation MS 0 8H4 Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission Region 11 Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303

ENCLOSURE I Changes to ECCS Evaluation Model E1-1

- 10 CFR 50.46 Reporting Text Our ref: LTR-LIS-08-66 January 31, 2008 ERRORS IN REACTOR VESSEL NOZZLE DATA COLLECTIONS (Non-Discretionary Change)

Background

Some minor errors were discovered in the-reactor vessel nozzle data collections that potentially affect the vessel inlet and outlet nozzle fluid volume, metal mass and surface area. The corrected values-have been evaluated for impact on current licensing-basis analysis results and will be incorporated into the plant-specific input databases on a forward-fit basis. 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) 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The differences in the vessel inlet and outlet nozzle fluid volume, metal mass and surface area are relatively minor and would be expected to-produce a negligible effect on large break and small break LOCA-analysis results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.

E1-2

Attachment I - 10 CFR 50.46 Reporting Text Our ref: LTR-LIS-08-66 January 31, 2008 PUMP WEIR RESISTANCE MODELING (Non-Discretionary Change)

Background

Review of the reactor coolant:pump data-collections identified instances of either including a weir resistance for a design without a weir or double-counting the weir resistance for a design with a weir. The corrected resistances have been evaluated for impact on existing analysis results and will be incorporated into the plant-specific input databases 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) 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect Resolving the identified discrepancies has been evaluated as having a negligible effect on existing results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.

E1-3

1. - 10 CFR 50.46 Reporting Text Our ref: LTR-LIS-08-66 January 31, 2008 GENERAL CODE MAINTENANCE (Discretionary Change)

Background

Various changes have been made to enhance the usability of the codes and to help preclude errors in analyses. This includes items such as 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) 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The nature of these changes leads to an estimated PCT impact of O°F.

E1-4

Attachment to LTR-LIS-08-292 April 14, 2008 CCFL GLOBAL VOLUME ERROR (Non-Discretionary Change)

Background

An error was identified during the course of a recent Best Estimate Large Break LOCA analysis in which the volume between the core barrel and the baffle plates in the CCFL region above the active fuel length was modeled incorrectly. The corrected values have been evaluated for impact on the current licensing-basis analysis results. This error represents a non-discretionary change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model(s) 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM Estimated Effect The CCFL global volume modeling error has been generically eyaluated to have a negligible impact on PCT for affected analyses and a penalty of 0 'F is assigned.

E1-5

-I Attachment to LTR-LIS-08-385 June 2, 2008 ERRORS IN REACTOR VESSEL LOWER PLENUM SURFACE AREA CALCULATIONS (Non-Discretionary Change)

Background

Two errors were discovered in the calculations of reactor vessel lower plenum surface area. The corrected values have been evaluated for impact on current licensing-basis analysis results and will be incorporated on a forward-fit basis. 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) 1981 Westinghouse Large Break LOCA Evaluation Model with BASH 1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP Estimated Effect The differences in vessel lower plenum surface area are relatively minor and would be expected to produce a negligible effect on large and small break LOCA analysis results, leading to an estimated PCT impact of 0°F for 10 CFR 50.46 reporting purposes.

E1-6

Attachment I - 10 CFR 50.46 Reporting Text Our ref: LTR-LIS-08-66 January 31, 2008 Plant Specific Text E1-7

- 10 CFR 50.46 Reporting Text Our ref. LTR-LIS-08-66 January 31, 2008 WAT CYCLE 9 PMID VIOLATION (Discretionary Change)

Background

The Watts Bar Unit 1 Cycle 9 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 9 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 I Cycle 9 was determined via a plant-specific evaluation to be 20'F for Reflood 1 and Reflood 2.

E1-8

ENCLOSURE 2 PCT Rackup Sheets E2-1

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 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:

1/15/08 C om posite Analysis Information EM:

CQD (1996)

Analysis Date:

8/1/98 Limiting Break Size:

Guillotine FQ:

2.5 FdH:

1.65 Fuel:

Vantage +

SGTP (%):

12 Notes:

Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (IF)

Ref.

Notes LICENSING BASIS Analysis-Of-Record PCT 1892 1,2 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I. 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 B. PLANNED PLANT MODIFICATION EVALUATIONS I

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. 2007 ECCS MODEL ASSESSMENTS I. HOTSPOT Fuel Relocation Error 65 11 D. OTHER*

1. None 0

LICENSING BASIS PCT + PCT ASSESSMENTS PCT=

1865 It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I 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.

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

E2-2

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 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:

1/15/08 C om posite 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-l 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 I," June 2007.

12. LTR-LIS-07-893, "10 CFR 50.46 Reporting Text for Watts Bar Unit I Cycle 9 RSAC PMID Violation Evaluation and Revised PCT Rackup Sheets," December 2007.

Notes:

None E2-3

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit I Cycle 9, RSG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 R eflood 1 Analysis Information EM:

CQD (1996)

Analysis Date:

8/1/98 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 I

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 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 6. PMID Violation Evaluation 20 12 C. 2007 ECCS MODEL ASSESSMENTS I. HOTSPOT Fuel Relocation Error 0

11 D. OTHER*

1. None 0

LICENSING BASIS PCT + PCT ASSESSMENTS PCT=

1738 It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I 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 I 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 I and 2, 10 CFR 50.46 Annual Notification and Reporting for 1998," March 5, 1999.

E2-4

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 3 1, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit I Cycle 9, RSG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 Reflood 1 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 I Cycle 9 RSAC PMID Violation Evaluation and Revised PCT Rackup Sheets," December 2007.

Notes:

None E2-5

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit I Cycle 9, RSG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 R eflood 2 Analysis Information EM:

CQD (1996)

Analysis Date:

8/1/98 Limiting Break Size:

Guillotine FQ:

2.5 FdH:

1.65 Fuel:

Vantage +

SGTP (%):

12 Notes:

Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (IF)

Ref.

Notes LICENSING BASIS Analysis-Of-Record PCT 1892 1,2 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I

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 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. 2007 ECCS MODEL ASSESSMENTS I. HOTSPOT Fuel Relocation Error 65 11 D. OTHER*

I.None 0

LICENSING BASIS PCT + PCT ASSESSMENTS PCT =

1865 It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I 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 I 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 I and 2, 10 CFR 50.46 Annual Notification and Reporting for 1998," March 5, 1999.

E2-6

- PCT Rack-up Sheets Our ref: LTR-LIS-08-66 January 31, 2008 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:

1/15/08 Reflood 2 4

WAT-D-10725,"Tennessee Valley Authority,,Watts BarNuclear 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 I," 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 E2-7

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit 1 R SG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 C om posite Analysis Information EM:

CQD (1996)

Analysis Date:

8/1/98 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 I

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 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. 2007 ECCS MODEL ASSESSMENTS I. HOTSPOT Fuel Relocation Error 65 11 D. OTHER*

I. None 0

LICENSING BASIS PCT + PCT ASSESSMENTS PCT =

1845 It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I 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 I 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.

E2-8

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit 1 R SG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 C om posite 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 I," June 2007.

Notes:

None E2-9

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit I R SG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 R eflood 1 Analysis Information EM:

CQD (1996)

Analysis Date:

8/1/98 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 I

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 B. PLANNED PLANT MODIFICATION EVALUATIONS I

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. 2007 ECCS MODEL ASSESSMENTS I. HOTSPOT Fuel Relocation Error 0

11 D. OTHER*

I None 0

LICENSING BASIS PCT + PCT ASSESSMENTS PCT=

1718 It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

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 I 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.

E2-10

- PCT Rack-up Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit 1 R SG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 R eflood 1 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-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 I," June 2007.

Notes:

None E2-11

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit 1 R SG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 R eflood 2 Analysis Information EM:

CQD (1996)

Analysis Date:

8/1/98 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 I

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 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. 2007 ECCS MODEL ASSESSMENTS I. HOTSPOT Fuel Relocation Error 65 11 D. OTHER*

1. None 0

LICENSING BASIS PCT + PCT ASSESSMENTS PCT =

1845 It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I 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 I 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 I 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.

E2-12

- PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name:

Watts Bar Unit 1 R SG Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 R eflood 2 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 I," June 2007.

Notes:

None E2-13

.1L - PCT Rackup Sheets Our ref: LTR-LIS-08-66 January 31, 2008 Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name:

Watts Bar Unit I Utility Name:

Tennessee Valley Authority Revision Date:

1/15/08 Analysis Information EM:

NOTRUMP Analysis Date:

5/17/04 FQ:

2.5 FdH:

1.65 Fuel:

RFA-2 SGTP (%):

12 Notes:

Mixed Core - Vantage + / Performance + / RFA-2 LICENSING BASIS Analysis-Of-Record PCT PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I. None B. PLANNED PLANT MODIFICATION EVALUATIONS I. Increased Stroke Time for the ECCS Valves C. 2007 ECCS MODEL ASSESSMENTS I

None D. OTHER*

1. Leaking SIS Relief Valve RSG Limiting Break Size:

4 inch Clad Temp ('F)

Ref.

Notes 1132 1

0 o

2 0

120 3

LICENSING BASIS PCT + PCT ASSESSMENTS PCT =

1252 It is recommended that the licensee determine if these PCT allocations be considered with respect to 10 CFR 50.46 reporting requirements.

References:

I 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-11360, "Safety Injection Pump Discharge Relief Valve Leakage Evaluation," July 2005.

Notes:

None E2-14