ML061040280

Emergency Core Cooling System (ECCS) Evaluation Model Changes - Annual Notification and Reporting for 2005

ML061040280
Person / Time
Site:	Watts Bar
Issue date:	04/12/2006
From:	Pace P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML061040280 (25)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 APR 1 2 2006 1O CFR 50.6 U. S. Nuclear Regulatory Commission ATrN: Document Control Desk Washington, D.C. 20555-0001 Gentlemen:

In the Matter of ) Docket No. 50-,90 Tennessee Valley Authority WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - EMERGENCY CORE COOLING SYSTEM (ECCS) EVALUATION MODEL CHANGES - ANNUAL NOTIFICATION AN]D REPORTING FOR 2005

References:

1. 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.

2. Watts Bar Nuclear Plant (WBN) Unit 1 -

Emergency Core Cooling System (ECCS)

Evaluation Model Changes - 30 Day Report -

10 CFR 50.46 Notification, dated August 16, 2005.

3. Watts Bar Nuclear Plant (WBN) Unit 1 -

Emergency Core Cooling System (ECCS)

Evaluation Model Changes - 30 Day Report and Annual Report, dated December 18, 2003.

This letter provides information that fulfils the annual reporting requirements of 10 CFR 50.46. The enclosed inEormation addresses changes or errors in the WBN ECCS evaluation model that affect the calculation of peak cladding temperature (PCT). WBN's ECCS evaluation model is contractually maintained by Westinghouse Electric Corporation and the last 10 CFR 50.46 annual report for WBN was submitted Ainfedonroc-ydepaWe

U.S. Nuclear Regulatory Commission Pagse 2 APR 12 2006 in a letter in Reference 1 and a subsequent 30 day report in Reference 2. The changes to the model that have been made since that time are described in Enclosure 1. The PCT margin allocations resulting from the changes listed in Enclosure 1 are summarized in Enclosure 2.

This submittal also completes the commitment in Reference 3 to perform a revised Small Break Loss-of-Coolant Accident (LOCA) analysis as part of the replacement steam generator project.

The Rackup Sheets for Cycle 8 Small Break LOCA which includes the replacement steam generators are also provided in .

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

S13c rely, P. L. Pace Manager, Site Licensing and Industry Affairs Enclosures cc (Enclosures):

NRC Resident Inspector Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 Mr. D. V. Pickett, Project Manager U.S. Nuclear Regulatory Commission MS 08G9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2738 U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303

ENCLOSURE 1 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 EMERGENCY CORE COOLING SYSTEM (ECCS)

PEAK CLAD TEMPERATURE (PCT) ANNUAL REPORT

SUMMARY

OF CHANGES BEST ESTIMATE LARGE BREAK - CODE QUALIFICATION DOCUMENT (CQD)

(1996) RELATED ITEMS

1. :REVISED ITERATION ALGORITHM FOR CALCULATING THE AVERAGE FUEL TEMPERATURE (Enhancements/Forward-Fit Discretionary Change

Background

'Under certain conditions, the iteration scheme to calculate an average fuel temperature in HOTSPOT converged slowly, exceeding the maximum iteration count. This led to an

-average fuel temperature calculation that was inconsistent with the WCOBRA/TRAC temperature for calculating the stored energy in the fuel. A revised iteration scheme, based on a Combination of a secant method and a parabolic interpolation with a bracketing scheme, was implemented to resolve the non-convergence issue. This change is considered to be a Discretionary change in accordance with Section 4.1.1 of WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting."

Affected Evaluation Models 1996 Westinghouse (W) 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

'rhe prior inconsistencies between the WCOBRA/TRAC temperature and the HOTSPOT average fuel temperature always resulted in a higher HOTSPOT average fuel temperature.

Therefore, a 0 (zero) degree Fahrenheit (F) impact is

• Conservatively assigned for 10 CFR 50.46 reporting purposes.

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2. PELLET RADIAL PROFILE OPTION (Enhancements/Forward-Fit Discretionary Change)

Background

The radial power profile of fuel pellets was previously assumed to be uniform when setting up the conduction network over the fuel pellet in HOTSPOT. However, the accuracy of this approximation decreases for highly burned fuel since the radial power profile tends to increase from the cente:-:

towards the outside of the fuel pellet at higher burnups.

As such, an option was added in HOTSPOT to use a non-uniform radial power profile consistent with the WCOBRA/TRAC code.

These changes were considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Models 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

'rhis change is for forward-fit purposes only, and has no effect on existing analyses.

3. IMPROVED AUTOMATION OF END OF BLOWDOWN TIME (Enhancements/Forward-Fit Discretionary Change)

Background

Heat transfer multipliers are considered in the uncertainly methodology as a function of the time period in the

ransient. The blowdown cooling heat transfer multipliers are applied during the time period following turnaround of

he blowdown heatup through the end of blowdown. For simplicity, the end of blowdown was originally defined as

-:he time when the system pressure dropped below 40 pounds per square inch atmosphere (psia). This definition was then Later improved by defining end of blowdown based on the time at which the system pressure stops decreasing. This definition has been further revised in order to improve the automated selection of the end of blowdown time. The

revised definition for the end of blowdown was improved by replacing system pressure stops decreasing criterion with a selection based on the time when the collapsed liquid level in the lower plenum reaches a minimum and begins to increase again. These changes were considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-13451.

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Affected Evaluation Models 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 correct end of blowdown time was selected in all prior analyses. Therefore, the estimated effect is zero degrees.

4. GENERAL CODE MAINTENANCE (Enhancements/Forward-Fit Discretionary Change)

Background

A number of coding changes were made as part of normal code maintenance. Examples include more descriptive file naming, improved automation in the ASTRUM codes, and improved input diagnostics in the WCOBRA/TRAC code. All of these changes are considered to be Discretionary changes in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Models 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 None of these changes affect the results of design basis analyses. Therefore, the estimated effect is zero degrees.

5. 'THERMODYNAMIC PROPERTIES FROM THERMO (Enhancements/Forward-Fit Discretionary Change)

Background

Subroutine THERMO supplies the thermodynamic properties for the WCOBRA/TRAC one-dimensional components. It is stated in Section 10 of WCAP-12945-P-A and WCAP-16009-P-A that THERMO supplies the thermodynamic properties valid for temperatures within the following range:

280K < T1 < 697K E1-3

However, the thermodynamic properties supplied by THERIMO are actually valid for temperatures within the following range:

277K < T, < 647K This is not a change in the methodology, but rather, a correction of the documentation. This change is considered to be Discretionary change in accordance with Section 4.1.1 of WCAP-13451.

Affected Evaluation Models 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 This change does not affect the results of design basis analyses. Therefore, the estimated effect is zero degrees.

6. PRESSURIZER FLUID VOLUMES (Non-Discretionary Change)

Background

Westinghouse has recommended that the previously transmitted pressurizer fluid volumes be replaced with nominal cold values. This change resolves a discrepancy in the prior calculations while providing a close approximation of the actual as-built values. The revised values have been evaluated for impact on current licensing basis analyses 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 Models 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.

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Estimated Effect The differences between the previously transmitted and revised volumes are very small and would be expected to produce a negligible effect on large and small break LOCA analysis results, leading to an estimated zero degree PCT impact.

7. VESSEL UNHEATED CONDUCTOR NODING (Non-Discretionary Change)

Background

A discrepancy was identified in a 1996 Westinghouse Best Estimate Large Break LOCA (BE LBLOCA) Evaluation Model analysis whereby some unheated conductors used node sizes that are inconsistent with the analysis input guidelines.

Inspection of selected other analyses using this Evaluation Model identified similar occurrences, and evaluations were completed to estimate the effect of these differences on typical large break LOCA analysis results. These changes represent a closely-related group of Non-Discretionary Changes in accordance with Section 4.1.2 of WCAP-13451.

Affected 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.

Estimated Effect Representative plant calculations using the 1996 Westinghouse BE LBLOCA Evaluation Model indicated that correcting the unheated conductor node sizes resulted in a small reduction in PCT that will conservatively be assigned a 0 degrees F effect for 10 CFR 50.46 reporting purposes.

Similar effects would be expected for the 1999 Westinghouse

]BE LBLOCA Evaluation Model for plants with Upper Plenum Injection, and analyses using this Evaluation Model will also be assigned a 0 degree F PCT effect. The 2004 Westinghouse BE LBLOCA Evaluation Model with ASTRUM is unaffected, since the discrepancy was identified prior to completion of the initial plant application.

8. CONTAINMENT RELATIVE HUMIDITY ASSUMPTION (Non-Discretionary Change)

Background

Large Break LOCA analyses have historically used maximum initial relative humidity to specify the initial containment El-5

air and steam partial pressures. This assumption is conservative for a given total initial containment pressure, but is non-conservative for a given initial containment air partial pressure. The historical assumption has been revised accordingly. This change represents a Non-Discretionary Change in accordance with Section 4.1.2 of WCAP-13451.

Affected Evaluation Model 1996 Westinghouse Best Estimate Large Break LOCA Evaluation Model.

1999 Westinghouse Best Estimate Large Break LOCA EvaluatiDn Model, Application to PWRS with Upper Plenum Injection.

SECY UPI WCOBRA/TRAC Large Break LOCA Evaluation Model.

Estimated Effect An evaluation for WBN concluded that no PCT assessments are required, leading to an estimated PCT effect of 0 degree F.

APPENDIX K SMALL BREAK - NOTRUMP RELATED ITEMS

1. Pressurizer Fluid Volumes (Non-Discretionary Change)

Background

Westinghouse has recommended that the previously-transmitted pressurizer fluid volumes be replaced with nominal cold values. This change resolves a discrepancy in the prior calculations while providing a close approximation of the actual as-built values. The revised values have been evaluated for impact on current licensing-basis analyses 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 Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH.

1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP.

Estimated Effect The differences between the previously transmitted and revised volumes are very small and would be expected to produce a negligible effect on large and small break LOCA E1-6

(SBLOCA) analysis results, leading to an estimated PCT impact of 0 degrees F for 10 CFR 50.46 reporting purposes.

2. LOWER GUIDE TUBE ASSEMBLY WEIGHT (Non-Discretionary Change)

Background

An error was discovered in the lower guide tube assembly weight for three units that resulted in a small over-estimation of the upper plenum metal mass. The corrected values have been evaluated for impact on current licensing-basis analyses 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 Models 1981 Westinghouse Large Break LOCA Evaluation Model with BASH.

1985 Westinghouse Small Break LOCA Evaluation Model with NOTRUMP.

Estimated Effect The differences in upper plenum metal mass are very small 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 degrees F for 10 CFR 50.46 reporting purposes.

3. DISCREPANCY IN NOTRUMP RWST DRAINDOWN CALCULATION (Non-Discretionary Change)

Background

For SBLOCA calculations where the break size is greater than the safety injection (SI) line diameter, and where the SI line is connected directly to the reactor coolant system (RCS), it is assumed that the broken loop SI flows do not inject to the RCS, but rather spill to containment.

Typically, this is modeled in NOTRUMP-EM analyses by setting the flows injected to the broken loop equal to zero, which neglects the continued depletion of the refueling water storage tank (RWST) inventory. As a result, the RWST draindown time is incorrectly calculated, potentially resulting in an inaccurate modeling of enthalpy changes and/or SI interruptions that can occur at switchover to sump recirculation. Therefore, the SI spilling flows need to be explicitly modeled in order to correctly calculate the RWST draindown time.

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Affected Evaluation Models 1985 Westinghouse Small Break LOCA Evaluation Model with WOTRUMP.

Estimated Effect For Westinghouse plants using the NOTRUMP-EM, the larger small breaks are typically non-limiting and the transients are of short duration. Therefore, correct modeling of the spilling flows in the RWST draindown calculation for these breaks would be expected to produce a negligible effect on SBLOCA results, leading to an estimated PCT impact of 0 degrees F for 10 CFR 50.46 reporting purposes.

4. GENERAL CODE MAINTENANCE (Enhancements/Forward-Fit Discretionary Change)

Background

Various changes in code input and output format have been made to enhance usability and help preclude errors in analyses. This includes both input changes (e.g., more relevant input variables defined and more common input values used as defaults) and input diagnostics designed to preclude unreasonable values from being used, as well as various changes to code output which have effect on calculated results. In addition, various updates were made to eliminate inactive coding, improve active coding, and enhance commenting, both for enhanced usability and to facilitate code debugging when necessary. 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 Models 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 0 degrees F.

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ENCLOSURE 2 WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 EMERGENCY CORE COOLING SYSTEM (ECCS)

PEAK CLAD TEMPERATURE (PCT) ANNUAL REPORT RACKUP SHEETS

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Wcstinghouse LOCA Peak Clad Tcmperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit I Cycle 7, OSG Utility Name: Tennessee Valley Authority Revis.ion Date: 2/23/06 Composite Analv:is Information EM: CQD (1996) Analysis Date: 8/1/98 Limiting Break Size: Guillotine FQ: 2.5 FlIIl: 1.65 Fuel: Vantage + SGTP (%): 10 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 aCannel DX Error .4 3

2. MONTECFDecaylleatUncertaintyError 4 6 3 . Input Error Resulting in Incomplete Solution Matrix 0 7 4 . Tavg Dias 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 C. 2005 ECCS MODEL ASSESSMENTS

1. None 0 D. OTIIER*

1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1790

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

References:

1 . WVCAP-14839, Rev. 1, "Best Estimate Analysis of die Large Break Loss of Coolant Accident for tie Watts Bar Nuclear Plan."

August 1998.

2 . WAT-D-10499, "Tennessee Val1cyAudtoritylVatts Bar NuclearPlant 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 . VAT-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 . VAT-D-10840, "Tennessee Valley Authority, Watts Bar Nuclear Plant Unit 1,Final Deliverables for 1.4% Uprate Program,"

August31, 2000.

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westinghousc LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit I Cycle 7, OSG Utility Name: Tennessee Valley Authority Revision Date: 2 /23106 Composite 6 . NVAT-D-10904, "10 CFR 50.46 Annual Notification and Reporting for 2000," February 2001.

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

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

9 . VAT-D-11285, "Evaluation of Proposed Changes to the Stroke Time for the ECCS Valves," November2004.

Notes:

None

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westiingliouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Nanie: Watts Bar Unit I Cycle 7, OSG Utility Name: Tennessee Valley Authority Revision Date: 2 /23/06 Reflood 1 Analysis Information EM: CQD (1996) Analysis Date: 8/1/98 Limiting Break Size: Guillotine FQ: 2.5 Fdll: 1.65 Fuel: Vantage + SGTP(%): 10 Notes: Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (0 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. MONTECFDecaylleatUncertainityError 4 5 3 . Input Error Resulting in Incomplete Solution Matrix 60 6

4. Tavg Bias Error 8 6 5 . Rcvised Blowdown leatup Uncertainty Distribution 5 7 B. PLANNED PLANT MODWICATION EVALUATIONS I Accumulator Line/Pressurizer Surge Line Data Evaluation -37 4 2 . Increased Accumulator Temperature Range Evaluation 4 4 3 . 1.4% Uprate Evaluation 12 4 4 . Increascd Stroke Time for the ECCS Valves 0 8 C. 2005 ECCS MODEL ASSESSMENTS

1. None 0 D. OTIIER*

1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1768

• 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 Plan.,"

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

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

August 31, 2000.

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

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 7, OSG' Utility Name: Tennessee Valley Authority Revison Date: 2/23/06 Reflood l 6 . VAT-D-11225, "10 CFR 50.46 Annual Notification and Reporting for2003," March 2004.

7 . VAT-D-11334, "10 CFR 50.46 Annual NotificationandRcportingfor2004, "April2005.

8 . VAT-D- 11285, "Evaluation of Proposed Chaniges to the Strokc Time for the ECCS Valves," Novernbcr 2004.

Notes:

None

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westinighouse LOCA Peak Clad Temperature Sumninary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 7, OSG(

Utility Name: Tennessee Valley Authority Revision Date: 2 /23/06 Reflood ,2 Analv is Information EM: CQD (1996) Analysis Datc: 8/1/98 Limiting Break Size: Guillotine FQ: 2.5 Fdll: 1.65 Fuel: Vantage + SGTP(%): 10 Notes: Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (*F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1892 1,2 PCT ASSESSMENTS (Delta PCI)

A. PRIOR ECCS MODEL ASSESSMENTS I. Vessel Channel DX Error -4 3

2. MIONTECFDecayh'eat UncertaintyError 4 6 3 . Input Error Resulting in Incomplete Solution MatrLx 0 7

4. Tavg Bias Error 8 7 5 . Revised Blowdown licatup Uncertainty Distribution 5 8 B. PLANNED PLANT MODIFICATION EVALUATIONS I . AccumulatorLinclPressurizerSurgeLine 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 C. 2005 ECCS MODEL ASSESSMENTS I None 0 D. OTIIER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1790

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

Reternces:

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.

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

August 31, 2000.

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 7, OS('r Utility Name: Tennessee Valley Authority Revision Date: 2/23/06 Reflood 2.

6 . VAT-D-10904, "10 CFR 50.46 Annual Notification and Rcporting 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," April2005.

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

Notes:

None

Attachment 2 of LTR-LIS-06-117- PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plait Name: Watts Bar Unit I Cycle 8, RSG Utility Name: Tennessee Valley Authority Revis ion Date: 2 /27/06 Composite AnnJs Information EM: CQD (1996) Analysis Date: 8/1/98 Limiting Break Size: Guillotine FQ: 2.5 Frill: 1.65 Fuel: Vantage + SGTP(%): 10 Notes: Mixed Core - Vantage + / Performance + / RFA-2, RSG (12% SGTP)

Clad Temp (°F) Ref. Not es LICENSING BASIS Analysis-Of-Record PCT 1892 1,2 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I. Vessel aCannel DX Error -4 3 2 . IONTECF Decay lleat Uncertainty Error 4 6 3 . Input Error Resulting in IncompIctc Solution lMatrix 0 7 4 . Tavg Bias Error 8 7 5 . RevisedBlowdownlIeatupUncertaintyDistribution 5 8 B. PLANNED PLANT MODIFICATION EVALUATIONS I . Accumulator Line/Prcssurizer Surgeine 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. 2005 ECCS MODEL ASSESSMENTS I. None 0 D. OTIIER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1780

• 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 Plan,"

August 1998.

2 . WVAT-D-10499, "Tennessee Valley Authority Vatts Bar Nuclear Plant Units 1 and 2, 10 CFR 50.46 Annual Notification and Reporting for 1997," February 27, 1998.

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

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

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit I Cycle 8, RSG Utility Namne: Tennessee Valley Authority Revision Date: 2 /27/06 Composite 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 for2003," March 2004.

8 . WAT-D-11334, "10 Cr-R 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.

Notes:

None

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Wcstinghouse LOCA Pcak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit I Cycle 8, RSG Utilityq Name: Tennessee Valley Authority Revision Date: 2/27/06 Reflood 1 Analvdis Information EM: CQD (1996) Analysis Date: 8/1/98 Limiting Break Size: Guillotine FQ: 2.5 Fdll: 1.65 Fuel: Vantage + SGTP (%): 10 Notes: Mixed Core - Vantage + / Performance + / RFA-2, RSG (12% SGTP)

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 llcatup Uncertainty Distribution 5 8 B. PLANNED PLANT MODIFICATION EVALUATIONS I Accumulator Line/Prcssurizer 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. 2005 ECCS MODEL ASSESSMENTS I. None 0 D. OTIIER*

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:

I . WCAP-14839. Rev. 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for thecVatts Bar Nuclear Plan,,"

August 1998.

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

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

Attachment 2 of LTR-LIS-06-117- PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Nanme: Watts Bar Unit 1 Cycle 8, RSGC Utilit y Name: Tennessee Valley Authority Revision Date: 2/27/06 Reflood 1 5 .VAT-D-10840, "Tennessee Vallcy Authority, Watts Bar NuclearPlant Unit 1, Final Deliverables for 1.4% Uprate Program,'

August 31.2000.

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

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

10 . NVTV-RSO-06-015, "LOCA & Non-LOCA Analysis Surnnary for Replacement Steam Generator," February 2006.

Notes:

None

Attachment 2 of LTR-LIS-06-117- PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Summary for Best Estimate Large Break Plant Name: Watts Bar Unit 1 Cycle 8, RS G Utility Name: Tennessee Valley Authority Revision Date: 2/27/06 Reflood 2 Analyd-s Information EMI: CQD (1996) Analysis Date: 8/1/98 Limiting Break Size: Guillotine FQ: 2.5 FdlI: 1.65 Fuel: Vantage + SGTP (%): 10 Notes: Mixed Core - Vantage + / Performance + / RFA-2, RSG (12% SGTP)

Clad Temp (0 F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1892 1,2 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I. Vesscl Clannel DX Error -4 3 2 . IONTECF Decay leat Uncertainty Error 4 6 3 . Input Error Resulting in Incomplete Solution Mlatrix 0 7 4 . Tavg Bias Error 8 7 5 . Revised Blowdown Ileatup 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. 2005 ECCS MODEL ASSESSMENTS

1. None 0 D. OTIIER*

I . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1780

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

References:

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

August 1998.

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

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

Attachment 2 of LTR-LIS-06-117- PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Suniniary for Best Estinmate Large Break Plant Nanie: Watts Bar Unit I Cycle 8, RSG Utilitv Name: Tennessee Valley Authority Revision Date: 2/27/06 Reflood 2 5 .WAT-D-10840, "Tennessee Valley Authority, Watts Bar NuclearPlant Unit 1, Final Deliverables for 1.4% Upratc Program,'

August 31, 2000.

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

7 . VAT-D-1 1225, "10 CFR 50.46 Annual Notification and Reporting for 2003," March 2004.

8 . VAT-D-1 1334, "10 CFR 50.46 Annual Notification and Reporting for 2004," April 2005.

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

10 . NVTV-RSG-06-0 15, "LOCA & Non-LOCA Analysis Summary for Replacement Steam Generator," February 2006.

Notes:

None

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westinghousc LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Watts Bar Unit 1 Cycle 7, 0S Utility Name: Tennessee Valley Authority Revision Date: 2/23/06 Analvios Infornation EMI: NOTRUMP Analysis Date: 11/1/96 Limiting Break Size: 4 inch FQ: 2.5 FdlI: 1.65 Fuel: Vantage + SGTP (%): 10 Notes: Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (°F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1126 1,2 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . NOTRUNIP Mixture Level Tracking / Region Depletion Errors 13 4 2 . NOTRUMP Bubble Rise / Drift Flux Model Inconsistency Corrections 35 5 B. PLANNED PLANT MODIFICATION EVALUATIONS I . Annular Blankets 10 3 2 . Increased Stroke Time for the ECCS Valves 0 6 C. 2005 ECCS MODEL ASSESSMENTS I. None 0 D. OTIIER*

I Tavg Uncertainty of 6 °F 1 1,2 2 . Leaking SIS Relief Valve 120 7 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1305

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

References:

1 . WVAT-D-10337, "Tennessee Valley Authority. Watts Bar NuclearPlant, Final SafetyEvaluation to Support Teclnical Specification Changes," March 5, 1997.

2 . WAT-D-10356, "Tennessee Valley Authority, Watts Bar NuclearPlant Units 1 & 2, Final Report and Safety Evaluation for tie 10% SGTP Program," June 2, 1997.

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

4 . WAT-D-108 10, "Tennessee Valley Authority, Watts Bar Nuclear Plant Unit 1, 10 CFR 50.46 Appendix K (BART/BASHINOTRUMP) Evaluation Model Mid-Year Notification and Reporting for 2000," June 30,2000.

5 . WAT-D-11195, "10 CFR 50.46 Mid-Year Notification and Reporting for 2003," November 2003.

6 . WAT-D-1 1285, "Evaluation of Proposed Changes to the Stroke Time for the ECCS Valves," November 2004.

7 . WAT-D-11360, "Safety Injection Pump Discharge Relief Valve Leakage Evaluation," July 2005.

Notes:

None

Attachment 2 of LTR-LIS-06-117 - PCT Rackup Sheets Westinghouse LOCA Peak Clad Temperature Summary for Appendix K Small Break Plant Name: Watts Bar Unit 1 Cycle 8, RSG Utility Name: Tennessee Valley Authority Revision Date: 2/27/06 Analv dis Infortnation EM: NOTRUMP Analysis Date: 5/17/04 Limiting Break Size: 4 inch FQ: 2.5 FdlI: 1.65 Fuel: RFA-2 SGTP (%): 12 Notes: Mixed Core - Vantage + / Performance + / RFA-2 Clad Temp (0 F) Ref. Notes LICENSING BASIS Analysis-Of-Record PCT 1132 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS I . None 0 B. PLANNED PLANT MODIFICATION EVALUATIONS I . Increased Stroke Time for the ECCS Valves 0 2 C. 2005 ECCS MODEL ASSESSMENTS I. None 0 D. OTIIER*

I . Leaking SIS Relief Valve 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:

1. WVTV-RSG-06-0 15, "LOCA & Non-LOCA Analysis Sumniry for Replacement Steam Generator," Fcbruary 2006.

2 . WVAT-D-11285, "Evaluation of Proposed Changes to the Stroke Time for the ECCS Valves," November 204.

3. NVAT-D-1 1360, "Safety Injection Pump Discharge Relief Valve Leakage Evaluation," July 2005.

Notes:

None