ML12296A254

CFR 50.46 - 30-Day Special Report

ML12296A254
Person / Time
Site:	Watts Bar
Issue date:	10/18/2012
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME5186
Download: ML12296A254 (10)

Text

Tennessee Valley Authority, 1101 Market Street; Chattanooga, Tennessee 37402 October 18, 2012 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 Day Special Report

Reference:

1. TVA Letter to NRC, "10 CFR 50.46 Annual Report for Model Year 2011,"

dated April 25, 2012 [ML12117A261]

2. NRC Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation," dated December 13, 2011 [ML113430785]

3. Letter LTR-NRC-01-6, from WEC to NRC, "U. S. Nuclear Regulatory Commission, 10 CFR 50.46 Annual Notification and Reporting for 2000,"

dated March 13, 2001 [ML010800135]

4. WCAP-1 3451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting," October 1992

5. Letter LTR-NRC-12-18 from WEC to NRC, "Westinghouse Response to December 16, 2011 NRC Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (TAC No. ME5186)," dated February 17, 2012

6. Letter LTR-NRC-12-27 from WEC to NRC, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," dated March 7, 2012 [ML12072A035]

Printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 October 18, 2012

7. NRC Memorandum, "Summary of March 8, 2012, Meeting with the Pressurized Water Reactor Owners Group Regarding a Potential Thermal Conductivity Degradation Program," dated April 17, 2012 [ML12103A181]

8. NRC Memorandum, "Summary of June 27, 2012, Meeting with the Pressurized Water Reactor Owners Group Regarding the Thermal Conductivity Degradation Program," dated August 31, 2012 [ML12220A605]

The purpose of this letter is to provide a 30-day report of changes to the calculated peak cladding temperature (PCT) for the Watts Bar Nuclear Plant (WBN), Unit 1, Emergency Core Cooling System (ECCS) evaluation model. This submittal satisfies the reporting requirements for a significant change or error in accordance with 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," paragraph (a)(3)(ii). The enclosed report provides a summary of the changes to the calculated PCT for the limiting ECCS analysis.

The PCT changes identified for WBN Unit 1 in the annual report submitted on April 25, 2012 (Reference 1) when expressed as the cumulative sums of the absolute magnitudes exceed the 50 degrees Fahrenheit ('F) threshold for a significant change or error as defined in 10 CFR 50.46(a)(3)(i). Any subsequently discovered change or error will necessarily continue to be considered significant for the purposes of reporting until such time as a reanalysis of the ECCS evaluation model is completed.

Accordingly, this 30-day report is being submitted as required by 10 CFR 50.46(a)(3)(ii) to report the impact of fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown on the PCT calculation for WBN Unit 1. United States Nuclear Regulatory Commission (NRC)

Information Notice 2011-21 (Reference 2) notified addressees of the potential for this phenomenon to cause errors in realistic ECCS evaluation models. Fuel pellet TCD and peaking factor burndown were not explicitly considered in the WBN Unit 1 ECCS Large Break Loss of Coolant Accident (LBLOCA) evaluation model analysis of record (AOR).

This 30-day report also includes the impact on the PCT calculation of the change from the Westinghouse (WEC) Fuel Rod Performance and Design (PAD) computer code Version 3.4 to PAD Version 4.0 for fuel rod design inputs into the WBN Unit 1 ECCS LBLOCA evaluation model AOR. The implementation of PAD Version 4.0 into the 1996 WEC Best Estimate Large Break LOCA Evaluation Model was described in Reference 3 as a forward-fit, Discretionary Change in accordance with Section 4.1.1 of WCAP-1 3451 (Reference 4). The plant-specific upgrade from PAD Version 3.4 to PAD Version 4.0 is considered a design input change into the LBLOCA AOR for WBN Unit 1.

U.S. Nuclear Regulatory Commission Page 3 October 18, 2012 The plant-specific evaluations concerning TCD and the upgrade from PAD Version 3.4 to Version 4.0 were performed using the methods described in References 5 and 6. The analytical approach for the effects on medium to high PCT margin plants with WEC fuel, which includes WBN Unit 1, was presented to the NRC in meetings with the Pressurized Water Reactor Owners Group held on March 8 and June 27, 2012 (References 7 and 8).

As indicated in the enclosed report, there is an estimated 60°F reduction in the net licensing basis PCT for the LBLOCA. Accounting for this change, the updated (net) licensing basis PCTs for the LBLOCA and Small Break Loss of Coolant Accident (SBLOCA) evaluation models remain below the 2200°F limit. However, the cumulative sum of the absolute magnitudes of the LBLOCA PCT changes for WBN Unit 1 remains above the 50°F threshold defined for a significant change or error, thus necessitating this 30-day report.

In accordance with 10 CFR 50.46(a)(3)(ii), changes to the ECCS LBLOCA and SBLOCA evaluation models are required to be reported to the NRC within 30 days when the cumulative sum of the absolute magnitudes of the resulting PCT changes exceeds 50 0F. In addition, the licensee is required to include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with the 10 CFR 50.46 requirements.

Compliance with the 10 CFR 50.46 requirements is demonstrated by the calculated LBLOCA and SBLOCA PCTs for WBN Unit 1 remaining below the 2200°F limit. As presented in the enclosed report, the updated (net) licensing basis LBLOCA PCT is 1805°F and the updated (net) licensing basis SBLOCA PCT is 11320F. Hence, the LBLOCA and SBLOCA PCTs remain below the 2200°F limit, and compliance with the requirements of 10 CFR 50.46 is thereby demonstrated. TVA has, therefore, concluded that no proposed schedule for providing a reanalysis or other action is required.

There are no regulatory commitments in this letter. Please direct questions concerning this report to Clyde Mackaman at (423) 751-2834.

Respectfully, J.W. Shea Z

er Vice President, Nuclear Licensing Enclosure and cc: See Page 4

U.S. Nuclear Regulatory Commission Page 4 October 18, 2012

Enclosure:

10 CFR 50.46 30-Day Report for Watts Bar Nuclear Plant Unit 1 cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant

ENCLOSURE 10 CFR 50.46 30-DAY REPORT FOR WATTS BAR NUCLEAR PLANT, UNIT 1 The Watts Bar Nuclear Plant (WBN), Unit 1, reactor core currently contains only the Westinghouse Electric Company (WEC) Robust Fuel Assembly (RFA-2TM) fuel design.

Description of Changes or Errors Relative to the Previous Report TCD and Peakinq Factor Burndown and PAD 4.0 Implementation The previous 10 CFR 50.46 report for WBN Unit 1 (Reference 1) was submitted to the United States Nuclear Regulatory Commission (NRC) on April 25, 2012. As indicated in the report, WEC WCAP-14839, Revision 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Unit 1 Watts Bar Nuclear Plant" (Reference 2), is the current analysis of record (AOR) for the Best Estimate Large Break Loss of Coolant Analysis (BE LBLOCA), with a baseline Peak Cladding Temperature (PCT) of 1892 degrees Fahrenheit (°F). The indicated AOR for the Small Break Loss of Coolant Accident (SBLOCA) is WTV-RSG-06-015, "LOCA &

Non-LOCA Analysis Summary for Replacement Steam Generator for WBN Unit 1" (Reference 3), with a baseline PCT of 11 32*F.

On December 13, 2011, NRC Information Notice 2011-21 (Reference 4) notified addressees of recent information obtained concerning the impact of irradiation on fuel thermal conductivity and its potential to cause significantly higher predicted PCT results in realistic emergency core cooling system (ECCS) evaluation models. Subsequently, in a letter dated September 20, 2012, WEC notified the Tennessee Valley Authority (TVA) that the effects of fuel pellet thermal conductivity degradation (TCD) and peaking factor burndown were not explicitly considered in the Watts Bar Unit 1 BE LBLOCA AOR. The letter also informed TVA that the WEC Fuel Rod Performance and Design (PAD) computer code had been upgraded from Version 3.4 to Version 4.0. In the letter, WEC provided the estimated effect on the PCT calculation for the WBN Unit 1 BE LBLOCA AOR of fuel pellet TCD and peaking factor burndown, as well as the estimated effect of the upgrade to fuel rod design inputs for the PAD software. The plant-specific evaluations concerning TCD and the upgrade from PAD Version 3.4 to Version 4.0 were performed using the methods described in References 5 and 6. The plant specific implementation of PAD Version 4.0 into the BE LBLOCA AOR for WBN Unit 1 is considered a design input change into the BE LBLOCA analysis. The analytical approach for the effects on medium to high PCT margin plants with WEC fuel, including WBN Unit 1, was presented to the NRC in meetings with the Pressurized Water Reactor Owners Group held on March 8 and June 27, 2012 (References 7 and 8). Note that the TCD evaluations only address the effects on the WBN Unit 1 BE LBLOCA AOR. Based on a qualitative analysis, as noted in Reference 5, the impact on the SBLOCA AOR was determined to be negligible.

The enclosed Table details the accumulated PCT impact resulting from the changes or errors in the LBLOCA and SBLOCA analyses since each of the respective AORs (References 1 and 2) was established. The composite values are provided for the effects on the LBLOCA analysis because they represent the worst case PCT impact from either the Reflood 1 or the Reflood 2 accident scenario sequence. As indicated in the Table, the change to the LBLOCA limiting composite PCT has decreased by 60*F due to the combined effects of the upgrade to PAD Version 4.0 (75°F decrease in PCT) and the TCD and peaking factor burndowh (15°F increase).

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ENCLOSURE 10 CFR 50.46 30-DAY REPORT FOR WATTS BAR NUCLEAR PLANT, UNIT 1 In accordance with 10 CFR 50.46(a)(3)(i), this change is considered significant because the absolute magnitudes of the accumulated changes and errors to the calculated PCT since the baseline BE LBLOCA AOR analysis (Reference 1) was performed exceeds 500F.

Vessel Channel DX Error In 1999, WEC made TVA aware of an error in the ECCS model which they referred to as "Vessel Channel DX Error (Including Investigation of Code Uncertainties)." This particular error involved the use of incorrect cell height in the calculating gap flow wall friction and interfacial drag coefficients.

This error has no impact on the SBLOCA PCT since the error pre-dates the current (2006)

AOR. The impact of this error on LBLOCA PCT has been previously reported as 4°F decrease in the PCT. TVA recently identified that a discussion of this error was not included in the 10 CFR 50.46 report for 1999 (Reference 9) or in a subsequent report. TVA has elected to include a discussion of the error in this report for documentation completeness.

ECCS Valve Stroke Time Changes In 2004, TVA asked WEC to evaluate the impact of increasing the stroke time for several ECCS valves on the AOR. Westinghouse determined that the only impact of the stroke time increase was on the LOCA AOR.

For the LBLOCA and SBLOCA AOR, WEC determined that the effect of this change was negligible. TVA recently identified that a discussion of this change was not included in the 10 CFR 50.46 report for 2004 (Reference 14) or in a subsequent report. TVA has elected to include a discussion of the change in this report for documentation completeness.

WBN Unit 1 RSG Upgrade In 2006, WEC revised the BE LBLOCA AOR to incorporate the changes associated with the replacement steam generators (RSGs) for WBN Unit 1. The RSG project upgraded the WBN Unit 1 steam generators (SGs) from the WEC Model D3 SGs to the WEC Model 68AXP SGs.

The WEC evaluation incorporated the RSG design into the model and included two transient calculations, one at 0 percent SG tube plugging and one at 12 percent.

The impact of the upgrade to the RSGs has been previously reported as a 1 0°F decrease in the LBLOCA AOR PCT. TVA recently identified that a discussion of this change was not included in the 10 CFR 50.46 report for 2006 (Reference 15) or in a subsequent report. WVA has elected to include a discussion of the change in this report for documentation completeness.

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

Summary of Changes to WBN Unit I PCT for LBLOCA and SBLOCA Year Description LBLOCA LBLOCA SBLOCA SBLOCA Reference APCT IAPCTI APCT IAPCTI

__F)

OF (OF)

(OF) 1998 BE LBLOCA AOR Baseline PCT 1892 2

2006 SBLOCA AOR Baseline PCT 1132 3

1999 Vessel Channel DX Error

-4 4

10*

2000 Increased Accumulator Room 4

4 10 Temperature Evaluation 2000 1.4% Uprate Evaluation 12 12 10 2000 Accumulator Line/Pressurizer

-131 131 10 Surge Line Data Evaluation 2000 MONTECF Decay Heat 4

4 11 Uncertainty Error 2001 WBN Specific LBLOCA Vessel 0

0 12 Geometry Input Errors 2003 Input Error Resulting in 0

0 13 Incomplete Solution Matrix 2003 Tavg Bias Error 8

8 13 2004 Increased Stroke Time for ECCS 0

0 14" Valves 2004 Revised Blowdown Heatup 5

5 14 Uncertainty Distribution 2006 Replacement Steam Generators

-10 10 0

0 15*

(D3 to 68AXP) 2006 HOTSPOT Fuel Relocation Error 65 65 0

0 15 E-3 of 6

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

Summary of Changes to WBN Unit I PCT for LBLOCA and SBLOCA Year Description LBLOCA LBLOCA SBLOCA SBLOCA Reference APCT IAPCTI APCT IAPCTI (OF)

(OF)

(F)

(OF) 2009 PMID Violation Evaluation 20 20 0

0 16 2012 PAD 4.0 Implementation

-75 75 2012 TCD and Peaking Factor 15 15 Burndown Updated (net) licensing basis P--

1805 1132 AOR PCT + I APCT Cumulative sum of PCT 353 0

changes I I APCTI

• Discussed in text of this enclosure E-4 of 6

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

1.

Letter from TVA to NRC, "10 CFR 50.46 Annual Report for Model Year 2011," dated April 25, 2012 [ML12117A261]

2.

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

3.

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

4.

NRC Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation", dated December 13, 2011 [ML113430785]

5.

Letter LTR-NRC-12-18 from WEC to NRC, "Westinghouse Response to December 16, 2011 NRC Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (TAC No.

ME5186)," dated February 17, 2012

6.

Letter LTR-NRC-12-27 from WEC to NRC, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)," dated March 7, 2012 [ML12072A035]

7.

NRC Memorandum, "Summary of March 8, 2012, Meeting with the Pressurized Water Reactor Owners Group Regarding a Potential Thermal Conductivity Degradation Program," dated April 17, 2012 [ML12103A181]

8.

NRC Memorandum, "Summary of June 27, 2012, Meeting with the Pressurized Water Reactor Owners Group Regarding the Thermal Conductivity Degradation Program,"

dated August 31, 2012 [ML12220A605]

9.

Letter from TVA to NRC, "Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - 30 Day Report," dated October 18, 1999 [ML073240684]

10.

Letter from TVA to NRC, "Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - 30 Day Report and Annual Notification and Reporting for 2000," dated October 26, 2000 [ML003764646]

11.

Letter from TVA to NRC, "Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - 30 Day Report and Revised Annual Notification Report for 2000," dated September 7, 2001 [ML012570290]

12.

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

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ENCLOSURE 10 CFR 50.46 30-DAY REPORT FOR WATTS BAR NUCLEAR PLANT, UNIT I

13.

Letter from TVA to NRC, "Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - 30 Day Report and Revised Annual Notification and Reporting for 2003," dated April 19, 2004 [ML041130196]

14.

Letter from TVA to NRC, "Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - 30 Day Annual Notification and Reporting for 2004," dated April 19, 2005 [ML051120164]

15.

Letter from TVA to NRC, "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 [ML071860388]

16.

Letter from TVA to NRC, "Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - Annual Notification and Reporting," dated July 2, 2008 [ML081980033]

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