CNL-15-053, Combined 10 CFR 50.46 - 30-day and Annual Report of Changes and Errors to the Calculated Peak Cladding Temperature (PCT) for the Emergency Cooling Core System (ECCS) Evaluation Model

Combined 10 CFR 50.46 - 30-day and Annual Report of Changes and Errors to the Calculated Peak Cladding Temperature (PCT) for the Emergency Cooling Core System (ECCS) Evaluation Model

ML15098A124
Person / Time
Site:	Watts Bar
Issue date:	03/30/2015
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNL-15-053 L44 150330 001
Download: ML15098A124 (14)

Text

L44 150330 001 Tennessee Valley Authority, 1101 Market Street, Chattanooga , TN 37402 CNL-15-053 March 30, 2015 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 NPF-90 NRC Docket No. 50-390 Watts Bar Nuclear Plant, Unit 2 Construction Permit No. CPPR-92 NRC Docket Nos. 50-391

Subject:

10 CFR 50.46- 30-Day and Annual Report for Watts Bar, Units 1 and 2

Reference:

1. TVA Letter to NRC, "1 0 CFR 50.46- Annual Report for Watts Bar, Lin its 1 and 2," dated April25, 2014 [ML14119A332]

2. TVA letter to NRC, "1 0 CFR 50.46- 30-Day report for Watts Bar, Unit 2,"

dated February 6, 2015 [ML15037A725]

The purpose of this letter is to provide a combined 30-day report and annual report of changes and errors to the calculated peak cladding temperature (PCT) for the Watts Bar Nuclear Plant (WBN), Units 1 and 2, Emergency Core Cooling System (ECCS) evaluation model. This report is required in accordance with Title 10 of the Code of Federal Regulations (1 0 CFR) 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, " paragraph (a)(3)(ii) .

The last 10 CFR 50.46 annual report for WBN Units 1 and 2 was submitted to the United States Nuclear Regulatory Commission (NRC) on April 25, 2014 (Reference 1). The last annual report indicated that in accordance with 10 CFR 50.46(a)(3)(ii), subsequent changes or errors affecting the WBN Unit 1 and 2 large break loss of coolant accident (LBLOCA) analysis are considered significant for reporting purposes because the absolute magnitude of the accumulated changes in PCT since the last LBLOCA analysis exceeds 50 degrees Fahrenheit.

On February 6, 2015, Tennessee Valley Authority (TVA) provided a 30-Day report of the estimated effect of a change in accumulator injection line resistances for the WBN Unit 2 small break loss-of-coolant accident (SBLOCA) analysis and the LBLOCA analysis (Reference 2).

The peak clad temperature (PCT) was unchanged .

U.S. Nuclear Regulatory Commission CNL-15-053 Page 2 March 30, 2015 As indicated in the enclosure, the current updated (net) licensing basis PCT for WBN Unit 1, LBLOCA remains unchanged at 1865 degrees Fahrenheit and the updated (net) licensing basis PCT for the SBLOCA remains unchanged at 1132 degrees Fahrenheit. Similarly, the current updated (net) licensing basis PCT for the WBN Unit 2, LBLOCA remains unchanged at 1711 degrees Fahrenheit and the updated (net) licensing basis PCT for the SBLOCA remains at 1184 degrees Fahrenheit.

There are no regulatory commitments in this letter. Please direct questions concerning this report to Gordon Arent at (423) 365-2004.

ly, 0~

J.

Enclosure:

Watts Bar Nuclear Plant, 10 CFR 50.46 Annual Report cc (Enclosures):

NRC Regional Administrator- Region II NRC Senior Resident Inspector- Watts Bar Nuclear Plant Unit 2

ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 In accordance with the reporting requirements of Title 10 of the Code of Federal Regulations (1 0 CFR) 50.46(a)(3)(ii) , the following is a summary of the limiting design basis loss of coolant accident (LOCA) analysis results established using the current Watts Bar Nuclear Plant (WBN)

Emergency Core Cooling System evaluation models for Units 1 and 2. This report describes the changes and errors affecting the calculated peak cladding temperatures (PCTs) since the last analysis of record was submitted to the Nuclear Regulatory Commission (NRC) .

The last 10 CFR 50.46 annual report for WBN Units 1 and 2 was submitted to the NRC on April 25, 2014 (Reference 1). A 10 CFR 50.46 30-Day Report for Watts Bar, Unit 2 was submitted to the NRC on February 6, 2015 (reference 2) .

Table 1 lists the changes and errors in the large break loss of coolant accident (LBLOCA) analysis for Unit 1 since the analysis of record (AOR) and the associated effect on PCT. Table 2 lists the changes and errors in the small break loss of coolant accident (SBLOCA) analyses for WBN Unit 1 since the AOR and the associated effect on PCT. The changes that were not previously identified in reference 1 are described in the notes to Tables 1 and 2.

Table 3 lists the changes and errors in the LBLOCA and SBLOCA analyses for WBN Unit 2 since each of the respective AORs and the associated effect on PCT. The changes that were not previo!Jsly identified in reference 1 and 2 are described in the notes to Table 3.

The updated (net) licensing basis PCT for the LBLOCA and SBLOCA remain unchanged for WBN Units 1 and 2 from the last annual report (Reference 1).

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ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 Table 1 Watts Bar Unit 1 LBLOCA Reflood 1 Reflood 2

~PCT I~PCTI ~PCT I~PCTI Note Reference Year Description (oF) (oF) (oF) (oF) 1998 BE LBLOCA AOR PCT 1656 - -- 1892 - -- - -- 3 1999 Vessel Channel OX Error 56 56 -4 4 - -- 4 2000 Increased Accumulator Room 4 4 4 4 - -- 4 Temperature Evaluation 2000 1.4% Uprate Evaluation 12 12 12 12 - -- 4 2000 Accumulator Line/Pressurizer

-37 37 -131 131 --- 4 Sucge Line Data Evaluation 2000 MONTECF Decay Heat 4 4 4 4 --- 5 Uncertainty Error 2001 WBN Specific LBLOCA 0 0 0 0 --- 6 Vessel Geometry Input Errors 2003 Input Error Resulting in 60 60 0 0 - -- 7 Incomplete Solution Matrix 2003 Ta'{g_ Bias Error 8 8 8 8 --- 7 2004 Increased Stroke Time for 0 0 0 0 - -- 8 ECCS Valves 2004 Revised Slowdown Heatup 5 5 5 5 - -- 8 Uncertainty Distribution 2006 Replacement Steam

-50 50 -10 10 --- 9 Generators (03 to 68AXP) 2006 HOTSPOT' Fuel Relocation 0 0 65 65 --- 9 Error CNL-15-053 E 2 of 12

ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 Table 1 Watts Bar Unit 1 LBLOCA Reflood 1 Reflood 2

!lPCT I!lPCTI !lPCT lllPCTI Note Reference Year Description (oF) (oF) (oF) (oF) 2012 PMID/PBOT Violation 20 20 20 20 --- 10, 11 Evaluation 2012 TCD and Peaking Factor 114 114 15 15 - -- 10, 11 Burndown 2013 WCOBRA!TRAC TM History 0 0 0 0 12 File Dimension Error 2013 General Code Maintenance 0 0 0 0 --- 13 2013 HOTSPOT' Burst Temperature Calculation for 0 0 0 0 - -- 13 Zl RLO TM Cladding 2013 HOTSPOTTM Iteration Algorithm for Calculation 0 0 0 0 - -- 13 Initial Fuel Pellet Average Temperature 2013 WCOBRA!TRAC TM Automated Restart Process 0 0 0 0 --- 13 Loqic Error 2013 Rod Internal Pressure 0 0 0 0 --- 13 Calculation Error 2013 Elevations for Heat Slab 0 0 0 0 - -- 14 Temperature Initialization 2013 Heat Transfer Model Error 0 0 0 0 14 Corrections 2013 Correction to Heat Transfer

--- 14 0 0 0 0 Node Initialization 2013 Mass Conservation Error Fix 0 0 0 0 --- 14 2013 Correction to Split Channel 0 0 0 0 --- 14 Momentum Equation CNL-15-053 E 3 of 12

ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 Table 1 Watts Bar Unit 1 LBLOCA Reflood 1 Reflood 2 L1PCT I L1PCTI L1PCT I L1PCTI Note Reference Year Description (oF) (oF) (oF) (oF) 2013 Heat Transfer Logic Correction for Rod Burst 0 0 0 0 - -- 14 Calculation 2013 Changes to Vessel Superheated Steam 0 0 0 0 - -- 14 Properties 2013 Update to Metal Density Reference Temperatures 0 0 0 0 - -- 14 2013 Decay Heat Model Error 0 0 0 0 14 Corrections 2013 Correction to the Pipe Exit 0 0 0 0 14 Pressure Drop Error 2013 WCOBRAITRAC File

- -- 14 0 0 0 0 Dimension Error Correction 2013 Revised Heat Transfer

-40 40 -85 85 14 Multiplier Distributions 2013 HOTSPOT Burst Strain Error 20 20 70 70 --- 15 General Computer Code 2014 Maintenance 0 0 0 0 1 ---

Revised Uncertainty in 2014 LBLOCA Monte Carl 0 0 0 0 2 - --

Simulations Updated (net) licensing basis PCT

- -- 1832 - -- 1865 -- -

AOR PCT + L L1PCT Cumulative sum of PCT changes - -- 430 - -- 433 L I L1PCTI CNL-15-053 E 4 of 12

ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 Notes:

1. Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding; and eliminating inactive coding.

2. As part of the WCOBRAITRAC validation basis for use in Best-Estimate (BE) Large-Break Loss-of-Coolant Accident (LBLOCA) analysis, simulations of many separate effects tests (SETs) and integral effects tests (lETs) were performed . In the Westinghouse 1996 BE LBLOCA evaluation model (CQD evaluation model), the simulations of the test data which included core heat transfer measurements were used in the uncertainty methodology in two ways.

a. To develop heat transfer multiplier distributions which are used in the HOTSPOT code for local hot rod calculations.

b. To determine the minimum code uncertainties for blowdown and reflood that are used in the MONTECF code for the Monte Carlo uncertainty calculations.

The effects of several changes and error corrections to the WCOBRAITRAC code were previously reported in 2013, including effects of revised heat transfer multiplier distributions.

As a result of previously reported changes and error corrections made to WCOBRAITRAC , the total code uncertainty values utilized in the MONTECF code for both blowdown and reflood were recalculated. The recalculated total code uncertainty values are lower than the previous values.

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ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 Table 2 Watts Bar Unit 1 SBLOCA SBLOCA SBLOCA Year Description t\PCT lt1PCTI Note Reference (oF) (oF) 2006 SBLOCA AOR PCT 1132 --- - -- 16 NOTRUMP-EM 1M 2013 Evaluation of Fuel 0 0 - -- 13 Pellet TCD General Computer 0 0 1 - --

2014 Code Maintenance Fuel Rod Gap 0 0 2 - --

2014 Conductance Error Radiation Heat 0 0 3 ---

2014 Transfer Model Error SBLOCTA Pre-DNB Cladding Surface 2014 Heat Transfer 0 0 4 - --

Coefficient Calculation Evaluation of 8-inch 0 0 5 -- -

2014 Axial Blankets Updated (net) licensing basis

--- PCT 1132 --- - -- ---

AOR PCT + l: t\PCT Cumulative sum of PCT changes - -- 0 - -- - --

L I L\PCTI Notes:

1. Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks; enhancing the code output; optimizing active coding ; and eliminating inactive coding .

2. An incorrect temperature was used in the calculation of the cladding emissivity term in the fuel rod gap conductance model in the NOTRUMP computer code (reactor coolant system response model)

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ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014

3. Two errors were discovered in the calculation of the radiation heat transfer coefficient within the fuel rod model of the NOTRUMP computer code (reactor coolant system response model). First, existing logic did not preclude non-physical negative or large (negative or positive) radiation heat transfer coefficients from being calculated. These erroneous calculations occurred when the vapor temperature exceeded the cladding surface temperature or when the predicted temperature difference was less than 1°F.

Second, a temperature term incorrectly used degrees Fahrenheit instead of Rankine.

4. Two errors were discovered in the pre-departure from nucleate boiling (pre-ON B) cladding surface heat transfer coefficient calculation in the SBLOCTA code (cladding heat-up calculations). The first error is a result of inconsistent time units (hours vs. seconds) in the parameters used for the calculation of the Reynolds and Prandtl numbers, and the second error is an incorrect diameter used to develop the area term in the cladding surface heat flux calculation .

5. It was identified that the Watts Bar Unit 1 small-break loss-of-coolant accident (SBLOCA) analysis of record (AOR) explicitly analyzed 6-inch axial blankets at the top and bottom of the active fuel rod; however, the plant operates with 8-inch axial blankets.

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ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 Table 3 Watts Bar Unit 2 LBLOCA and SBLOCA Year Description LBLOCA LBLOCA SBLOCA SBLOCA Note Reference

~PCT I~PCTI ~PCT I~PCTI (oF) (oF) (oF) (oF) 2013 LBLOCA AOR PCT 1766 -- --- - -- 17 2010 SBLOCA AOR PCT --- -- - 1184 - -- --- 18 2013 Elevations for Heat Slab 19 0 0 --- --- - --

Temperature Initialization 2013 Heat Transfer Model Error - -- 19 0 0 --- - --

Corrections 2013 Correction to Heat Transfer 19 0 0 - -- - -- ---

Node Initialization 2013 Mass Conservation Error Fix 0 0 --- - -- --- 19 2013 Correction to Split Channel - -- 19 0 0 - -- - --

Momentum Equation 2013 Heat Transfer Logic 19 Correction for Rod Burst 0 0 --- --- ---

Calculation 2013 Changes to Vessel 19 Superheated Steam 0 0 --- --- ---

Properties 2013 Update to Metal Density 19 0 0 - -- - -- ---

Reference Temperatures 2013 Decay Heat Model Error - -- 19 0 0 --- - --

Corrections 2013 Correction to the Pipe Exit --- 19 0 0 - -- ---

Pressure Drop Error 2013 WCOBRAITRAC File - -- 19 0 0 - -- - --

Dimension Error Correction 2013 Revised Heat Transfer 19

-55 55 --- - -- - --

Multiplier Distributions 2013 Initial Fuel Pellet Average 0 0 19 Temperature Uncertainty Calculation - -- - -- - --

2013 HOTSPOT Burst Strain Error 0 0 - -- --- - -- 15 CNL-15-053 E 8 of 12

ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014 Year Description LBLOCA LBLOCA SBLOCA SBLOCA Note Reference i1PCT IL1PCTI 1'1PCT I 1'1PCTI (oF) (oF) (oF) (oF)

Cold Leg Accumulator 2014 Injection Lines Hydraulic 0 0 0 0 - -- 2 Resistance Changes General Computer Code 2014 Maintenance 0 0 0 0 1 ---

Errors in Decay Group 2014 Uncertainty Factors 0 0 - -- - -- 2 - --

Treatment of Burnup Effects 2014 on Thermal Conductivity 0 0 --- --- 3 ---

Degradation Fuel Rod Gap Conductance 2014 Error --- --- 0 0 4 - --

Radiation Heat Transfer 2014 Model Error - -- - -- 0 0 5 - --

SBLOCTA Pre-DNB Cladding Surface Heat 2014 Transfer Coefficient --- - -- 0 0 6 - --

Calculation Updated (net) licensing basis PCT (AOR PCT + L 1'1PCT) 1711 1184 Cumulative sum of PCT changes

- -- - -- 55 0 0 - -- ---

L I 1'1PCTI Notes:

1. Various changes have been made to enhance the usability of codes and to streamline future analyses. Examples of these changes include modifying input variable definitions, units and defaults; improving the input diagnostic checks ; enhancing the code output; optimizing active coding ; and eliminating inactive coding .

2. Errors in the calculation of decay heat were discovered in the WCOBRA!TRAC code. The decay group uncertainty factors for each fissile isotope are provided in Table 8-14 of WCAP-16009-P-A. The uncertainty factors for 239Pu were applied to 238U , and those for 238U were applied to 239Pu. This error causes an over-prediction of the uncertainty in decay power from 239Pu and an under-prediction of the uncertainty in decay power from 238U. Further, the decay group uncertainty factor for Decay Group 6 of 235U was erroneously coded as 2.5% instead of 2.25%.

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ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014

3. It was discovered that some approximations made in the Watts Bar Unit 2 plant specific adaptation of the ASTRUM methodology to address TCD could under-estimate rod power and local peaking in second-cycle assemblies. Also , the analysis assumes that the modeled hot assembly is surrounded by assemblies with power consistent with the core-average. However, second-cycle assemblies are often face-adjacent with feed assemblies, which can have power higher than the core-average. This can result in an over-estimation of the gamma energy deposited away from the hot rod and hot assembly.

4. An incorrect temperature was used in the calculation of the cladding emissivity term in the fuel rod gap conductance model in the NOTRUMP computer code (reactor coolant system response model)

5. Two errors were discovered in the calculation of the radiation heat transfer coefficient within the fuel rod model of the NOTRUMP computer code (reactor coolant system response model). First, existing logic did not preclude non-physical negative or large (negative or positive) radiation heat transfer coefficients from being calculated . These erroneous calculations occurred when the vapor temperature exceeded the cladding surface temperature or when the predicted temperature difference was less than 1°F.

Second , a temperature term incorrectly used degrees Fahrenheit instead of Rankine.

6. Two errors were discovered in the pre-departure from nucleate boiling (pre-DNB) cladding surface heat transfer coefficient calculation in the SBLOCTA code (cladding heat-up calculations) . The first error is a result of inconsistent time units (hours vs. seconds) in the parameters used for the calculation of the Reynolds and Prandtl numbers, and the second error is an incorrect diameter used to develop the area term in the cladding surface heat flux calculation .

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ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014

References:

1. Letter from TVA to NRC, 10 CFR 50.46 Day and Annual Report for 2013, dated April 25, 2014 [ML14119A332]

2. Letter from TVA to NRC, 10 CFR 50.46 Day Report for Watts Bar, Unit 2 dated February 6, 2015 [ML15037A725]

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

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

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

6. 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]

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

8. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes - Annual Notification and Reporting for2004, dated April 19, 2005 [ML051120164]

9. Letter from TVA to NRC, Watts Bar Nuclear Plant (WBN) Unit 1 - Emergency Core Cooling System (ECCS) Evaluation Model Changes Day Report and Annual Notification and Reporting for 2006, dated July 3, 2007 [ML071860388]

10. Letter from TVA to NRC, Supplement to 10 CFR 50.46 Day Special Report, dated February 13, 2013 [ML13046A002]

11. Letter from TVA to NRC, 10 CFR 50.46 Day Special Report, dated October 18, 2012

[ML12296A254]

12. Letter from TVA to NRC, 10 CFR 50.46 Day Report for Watts Bar Unit 1, dated March 19, 2013 [ML13080A405]

13. Letter from TVA to NRC, 10 CFR 50.46 Day and Annual Report for 2012, dated April 25, 2013 [ML13120A005]

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ENCLOSURE Watts Bar Nuclear Plant 10 CFR 50.46 30-Day Report and Annual Report for 2014

14. Letter from TVA to NRC , "10 CFR 50.46- 30 day Report for Watts Bar, Unit 1," dated August 28, 2013 [ML13267A034]

15. Letter from TVA to NRC, "10 CFR 50.46-30 day Report for Watts Bar, Units 1 and 2," dated February 28, 2014. [ML14064A431]

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

17. WCAP-17093-P , Revision 1, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Unit 2 Nuclear Plant Using the ASTRUM Methodology," June 2013

18. WBT-D-1460, "Final Small Break LOCA Summary Report for WBN Unit 2," January 2010

19. Letter from TVA to NRC, "Watts Bar Nuclear Plant, Unit 2- Emergency Core Cooling System Evaluation Model Changes- 30 Day Report- 10 CFR 50.46 Notification ," dated August 28, 2013.[ML13246A076]

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