ML12088A104

Response to Information Request Pursuant to 10 CFR 50-.54 (F) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse-Furnished.

ML12088A104
Person / Time
Site:	Cook
Issue date:	03/19/2012
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC M99899
Download: ML12088A104 (16)

Text

z INDIANA MICHIGAN Indiana Michigan Power POVERO One Cook Place Bridgman, MI 49106 A unit of American Electric Power IndianaMichiganPower.com March 19, 2012 AEP-NRC-2012-13 10 CFR 50.54(f) 10 CFR 50.46 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 RESPONSE TO INFORMATION REQUEST PURSUANT TO 10 CFR 50.54(f) RELATED TO THE ESTIMATED EFFECT ON PEAK CLADDING TEMPERATURE RESULTING FROM THERMAL CONDUCTIVITY DEGRADATION IN THE WESTINGHOUSE-FURNISHED REALISTIC EMERGENCY CORE COOLING SYSTEM EVALUATION (TAC .NO. M99899)

References

1. Letter from M. G. Evans, U. S. Nuclear Regulatory Commission to L. J. Weber, Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Information Request Pursuant To 10 CFR 50.54(f) Related To The Estimated Effect On Peak Cladding Temperature Resulting From Thermal Conductivity Degradation In The Westinghouse-Furnished Realistic Emergency Core Cooling System Evaluation (TAC NO. M99899)," dated February 16, 2012, Agencywide Documents Access and Management System (ADAMS)

Accession No. ML12041A384

2. Letter from J. A. Gresham, Westinghouse Electric Company (WEC), to the U.S. Nuclear Regulatory Commission, "Westinghouse Input Supporting Licensee Response to NRC .10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)", LTR-NRC-12-27, dated March 7, 2012

Dear Sir or Madam:

By Reference 1, the Nuclear Regulatory Commission (NRC) requested information pursuant to 10 CFR 50.54(f) from Indiana Michigan Power Company (A&M), licensee for the Donald C. Cook Nuclear Plant (CNP) Units 1 and 2. Information was requested regarding the effect of a potentially significant error, as defined in 10 CFR 50.46(a)(3)(i), associated with thermal conductivity degradation (TCD), on peak cladding temperature (PCT) in the Westinghouse Electric Company (WEC)-furnished realistic emergency core cooling system (ECCS) evaluation models. Acx)q M(ZAz

U. S. Nuclear Regulatory Commission AEP-NRC-2012-13 Page 2 is an affirmation. Enclosure 2 contains I&M's 30-day response for the requested information pursuant to the 10 CFR 50.54(f) letter for CNP. By Reference 2, WEC submitted a description of the methodology and assumptions used to determine the estimated PCT impact due to TCD directly to the NRC in support of I&M's response to address issue 2 in the request for information.

The estimated impact on the CNP Large Break LOCA (LBLOCA) Evaluation Model from fuel TCD represents a significant change in PCT, as defined in 10 CFR 50.46(a)(3)(i). 10 CFR 50.46(a)(3)(ii) requires the licensee to provide a report within 30 days, including a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with 10 CFR 50.46. The 50.46 report and proposed reanalysis schedule is provided as Enclosure 3 to this letter.

A new regulatory commitment is provided as Enclosure 4 to this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President MCS/jr

Enclosures:

1. Affirmation

2. Response to NRC Information Request Pursuant to 50.54(F) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse Furnished Realistic Emergency Core Cooling System Evaluation

3. 10 CFR 50.46 Report and Proposed Schedule for Reanalysis

4. Commitment

U. S. Nuclear Regulatory Commission AEP-NRC-2012-13 Page 3 c: E. J. Leeds, NRR J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosures MDEQ - WHMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III P. S. Tam, NRC Washington, DC to AEP-NRC-2012-13 AFFIRMATION AFFIRMATION I, Joel P. Gebbie, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.

Indiana Michigan Power Company Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS DAYOF ,2012 Notary Publir/'

My Commission Expires / - a I - &0)

Enclosure 2 to AEP-NRC-2012-13 Response to NRC Information Request Pursuant to 50.54(F) Related to the Estimated Effect on Peak Cladding Temperature Resulting from Thermal Conductivity Degradation in the Westinghouse Furnished Realistic Emergency Core Cooling System Evaluation to AEP-NRC-2012-13 Page 1 of 6 By Reference 1, Indiana Michigan Power Company (I&M), licensee for the Donald C. Cook Nuclear Plant (CNP) Units 1 and 2 was required to provide information regarding the effect of a potentially significant error, as defined in 10 CFR 50.46(a)(3)(i), associated with thermal conductivity degradation (TCD), on peak cladding temperature (PCT) in the Westinghouse Electric Company (WEC)-furnished realistic emergency core cooling system (ECCS) evaluation models.

Items to be specifically addressed:

(1) An estimation of the effect of the thermal conductivity degradation error on the peak fuel cladding temperature calculation for the emergency core cooling system evaluations at Donald C. Cook Nuclear Plant, Units 1 and 2.

(2) A description of the methodology and assumptions used to determine the estimates.

This description shall include consideration of experimental data relevant to thermal conductivity degradation and specific information regarding any computer code model changes which were necessary to address these data.

I&M Response The Nuclear Regulatory Commission (NRC) approved 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM (Reference 2) is based on the PAD 4.0 fuel performance code (Reference 3). PAD 4.0 was licensed without explicitly considering fuel pellet TCD with burnup. Explicit modeling of fuel pellet TCD in the fuel performance code leads to changes in the fuel rod design parameters beyond beginning-of life which are input to the large-break LOCA (LBLOCA) analyses. The effects of explicitly modeling fuel pellet TCD on the CNP Units 1 and 2 LBLOCA analyses have been considered as described in Reference 4.

Fuel performance data that accounts for fuel pellet TCD was used as input to the CNP Units 1 and 2 evaluations. The new PAD fuel performance data was generated with a representative model (using an unlicensed model) that includes explicit modeling of fuel pellet TCD as described in Reference 4. Therefore the evaluations performed consider the fuel pellet TCD effects cited in NRC Information Notice 2011-21 (Reference 5).

Quantitative evaluations, as discussed in Reference 4, were performed to assess the PCT effect of TCD and peaking factor burndown with other considerations of burnup on the CNP large-break Loss-of Coolant Accident (LBLOCA) analyses and concluded that the estimated PCT impact is 384°F for Unit 1 and 73°F for Unit 2. The peaking factor burndown included in the evaluations is provided in Tables 1 and 2. I&M and WEC utilized processes which ensure that the LBLOCA analysis input values conservatively bound the as-operated plant values for those parameters.

to AEP-NRC-2012-13 Page 2 of 6 Table 1 CNP Unit 1 Peaking Factors Versus Rod Burnup Rod Burnup FQ Steady-State FQ Transient(I)

(MWD/MTU) FDH(1),(2) 0 1.70 2.15 1.545 28,000 1.70 2.15 1.545 30,000 1.50 1.90 1.404 60,000 1.40 1.77 1.28 62,000 1.40 1.77 1.28 (1) Includes uncertainties.

(2) Hot assembly radial peaking factor uses same peaking factor burndown, since it is a function of FDH.

Table 2 CNP Unit 2 Peaking Factors Versus Rod Burnup Rod Burnup FQ Steady-State FQ Transient(I) FDH(1),(2)

(MWD/MTU) 0 1.85 2.335 1.61 30,000 1.85 2.335 1.61 60,000 1.40 1.767 1.3 62,000 1.40 1.767 1.3 (1) Includes uncertainties.

(2) Hot assembly radial peaking factor uses same peaking factor burndown, since it is a function of FDH.

to AEP-NRC-2012-13 Page 3 of 6 To demonstrate compliance with the 10 CFR 50.46(b)(1) acceptance criterion concerning PCT when explicitly considering fuel pellet TCD and peaking factor burndown in the CNP Units 1 and 2 LBLOCA analyses, design input values were revised to more closely represent current plant operation. These input changes are not changes to the approved 2004 Westinghouse Realistic Large Break LOCA Evaluation Model Using ASTRUM.

CNP Units 1 and 2 Updated Design Inputs for Best-Estimate Large-Break LOCA Evaluations Considering.the Effects of Fuel Thermal Conductivity Degradation Design Input Updated Value AOR Value SG Tube Plugging Unit 1 <2% Unit 1 < 10%

Unit 2:< 1% Unit 2:< 10%

Peaking Factors See Tables 1 and 2 See COLR(1)

Hot Full Power Nominal Tavg(2) Unit 1 = 556 0 F Unit 1 = 575.4 0 F Unit 2 = 574°F Unit 2 = 578.1°F Accumulator Temperature Unit 1 70°F < TACC < 100°F Unit 1 60°F < TACC < 120°F (TAcc) Unit 2 60°F < TAcc < 115°F Unit 2 60°F < TACC < 120°F Safety Injection Flow(3) Minimum Minimum Safety Injection Temperature 70 0 F:* Ts1 5 100OF 70 0 F

• Ts1:5 105 0 F (Ts,) 70F<T,__0°_7_F<Ts__0 _

Safety Injection Initiation Delay < 17 sec (with offsite power) <27 sec (with offsite power)

Time (4) < 28 sec (without offsite < 54 sec (without offsite power) power)

Unit 1 see License Amendment Request Figure Unit 1 see Figure 1 17 [Reference 6]

Containment Pressure Unit 2 see License Unit 2 unchanged Amendment Request Figure 17 [Reference 7] and Reference 8 Notes:

(1) Refer to Unit 1 Cycle 24 and Unit 2 Cycle 19 COLR.

(2) Uncertainty band applied to nominal value remains unchanged from AOR.

(3) Updated based on LBLOCA representative minimum RWST level and containment spilling assumption.

(4) Updated based on current surveillance testing requirements to AEP-NRC-2012-13 Page 4 of 6 Figure 1 Analyzed Versus Calculated Containment Backpressure WCOBRA/TRAC Contai nment Backpressure LOTIC2 Calculated Containment Backpressure 26 24-f 22-C-,

220

( 8- .... . . .. . . . . . . . . .. . . .

c.J)

C!)

18 0 100 200 300 400 500 finme (s)

I&M and WEC utilize processes which ensure that LOCA analysis input values conservatively bound the as-operated plant values for those parameters.

The evaluation method discussed in Reference 4 was used to determine the estimated effect of fuel pellet TCD and peaking factor burndown. First, the integrated PCT was calculated to demonstrate compliance with the 10 CFR 50.46(b)(1) criterion when the design input changes and TCD and peaking factor burndown were considered. Then, the margin PCT was calculated, including only the design input changes.

For the integrated PCT calculation, WEC performed a total of 19 WCOBRA/TRAC executions for Unit 1, and a total of 25 WCOBRAITRAC executions for Unit 2. The uncertainty attributes of these executions (hereafter referred to as "cases") were taken from among the most limiting cases from the original 124-run ASTRUM analyses. The evaluations considered an adequate range of burnup such that the effects of TCD and related burnup effects were captured.

HOTSPOT executions were performed for each WCOBRAITRAC case to consider the effect of local uncertainties for both IFBA (Integral Fuel Burnable Absorber) and non-IFBA fuel.

For the margin PCT calculation, WCOBRA/TRAC cases were executed until an estimated trend could be established. Again, the uncertainty attributes were taken from among the most limiting cases from the original 124-run ASTRUM analyses.

to AEP-NRC-2012-13 Page 5 of 6 Consistent with the ASTRUM methodology, the most limiting PCT from each evaluation was taken as the representative PCT. The limiting integrated PCT case, considering all design input changes and TCD and peaking factor burndown, was 2131'F for Unit 1 and 1941°F for Unit 2, which is less than the 2200°F acceptance criterion. Considering only the design input changes, the margin PCT was 1747 0 F for Unit 1 and 1868°F for Unit 2.

Given the current analyses of record (AOR) PCT (2128°F(U1) and 2107°F(U2)), the estimated effect of the design input changes for 10 CFR 50.46 reporting purposes is -381OF for Unit 1 and

-239 0 F for Unit 2. The estimated effect of TCD and peaking factor burndown is the difference between the margin PCT and the integrated PCT, or +384 0 F for Unit 1 and +73 0 F for Unit 2. The results of the evaluations are summarized below.

Summary of CNP Units 1 and 2 Evaluations in Response to NRC 10 CFR 50.54(f) Request TCD/peaking Margin Update Unit AOR PCT Integrated PCT Margin PCT factor burndown PCT Benefit (OF) (OF) (OF) PCT Penalty (OF)

(OF) 1 2128 2131 1747 384 -381 2 2107 1941 1868 73 -239 References for Enclosure 2:

1. Letter from M. G. Evans, U. S. Nuclear Regulatory Commission to L. J. Weber, Indiana Michigan Power Company, "Donald C. Cook Nuclear Plant Units 1 and 2 - Information Request Pursuant To 10 CFR 50.54(f) Related To The Estimated Effect On Peak Cladding Temperature Resulting From Thermal Conductivity Degradation In The Westinghouse-Furnished Realistic Emergency Core Cooling System Evaluation (TAC NO. M99899)," dated February 16, 2012, Agencywide Documents Access and Management System (ADAMS) Accession No. ML12041A384

2. WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," January 2005.

(Westinghouse Proprietary Class 2)

3. WCAP-15063-P-A, Revision 1 with Errata, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," July 2000. (Westinghouse Proprietary Class 2)

4. Letter from Westinghouse Electric Company (WEC) to the U.S. Nuclear Regulatory Commission, "Westinghouse Input Supporting Licensee Response to NRC 10 CFR 50.54(f) Letter Regarding Nuclear Fuel Thermal Conductivity Degradation (Proprietary/Non-Proprietary)", LTR-NRC-12-27, dated March 7, 2012 to AEP-NRC-2012-13 Page 6 of 6

5. NRC Information Notice 2011-21, McGinty, T. J., and Dudes, L. A., "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting From Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011

6. AEP-NRC-7565-01, "Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315 License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," December 2007. (ML080090268). Amendment approved on October 27, 2008 (ML082670351)

7. AEP-NRC-2009-23, "Donald C. Cook Nuclear Plant Unit 2 Docket No. 50-316 License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," March 2009. (ML090930453). Amendment approved on March 31, 2011 (ML110730783)

8. Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "Errors in Containment Backpressure Calculation in Large Break Loss-of-Coolant Accident Analysis," AEP-NRC-2011-35, dated June 16, 2011 (ML11171A655)

Enclosure 3 to AEP-NRC-2012-13 10 CFR 50.46 30-day Report and Proposed Schedule for Reanalysis to AEP-NRC-2012-13 Page 1 of 3 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Plant Name: Donald C. Cook Unit 1 Utility Name: American Electric Power Revision Date: 3/11/2012 Analysis Information EM: ASTRUM (2004) Analysis Date: 11/20/2007 Limiting Break Size: Split FQ: 2.15 FdH: 1.55 Fuel: 15x15 Upgrade SGTP (%): 10 Notes: Post-analysis evaluation of 2% SGTP and FdH of 1.545 Clad Temp ('F)

LICENSING BASIS Analysis-Of-Record PCT 2128 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

1. Update to LOTIC2 Calculated Containment Pressure 0 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Design Input Changes with Respect to Plant Operation -381 C. 2012 ECCS MODEL ASSESSMENTS

1. Evaluation of TCD and Peaking Factor Burndown +384 D. OTHER 1 . None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 2131 to AEP-NRC-2012-13 Page 2 of 3 Westinghouse LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Plant Name: Donald C. Cook Unit 2 Utility Name: American Electric Power Revision Date: 3/11/2012 Analysis Information EM: ASTRUM (2004) Analysis Date: 2/9/2009 Limiting Break Size: Split FQ: 2.335 FdH: 1.644 Fuel: Vantage 5 SGTP (%): 10 Notes: Post-analysis evaluation of 1% SGTP and FdH of 1.61 Clad Temp (OF)

LICENSING BASIS Analysis-Of-Record PCT 2107 PCT ASSESSMENTS (Delta PCT)

A. PRIOR ECCS MODEL ASSESSMENTS

1. Update to LOTIC2 Calculated Containment Pressure 0 B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Design Input Changes with Respect to Plant Operation -239 C. 2012 ECCS MODEL ASSESSMENTS

1. Evaluation of TCD and Peaking Factor Burndown +73 D. OTHER

1. None 0 LICENSING BASIS PCT + PCT ASSESSMENTS PCT = 1941 to AEP-NRC-2012-13 Page 3 of 3 Proposed Schedule for Reanalysis By December 15, 2016, I&M will submit to the NRC for review and approval a LBLOCA analysis that applies NRC-approved methods that include the effects of fuel TCD. The date for the analysis submittal is projected on the following milestones needed to perform a revised licensing basis LBLOCA analysis with an NRC-approved ECCS Evaluation Model that explicitly accounts for TCD:

1) Submittal by Westinghouse, to the NRC for review and approval, of a revised fuel performance and LBLOCA Evaluation Model methodologies that include the effects of TCD.

2) Prior NRC approval of a fuel performance analysis methodology that includes the effects of TCD.

3) Prior NRC approval of a LBLOCA Evaluation Model that includes the effects of TCD and accommodates the ongoing 10CFR50.46(c) rulemaking process.

to AEP-NRC-2012-13 REGULATORY COMMITMENT The following table identifies those actions committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment Date I&M will submit to the NRC for review a LBLOCA analysis December 15, 2016 that applies NRC-approved methods that include the effects of fuel TCD. The date for the analysis submittal is projected on the following milestones needed to perform a revised licensing basis LBLOCA analysis with an NRC-approved ECCS Evaluation Model that explicitly accounts for TCD:

1) Submittal by Westinghouse, to the NRC for review and approval, of a revised fuel performance and LBLOCA Evaluation Model methodologies that include the effects of TCD.

2) Prior NRC approval of a fuel performance analysis methodology that includes the effects of TCD.

3) Prior NRC approval of a LBLOCA Evaluation Model that includes the effects of TCD and accommodates the ongoing 10CFR50.46(c) rulemakinQ process.