AEP-NRC-2018-21, 30-Day Report of Changes to or Errors in an Evaluation Model

30-Day Report of Changes to or Errors in an Evaluation Model

ML18130A580
Person / Time
Site:	Cook
Issue date:	05/04/2018
From:	Lies Q Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2018-21
Download: ML18130A580 (5)

Text

m INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER One Cook Place Bridgman, Ml 49106 A unit ofAmerican Electric Power lndianaMichiganPower.com May 4, 2018 . AEP-NRC-2018-21 10 CFR 50.46 Docket No.: 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 2 30-DAY REPORT OF CHANGES TO OR ERRORS IN AN EVALUATION MODEL Pursuant to 10 CFR 50.46, Indiana Michigan Power Company (l&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 2, is reporting significant changes to or errors in emergency core cooling system evaluation model (EM), or in the application of such a model that affects the calculated peak fuel cladding temperature (PCT). By report dated October 2, 2017, Westinghouse notified l&M of EM changes, for a planned plant modification, which significantly affected the Small-Break Loss-:of-Coolant Accident (SBLOCA) analysis for CNP Unit 2. The impact of this is not significant to the CNP Unit 1 SBLOCA Analysis Calculated PCT. The CNP Unit 1 and Unit 2 Large-Break LOCA analyses are not affected by this EM change. to this letter provides a description of the SBLOCA EM change and the associated impact to the CNP Unit 2 SBLOCA analysis of record and the analysis performed for the CNP Unit 2 upflow conversion modification. The upflow conversion modification field work was completed on April 17, 2018. Based on information provided by Westinghouse, an assessment of the EM change resulted in a PCT increase of 75°F for Unit 2.

The estimated impact on the CNP SBLOCA EM 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 proposed reanalysis schedule is provided in to this letter.

A new regulatory commitment is provided as Enclosure 2 to this letter.. Should you have any questions, please contact Mr. Michael K. $carpello, Regulatory Affairs Director, at (269) 466-2649.

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~ane Lies Site Vice President DB/kmh

U.S. Nuclear Regulatory Commission AEP-NRC-2018-21 Page 2

Enclosures:

1. Donald C. Cook Nuclear Plant Unit 2 Report of Significant Changes Related to Westinghouse Small-Break Loss-of-Coolant Analysis Emergency Core Cooling System Evaluation Model

2. Regulatory Commitment c: R. J. Ancona - MPSC A. W. Dietrich, NRC Washington, D.C.

MDEQ- RMD/RPS NRC Resident Inspector K. S. West, NRC Region 111 A. J. Williamson -AEP Ft. Wayne, w/o enclosures

Enclosure 1 to AEP-NRC-2018-21 Donald C. Cook Nuclear Plant Unit 2 Report of Significant Changes Related to Westinghouse Small-Break Loss-of-Coolant Analysis Emergency Core Cooling System Evaluation Model Abbreviations:

CNP Donald C. Cook Nuclear Plant OF degrees Fahrenheit FdH nuclear enthalpy rise hot channel factor Fa heat flux hot channel factor HHSI high head safety injection (Safety Injection System at CNP) l&M Indiana Michigan Power Company LOCA loss of coolant accident MWt megawatts - thermal PCT peak cladding temperature RHR Residual Heat Removal SGTP steam generator tube plugging Summary Pursuant to 10 CFR 50.46, l&M, the licensee for CNP, is submitting a 30-day report of LOCA evaluation model changes resulting in a significant change in calculated PCT for the CNP Unit 2 Small Break LOCA analysis. A significant change is defined as a change or error identified in the model which results in a calculated change to PCT greater than 50°F or cumulative changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 50°F.

The following page summarizes the impact of the Unit 2 Reactor Vessel Upflow Conversion modification on the CNP Unit 2 Small Break LOCA analysis of record. The upflow conversion was completed in accordance with CNP Engineering Change 55458, Unit 2 Upflow Conversion modification. Note that the PCT impact on the Unit 2 Large Break LOCA due to the Reactor Vessel Upflow Conversion modification was not significant and is *not reported herein.

to AEP-NRC-2018-21 Page 2 CNP UNIT 2 LOCA Peak Clad Temperature Summary for Appendix K Small Break Evaluation Model: NOTRUMP Fa= 2.32 FdH = 1.62 SGTP = 10% 4 inch cold leg break Analysis Date: April 25, 2011 Note: 3600 MWt power level used in this analysis bounds the Unit 2 3468 MWt steady state power limit in the operating license.

LICENSING BASIS Analysis-of-Record PCT= 1274°F (a)

MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. None 0°F B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Reactor Vessel Upflow Conversion 75°F C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER 0°F LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1349°F Note:

a. Analysis models RHR injection flow diversion to RHR spray and HHSI cross-tie valves open during cold leg recirculation.

ENCLOSURE 2 TO AEP-NRC-2018-21 REGULATORY COMMITMENT The following table identifies the revised action committed to by Indiana Michigan Power Company (l&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by l&M. They are described to the U.S. Nuclear Regulatory Commission (NRG) for the NRC's information and are not regulatory commitments.

Commitment Date l&M will submit to the NRG for review a Unit 2 small The Unit 2 SBLOCA analysis break loss of coolant accident (SBLOCA) analysis will be submitted 28 months that applies NRG approved methods using from the approval of the Westinghouse's WCAP-16996-P, "Realistic LOCA supplement to topical report Evaluation Methodology Applied to Full Spectrum of WCAP-16996-P, and any Break Sizes (Full Spectrum LOCA Methodology)." required supplements that The date for the submittal of the analysis will align support the new 10 CFR 50.46 with an existing commitment to submit a Unit 2 large rule and would be needed for break loss of coolant accident analysis, (Reference the analysis.

letter from J. P. Gebbie, l&M, to NRG, "Donald C. Cook Nuclear Plant Units 1 and 2, U. S.

Nuclear Regulatory Commission Commitment Change Related to Estimated Effect of Peak Cladding Temperature Resulting from Thermal Conductivity Degradation," dated June 9, 2015).