AEP-NRC-2017-47, Annual Report of Loss-Of-Coolant Accident Evaluation Model Changes

Annual Report of Loss-Of-Coolant Accident Evaluation Model Changes

ML17244A238
Person / Time
Site:	Cook
Issue date:	08/30/2017
From:	Lies Q Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2017-47
Download: ML17244A238 (11)

Text

m INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWER One Cook Place Bridgman, Ml 49106 A unit ofAmerican Electric Power lndianaMichiganPower.com August 30, 2017 AEP-NRC-2017-47 10 CFR 50.46 Docket Nos.: 50-315 50-316 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Units 1 and 2 ANNUAL REPORT OF LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL CHANGES Pursuant to 10 CFR 50.46, . Indiana Michigan Power Company {l&M), the licensee for Donald C. Cook Nuclear Plant (CNP), is transmitting an annual report of loss-of-coolant accident (LOCA) evaluation model changes affecting the peak cladding temperature (PCT) for CNP Unit 1 and Unit 2. l&M is providing, as Enclosure 1, to this letter, the Unit 1 and Unit 2 Large Break and Small Break LOCA Analyses-of-Record PCT values and error assessments for calendar year 2016.

Also pursuant to 10 CFR 50.46, Enclosure 2, to this letter provides descriptions of errors in LOCA models that have been used for Unit 1 and Unfr 2, which have since been corrected, but were not previously reported within 30 days to the NRC pursuant to 10 CFR 50.46.

There are no new or revised commitments in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sin?_~E ls:ne Lies Site Vice President JMT/mll

U. S. Nuclear Regulatory Commission AEP-NRC-2017-47 Page 2

Enclosures:

1. Donald C. Cook Nuclear Plant Units 1 and 2, Large and Small Break Loss-of-Coolant Accident Peak Clad Temperature Summary

2. Donald C. Cook Nuclear Plant Units 1 and 2, Description of Errors in Loss of Coolant Accident Evaluation Models c: R. J. Ancona, MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region Ill J. Rankin, NRC Washington, D.C.

A. J. Williamson, AEP Ft. Wayne, w/o enclosures

ENCLOSURE 1 TO AEP-NRC-2017-47 DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENT PEAK-CLAD TEMPERATURE

SUMMARY

Abbreviations:

10 CFR Title 10 of the Code Of Federal Regulations ADAMS Agencywide Documents Access and Management System CNP Donald C. Cook Nuclear Plant OF degrees Fahrenheit ECCS emergency core cooling system EM evaluation methodology 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 LB large break LOCA loss of coolant accident MWt megawatts - thermal NOP/NOT normal operating pressure/normal operating temperature NRC Nuclear Regulatory Commission PCT peak cladding temperature RHR residual heat removal SGTP steam generator tube plugging SB small break TCD thermal conductivity degradation Summary:

By letter dated March 19, 2012, (ADAMS Accession No. ML12088A104), and supplemented by letter dated June 11, 2012, (ADAMS Accession No. ML12173A025), l&M, the licensee for CNP Units 1 and 2, submitted a report describing the impact of fuel pellet TCD on the LB LOCA ECCS evaluation model, and an estimate of the effect on the predicted PCT for CNP Units 1 and 2. This report was submitted pursuant to 10 CFR Part 50, Section 50.46, Paragraph (a)(3), and referred to a letter from Westinghouse Electric Company dated March 7, 2012, (ADAMS Accession No. ML12072A035). The report was subsequently found to be acceptable by NRC letter dated March 7, 2013, (ADAMS Accession No. ML13077A137).

By letter dated August 30, 2013, (ADAMS Accession No. ML13247A174), l&M, the licensee for CNP Unit 1, submitted a report describing the impact of Revised Heat Transfer Multiplier Distributions on the predicted PCT for CNP Unit 1. This report was submitted pursuant to 10 CFR Part 50, Section 50.46, Paragraph (a)(3). By Westinghouse letter LTR-LIS-13-360, "D. C. Cook Units 1 and 2 10 CFR 50.46 Report for Revised Heat Transfer Multiplier Distributions,"

dated July 31, 2013, Westinghouse Electric Company notified l&M, the licensee for CNP Unit 1, of significant errors in the EM for the LB LOCA analysis of record for CNP Unit 1. By Westinghouse letter LTR-LIS-13-406, "Additional Information on the Evaluation of Revised Heat Transfer Multiplier to AEP-NRC-2017-47 Page 2 Distributions," dated August 14, 2013, Westinghouse Electric Company provided l&M additional detail on the nature of the errors and the corrections made. As documented in the subsequent rack-up sheets the error results in a benefit to the calculated PCT.

By letter dated February 27, 2014, (ADAMS Accession No. ML14063A043), l&M, the licensee for CNP Unit 1, submitted a report describing the impact of an Error in Burst Strain Application on the predicted PCT for CNP Unit 1. This report was submitted pursuant to 1O CFR Part 50, Section 50.46, Paragraph (a)(3). By Westinghouse letter LTR-LIS-14-44, "D. C. Cook Units 1 and 2 10 CFR 50.46 Report for the HOTSPOT Burst Strain Error Correction," dated January 29, 2014, Westinghouse Electric Company notified l&M, the licensee for CNP Unit 1, of significant errors in the EM for the LB LOCA analysis of record for CNP Unit 1.

By letter dated May 20, 2016, (ADAMS Accession No. ML16145A291), l&M, the licensee for CNP Unit 1, submitted a report of significant changes to the ECCS EM as a result of implementation of the modification to restore NOP/NOT in CNP Unit 1. This report was submitted pursuant to 10 CFR Part 50, Section 50.46, Paragraph (a)(3).

The following pages summarize the impact of TCD, peaking factor burndown, heat transfer multiplier distribution revisions, error in burst strain application, decay group uncertainty factors errors, changes to grid blockage ratio and porosity (Unit 2 only), and plant modification evaluations on the CNP Units 1 and 2 LB LOCA analyses of record. In addition, pages are included that summarize the SB LOCA PCT analyses of record for CNP Units 1 and 2.

to AEP-NRC-2017-47 Page 3 CNP UNIT 1 RETIRED UNIT 1 *CYCLE 26 LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break_

Evaluation Model: ASTRUM (2004)

Fa= 2.15 FdH = 1.55 SGTP = 10%ca.J Break Size: Split Analysis Date: November 20, 2007 LICENSING BASIS Analysis-of-Record PCT= 2128°F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. Evaluation of TCD and Peaking Factor Burndown 384°F(a)

2. Revised Heat Transfer Multiplier Distributions -55°F

3. Error in Burst Strain Application 85°F

4. Decay Group Uncertainty Factors Errors -29°F B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Plant Evaluations associated with TCD -381°F(a)

C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 2132°F Notes:

a. These assessments are coupled via an evaluation of burnup effects which include thermal conductivity degradation, peaking factor burndown, and design input changes (e.g., reduction in the maximum allowed steam generator tube plugging from 10% to 2.5% and maximum FdH reduced to 1.545).

to AEP-NRC-2017-47 Page4 CNP UNIT 1 LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Evaluation Model: ASTRUM (2004)

Fo= 2.15 FdH = 1.55 SGTP = 10% Break Size: Split Analysis Date: November 20, 2007 LICENSING BASIS Analysis-of-Record PCT= 2128°F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. Return to NOP/NOT Including Pellet Thermal 404°F(a)

Conductivity Degradation and Peaking Factor Burndown

2. Revised Heat Transfer Multiplier Distributions for -91°F NOP/NOT Conditions

3. Error in Burst Strain Application 85°F

4. Decay Group Uncertainty Factors Errors -29°F B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Design Input Changes with Respect to Plant Operation -489°F(a) for Return to NOP/NOT Evaluation C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER 0°F LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 2008°F Notes:

a. These assessments are coupled via an evaluation of burnup effects which include thermal conductivity degradation, peaking factor burndown, and design input changes.

to AEP-NRC-2017-47 Page 5 CNP UNIT 1 LOCA Peak Clad Temperature Summary for Appendix K Small Break Evaluation Model: NOTRUMP Fa=2.32 FdH=1.55 SGTP=10% 3.25 inch cold leg break Analysis Date: January 6, 2012 Notes: 3304 MWt (plus 0.34% calorimetric uncertainty)

LICENSING BASIS Analysis-of-Record PCT= 1725°F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS B. PLANNED PLANT MODIFICATION EVALUATIONS C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1725°F to AEP-NRC-2017-47 Page6 CNP UNIT 2 LOCA Peak Clad Temperature Summary forASTRUM Best Estimate Large Break Evaluation Model: ASTRlJM (2004)

-Fa= 2.335 FdH = 1.644 SGTP = 10%<a-> Break Size: Split Analysis Date: February 9, 2009 LICENSING BASIS Analysis-of-Record PCT= 2107°F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. Evaluation of TCD and Peaking Factor Burndown 73°F(a)

2. Changes to Grid Blockage Ratio and Porosity 16°F

3. Revised Heat Transfer Multiplier Distributions -3°F

4. Error in Burst Strain Application 13°F B. PLANNED PLANT MODIFICATION EVALUATIONS

1. Plant Evaluations associated with TCD -239°F(a)

2. Evaluation of 12 Stainless Steel Filler Rods in 2 1°F(b)

Reconstituted Fuel Assemblies C. NEW 10 CFR 50.46 ASSESSMENTS

1. None 0°F D. OTHER 0°F LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1968°F Notes:

a. These assessments are coupled via an evaluation of burnup effects which include thermal conductivity degradation, peaking factor burndown, and design input changes (e.g., reduction in the maximum allowed steam generator tube plugging from 10% to 1.5% and maximum FdH reduced to 1.61).

b. This PCT impact is only applicable to the Unit 2 Cycle 23, which began December 2016 and is scheduled to end March 2018.

to AEP-NRC-2017-47 Page 7 CNP UNIT2 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 Notes: The 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 0°F C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER 0°F LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1274°F Notes:

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-2017-47 Donald C. Cook Nuclear Plant Units 1 and 2 Description of Err-ors in Loss of Coolant Accident Evaluation Models In August of 2010, two errors were identified in the large break loss-of-coolant-accident (LBLOCA) models used at Donald C. Cook Nuclear Plant (CNP). These errors are described below. The errors were potentially classified as significant in accordance with 10 CFR 50.46(a)(3)(i) since the LBLOCA model error potentially results in a change to the calculated peak cladding temperature (PCT) of more than 50°F, and the sum of the absolute values of the individual PCT changes resulting from the LBLOCA model errors is greater than 50°F.

ERROR IN LBLOCA MODEL On August 25, 2010, Westinghouse informed Indiana Michigan Power Company (l&M) of two errors involving the Unit 1 and Unit 2 LBLOCA analyses which affected the Unit 1 analyses, which had been approved by the Nuclear Regulatory Commission (NRC) on October 17, 2008. Westinghouse also presented l&M with a Reasonable Assurance of Safe Operation (RASO) based on conservatisms within the analyses.

Nature of the Errors The LBLOCA analysis was performed using a plant specific adaptation of the NRG-approved methodology set forth in Westinghouse Topical Report WCAP 16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)." The ASTRUM evaluation model utilizes the LOTIC2 computer code to determine a containment backpressure boundary condition for the WCOBRA!TRAC computer program used to calculate the PCT. The two errors were identified in the LOTIC2 calculation. The first error involved failure to include, in the containment backpressure calculation, the effects of the injected Emergency Core Cooling System (ECCS) mass and energy that spill from the Reactor Coolant System pipe break. Accounting for the addition of the relatively cooler ECCS water to the containment would cause the LOTIC2 predicted containment backpressure to decrease below that assumed in the WCOBRA!TRAC calculations. The second error involved use of an incorrect, non-conservative, energy conversion factor for mass and energy releases into containment.

Effect on Limiting ECCS Analysis In the August 25, 2010, notification Westinghouse informed l&M that they had not determined the estimate of effect due to the lower containment backpressure using WCOBRA!TRAC at that time.

However, Westinghouse stated that, due to the sensitivity of WCOBRA!TRAC at higher peak cladding temperature (PCT), if the code was to be executed with a lower containment backpressure, a possibility existed-for the predicted PCT to exceed 2200°F. Since the possibility existed for the predicted PCT to exceed 2200°F using an NRG-approved evaluation model based on a realistic computer code, Westinghouse stated that it would be considered appropriate to communicate this issue and the RASO to l&M. The RASO concluded that the conservatisms in the operating conditions assumed in the analysis, in the key physical models in LOTIC2, and in the analysis models contain sufficient margins to support the conclusion that the acceptance criteria to AEP-NRC-2017-47 Page 2 would be met at a high level of probability, as required by 10 CFR 50.46. Following the receipt of this letter, l&M implemented procedural and operational changes and failed to require a quantified impact from Westinghouse.

Commitment to Provide Reanalysis!Take Other Actions to Show Compliance with 10 CFR 50.46 To provide the analytical margin needed to offset the impact of these two errors, the minimum containment spray temperature assumed in the LOTIC2 calculation has been revised. With this revised assumption, the LOTIC2 calculated containment backpressure adequately bounds the containment backpressure that was input to the ASTRUM evaluation model; while accounting for both the above identified errors. l&M has revised the containment spray temperature assumption by changing the procedures that govern operation and testing of the Essential Service Water (ESW) system which provides cooling water to the Containment Spray System (CTS) heat exchanger.

Conclusion 10 CFR 50.46(a)(3)(ii) requires a licensee to report all errors and estimated effects at least annually and "If the change or error is significant, the applicant or licensee shall provide this report within 30 days and include with the report a proposed schedule for providing a reanalysis or taking other action as may be needed to show compliance with § 50.46 requirements." Therefore, pursuant to 10 CFR 50.46, l&M failed to estimate the impact and report the LBLOCA Analyses errors within 30 days of identification.

Inclusion of this enclosure in l&M's 2016 Annual Report will satisfy the prov1s1ons of 1O CFR 50.46(a)(3)(ii), with the exception of the 30-day reporting time limit that was not met. The failure to comply with the 30 day reporting requirement is being addressed by the CNP corrective action program.