AEP-NRC-2013-32, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes

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

ML13247A172
Person / Time
Site:	Cook
Issue date:	08/30/2013
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2013-32
Download: ML13247A172 (7)

Text

INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant HWER One Cook Place Bridgman, M149106 A unit of American Electric Power India naMichigan Power.com August 30, 2013 AEP-NRC-2013-32 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

References:

1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S.

Nuclear Regulatory Commission (NRC), "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 on the Westinghouse-Furnished Realistic Emergency Core Cooling System Evaluation,"

AEP-NRC-2012-13, dated March 19, 2012, (ADAMS Accession No. ML12088A104).

2. Letter from M. H. Carlson, I&M, to NRC, "Donald C. Cook Nuclear Plant Units 1 and 2 Response to Request for Information 10 CFR 50.46 Report for Emergency Core Cooling System Model Change or Error Associated with Thermal Conductivity Degradation," dated June 11, 2012, (ADAMS Accession No. ML12173A025).

3. Letter from P. S. Tam, NRC, to Lawrence J. Weber, I&M, "D. C. Cook Nuclear Plant (DCCNP), Units 1 and 2 - Closeout of 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," dated April 2, 2012, (ADAMS Accession No. ML12088A376).

4. Letter from T. J. Wengert, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2 - Evaluation of Report concerning Significant Emergency Core Cooling System Evaluation Model Error Related to Nuclear Fuel Thermal Conductivity Degradation (TAC NOS. ME8322 and ME8323)," dated March 7, 2013, (ADAMS Accession No. ML13077A137).

U.S. Nuclear Regulatory Commission AEP-NRC-2013-32 Page 2 Pursuant to 10 CFR 50.46, Indiana Michigan Power Company, 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. CNP is providing, as an enclosure 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 2012.

The impacts to Large Break LOCA PCT due to fuel pellet thermal conductivity degradation were previously reported to the Nuclear Regulatory Commission (NRC) in Reference 1 and supplemented by Reference 2. NRC staff review and acceptance of the impact to PCT due to fuel pellet thermal conductivity degradation was documented in Reference 4. These impacts to the LOCA analyses are discussed in the enclosure, and are included on the PCT reporting sheets.

Several other changes were made to the Large Break LOCA and Small Break LOCA evaluation models during the reporting period. The specific details of these changes were evaluated as having no impact on the calculated PCTs. Since there was no PCT impact due to these changes, they are not included in the PCT reporting sheets.

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.

Sincerely, Joel P. Gebbie Site Vice President DMB/kmh

Enclosure:

Donald C. Cook Nuclear Plant Units 1 and 2, Large and Small Break Loss-of-Coolant Accident Peak Clad Temperature Summary c: J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosures MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III T. J. Wengert, NRC Washington, DC

ENCLOSURE TO AEP-NRC-2013-32 DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE

SUMMARY

Abbreviations:

CNP Donald C. Cook Nuclear Plant OF degrees Fahrenheit ECCS emergency core cooling system EM evaluation methodology FdH nuclear enthalpy rise hot channel factor FQ heat flux hot channel factor HHSI high head safety injection (Safety Injection System at CNP)

I&M Indiana Michigan Power Company LOCA loss of coolant accident MWt megawatts - thermal NRC Nuclear Regulatory Commission PCT peak cladding temperature RHR Residual Heat Removal SGTP steam generator tube plugging 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), I&M, the licensee for CNP Units 1 and 2, submitted a report describing the impact of fuel pellet TCD on the Large Break 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 Title 10 of the Code of Federal Regulations (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).

A new small break LOCA analysis of record was implemented for CNP Unit 1 in 2012. The revised small break LOCA was provided to the NRC by letter dated August 31, 2012 (ADAMS Accession No. ML12256A685). The enclosure to the letter provides a report of the revised analysis. The results of the revised small break LOCA analysis conform to the emergency core cooling system acceptance criteria of 10 CFR 50.46. The calculated peak cladding temperature of 1725 0 F is below the 10 CFR 50.46 limit of 22001F.

The following pages summarize the impact of TCD, peaking factor burndown, and plant modification evaluations on the CNP Units 1 and 2 Large Break LOCA analyses of record. In addition, pages are included that summarize the small break LOCA PCT analyses of record for CNP Units 1 and 2.

Enclosure to AEP-NRC-2013-32 Page 2 CNP UNIT 1 LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Evaluation Model: ASTRUM (2004)

FQ= 2.15 FdH = 1.55 SGTP = 10%(a) 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. None 0OF B. PLANNED PLANT MODIFICATION EVALUATIONS -381 °F(a)

C. NEW 10 CFR 50.46 ASSESSMENTS 384 0 F(a)

D. OTHER 0OF LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2131°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%

and maximum FdH reduced to 1.545). Evaluation details provided in a letter dated March 19, 2012, (ADAMS Accession No. ML12088A104), and supplemented by letter dated June 11, 2012, (ADAMS Accession No. ML12173A025), and subsequently found acceptable by NRC letter dated March 7, 2013 (ADAMS Accession No. ML13077A137).

Enclosure to AEP-NRC-2013-32 Page 3 CNP UNIT 1 LOCA Peak Clad Temperature Summary for Appendix K Small Break Evaluation Model: NOTRUMP FQ=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

1. None 0OF B. PLANNED PLANT MODIFICATION EVALUATIONS 0OF C. NEW 10 CFR 50.46 ASSESSMENTS 0OF D. OTHER 0OF LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1725°F

Enclosure to AEP-NRC-2013-32 Page 4 CNP UNIT 2 LOCA Peak Clad Temperature Summary for ASTRUM Best Estimate Large Break Evaluation Model: ASTRUM (2004)

FQ = 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. None 0OF B. PLANNED PLANT MODIFICATION EVALUATIONS -2390 F(a)

C. NEW 10 CFR 50.46 ASSESSMENTS 73 0 F(a)

D. OTHER 0OF LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 19410 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%

and maximum FdH reduced to 1.61). Evaluation details provided in a letter dated March 19, 2012, (ADAMS Accession No. ML12088A104), and supplemented by letter dated June 11, 2012, (ADAMS Accession No. ML12173A025), and subsequently found acceptable by NRC letter dated March 7, 2013 (ADAMS Accession No. ML13077A137).

Enclosure to AEP-NRC-2013-32 Page 5 CNP UNIT 2 LOCA Peak Clad Temperature Summary for Appendix K Small Break Evaluation Model: NOTRUMP FQ= 2.32 FdH= 1.62 SGTP = 10% 4 inch cold leg break Analysis Date: April 25, 2011 Notes: 3600 MWt 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 0OF C. NEW 10 CFR 50.46 ASSESSMENTS 0OF D. OTHER 0°F LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1274°F Notes:

a. The 3600 MWt power level used in this analysis bounds the Unit 2 3468 MWt steady state power limit in the operating license.