AEP-NRC-2012-69, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes

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

ML12255A404
Person / Time
Site:	Cook
Issue date:	08/31/2012
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2012-69
Download: ML12255A404 (6)

Text

INDIANA MICHIGAN Indiana Michigan Power POWER One Cook Place Bridgman, MI 49106 A upit of American Electric Power India naMichiganPower.com August 31, 2012 AEP-NRC-2012-69 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 (I&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. 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 2011.

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

ENCLOSURE TO AEP-NRC-2012-69 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 FAH 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 SI Safety Injection

Enclosure to AEP-NRC-2012-69 Page 1 CNP UNIT 1 LARGE BREAK LOCA Evaluation Model: ASTRUM (2004) 0 FQ= 2.15 FAH = 1.55 SGTP = 10% Break Size: Split Operational Parameters: 3304 MWt (plus 0.34% uncertainty) Reactor Power LICENSING BASIS Analysis-of-Record, November 20071 PCT = 2128°F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. None 0OF B. PLANNED PLANT MODIFICATION EVALUATIONS

1. None 0OF 2

C. NEW 10 CFR 50.46 ASSESSMENTS

1. None 0°F D. OTHER

1. None 0°F LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2128°F The CNP Unit 1 Large Break LOCA Analysis-of-Record identified in this report is the same as the Analysis-of-Record identified in the annual report submitted in 2011 (ML11252B081). However, the designated analysis date has been changed to reflect the date that the analysis was performed.

2 Letter from J. P. Gebbie, I&M, to the NRC Document Control Desk, "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)," AEP-NRC-2012-13, dated March 19, 2012, (ML12088A104) provided the estimated impact on the CNP Unit 1 and Unit 2 Large Break LOCA Evaluation Model from fuel thermal conductivity degradation. This impact will be addressed in the 10 CFR 50.46 annual report submitted for calendar year 2012.

Enclosure to AEP-NRC-2012-69 Page 2 CNP UNIT 1 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ=2.32 FAH=1.55 SGTP=30% 3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 3250 MWt (plus 2% uncertainty) Reactor Power' LICENSING BASIS Analysis-of-Record, June 20002.3 PCT= 1720'F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. Asymmetric HHSI Delivery +50°F

2. Reduction in Turbine Driven Auxiliary Feedwater Flow +109 0 F

3. Burst and Blockage / Time in Life +111 0F B. PLANNED PLANT MODIFICATION EVALUATIONS

1. None 00 F C. NEW 10 CFR 50.46 ASSESSMENTS

1. None O0 F D. OTHER

1. None 0°F LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1990'F The 3250 MWt power level used in this analysis bounds the Unit 1 3304 MWt steady state power limit in the operating license after adjusting for recapture of feedwater flow measurement and power calorimetric uncertainty.

2 The CNP Unit 1 Small Break LOCA Analysis-of-Record identified in this report is the same as the Analysis-of-Record identified in the annual report submitted in 2011 (ML11252B081). However, the designated analysis date has been changed to reflect the date that the analysis was performed.

3 Letter from R. A. Hruby, I&M, to the NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Docket Nos. 50-315 and 50-316, Schedule for Submittal of Revised Unit 1 and Unit 2 Small Break Loss of Coolant Accident Analyses Addressing Residual Heat Removal System Spray Issue," AEP-NRC-2010-30, dated March 29, 2010, (ML100960423) informed the NRC staff that a revised Unit 1 Small Break LOCA analysis would be transmitted no later then August 31, 2012.

The revised analysis is being transmitted via separate correspondence, and will be reflected in the 10 CFR 50.46 annual report submitted for calendar year 2012.

Enclosure to AEP-NRC-2012-69 Page 3 CNP UNIT 2 LARGE BREAK LOCA Evaluation Model: ASTRUM (2004)

FQ= 2.3 3 5 FAH = 1.644 SGTP = 10% Break Size: Split Operational Parameters: 3468 MWt (plus 0.34% uncertainty) Reactor Power LICENSING BASIS Analysis-of-Record, February 2009 PCT = 2107'F A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. None 00 F B. PLANNED PLANT MODIFICATION EVALUATIONS

1. None 00 F C. NEW 10 CFR 50.46 ASSESSMENTS1 I. None 00 F D. OTHER

1. None 00 F LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2107'F Letter from J. P. Gebbie, I&M, to the NRC Document Control Desk, "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)," AEP-NRC-2012-13, dated March 19, 2012, (ML12088A104) provided the estimated impact on the CNP Unit 1 and Unit 2 Large Break LOCA Evaluation Model from fuel thermal conductivity degradation. This impact will be addressed in the 10 CFR 50.46 annual report submitted for calendar year 2012.

Enclosure to AEP-NRC-2012-69 Page 4 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ= 2.32 FAH= 1.62 SGTP= 10% 4inch cold leg break Operational Parameters: RHR injection flow diversion to RHR spray and HHSI tram cross-tie valves open, 3600 MWt (plus 0.34% uncertainty) Reactor Power' LICENSING BASIS Analysis-of-Record, April 2011 PCT= 1274 0 F A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. None 00 F B. PLANNED PLANT MODIFICATION EVALUATIONS

1. None 0°F C. NEW 10 CFR 50.46 ASSESSMENTS

1. None 00 F D. OTHER

1. None 00 F LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT= 1274°F The 3600 MWt power level used in this analysis bounds the Unit 1 3468 MWt steady state power limit in the operating license.