AEP-NRC-2010-65, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes

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

ML102510126
Person / Time
Site:	Cook
Issue date:	08/27/2010
From:	Carlson M Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2010-65
Download: ML102510126 (9)

Text

INDIANA MICHIGAN AOWER E

A unitofAmerican Electric Power Indiana Michigan Power One Cook Place Bridgman, MI 49106 Indiana Mic higa nPower.conm AEP-NRC-2010-65 10 CFR 50.46 August 27, 2010 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 R. A. Hruby, Indiana Michigan Power Company (I&M), to U. S.

Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Annual Report Of Loss-Of-Coolant Accident Evaluation Model Changes," AEP-NRC-2009-54, dated August 28, 2009 (ADAMS Accession No. ML092520238).

2.

Letter from R. A. Hruby, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit land 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 (ADAMS Accession No. ML100960423).

3. Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2, Annual Report and Thirty-Day Report of Loss of Coolant Accident Evaluation Model Changes,"

AEP:NRC:6046,.

dated August 11, 2006 (ADAMS Accession Number ML062340193).

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) model changes affecting the peak cladding temperature (PCT) for CNP Units 1 and 2. The enclosure to this letter provides the Unit 1 and Unit 2 large break and small break LOCA analyses of record PCT values and error assessments.

Note that the Unit 1 and Unit 2 small break LOCA analyses-of-record identified in the enclosure to this letter differ from those identified in the previous 10 CFR 50.46 annual report (Reference 1). In March 2010, I&M informed the NRC (Reference 2) that, until revised small break' analyses were completed and incorporated into the Updated Final Safety Analysis Report, I&M would consider the PACf

U. S. Nuclear Regulatory Commission AEP-NRC-2010-65 Page 2 Unit 1 and Unit 2 small break LOCA analyses-of-record to be those identified in Reference 3.

Accordingly, the Unit 1 and Unit 2 small break LOCA analyses-of-record identified in the enclosure to this letter are those identified in Reference 3.

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, Michael H. Carlson Vice President - Site Support Services JRW/jmr Enclosure Donald C. Cook Nuclear Plant (CNP) 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 enclosure MDEQ - WHMD/RPS NRC Resident Inspector M. A. Satorius, NRC Region III P. S. Tam - NRC Washington DC

ENCLOSURE TO AEP-NRC-2010-65 DONALD C. COOK NUCLEAR PLANT (CNP) UNITS 1 AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT.ACCIDENT, PEAK CLAD TEMPERATURE

SUMMARY

Enclosure to AEP-NRC-2010-65 Page 1 CNP UNIT 1 LARGE BREAK LOCA Evaluation Model: ASTRUM (2004)

FQ= 2.15 FAH = 1.55 SGTP = 10%

Break Size: Split Operational Parameters: 3304 MWt Reactor Power LICENSING BASIS Analysis-of-Record, October 2008 MARGIN ALLOCATIONS (Delta PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

None B.

PLANNED PLANT MODIFICATION EVALUATIONS

1.

None C.

NEW 10 CFR 50.46 ASSESSMENTS D.

OTHER LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT 2128'F 00F 00 F 00F 00F PCT 2128OF

Enclosure to AEP-NRC-2010-65 Page 2 CNP UNIT 1 SMALL BREAK LOCA Evaluation Model:

NOTRUMP FQ=2.32 FAH=1. 5 5 SGTP=30%

3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 3250 MWt Reactor Power1 Notes: ZI!RLO clad, IFM grids LICENSING BASIS Analysis-of-Record, December 2000 PCT= 1720°F MARGIN ALLOCATIONS (Delta PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

Asymmetric HHSI Delivery

2.

Reduction in Turbine Driven Auxiliary Feedwater Flow

3.

Burst and Blockage / Time in Life NEW 10 CFR 50.46 ASSESSMENTS OTHER B.

C.

+50°F

+1090F

+111 0F 0°F 0°F PCT= 1990°F LICENSING BASIS PCT+ MARGIN ALLOCATIONS 1 The 3250 MWt power level used in the reanalysis is acceptable because it 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.

Enclosure to AEP-NRC-2010-65 Page 3 CNP UNIT 2 LARGE BREAK LOCA Scenario 1 Evaluation Model: BASH FQ= 2.335 FA, = 1.644 SGTP = 15%

Break Size: Cd = 0.6 Operational Parameters: RHR System Cross-Tie Valves Closed, 3413 MWt Reactor Power' LICENSING BASIS Analysis-of-Record, December 1995 MARGIN ALLOCATIONS (Delta PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. ECCS double disk valve leakage

2.

BASH current limiting break size reanalysis to incorporate LOCBART spacer grid single phase heat transfer and LOCBART zirc-water oxidation error

3.

LOCBART Pellet Volumetric Heat Generation Rate Error2 B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1. Cycle 13 ZIRLO Fuel Evaluation

2.

Reduced Containment Spray Temperature C.

NEW 10 CFR 50.46 ASSESSMENTS D.'

OTHER PCT = 2051°F

+80F

+58 0F

+25 0F

-50OF

+470F 0°F 00 F LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2139°F

1. Power level used as basis for PCT acceptance is 3413 MWt due to the reanalysis (see Item A.2) to provide an integrated error effect on the limiting case. This reanalysis (Item A.2) is not considered the analysis-of-record due to the spectrum of break sizes not being reanalyzed to ensure that the limiting break size at 3413 MWt with the errors incorporated would not change. Thus, the analysis-of-record remains as the 1995 analysis at a power level of 3588 MWt. The difference between the limiting case PCT (205 1°F) and the PCT from the reanalysis of that limiting break size at 3413 MWt is the 58°F being reported. The 3413 MWt power level used in the reanalysis is acceptable because it bounds the Unit 2 3468 MWt steady state power limit in the operating license after adjusting for recapture of feedwater flow measurement and power calorimetric uncertainty.

2.

Includes 9°F penalty due to rebaselining of the limiting LOCBART calculation.

Enclosure to AEP-NRC-2010-65 Page 4 CNP UNIT 2 LARGE BREAK LOCA Scenario 2 Evaluation Model: BASH FQ= 2.335 F, = 1.644 SGTP = 15%4 Break Size: Cd = 0.6 Operational Parameters: RHR System Cross-Tie Valves Closed, 3413 MWt Reactor Power 3 LICENSING BASIS Analysis-of-Record, December 1995 MARGIN ALLOCATIONS (Delta PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

ECCS double disk valve leakage

2.

BASH current limiting break size reanalysis to incorporate LOCBART spacer grid single phase heat transfer and LOCBART zirc-water oxidation error

3.

LOCBART Pellet Volumetric Heat Generation Rate Error

4.

Increased Accumulator Water Temperature Evaluation 4 B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1.

Cycle 13 ZIRLO Fuel Evaluation C.

NEW 10 CFR 50.46 ASSESSMENTS D.

OTHER PCT = 205 1°F

+80F

+58 0F

+140F

+27 0F

-50°F 0OF 0OF LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2108OF

3.

Power level used as basis for PCT acceptance is 3413 MWt due to the reanalysis (see Item A.2) to provide an integrated error effect on the limiting case. This reanalysis (Item A.2) is not considered the analysis-of-record due to the spectrum of break sizes not being reanalyzed to ensure that the limiting break size at 3413 MWt with the errors incorporated would not change. Thus, the analysis-of-record remains as the 1995 analysis at a power level of 3588 MWt. The difference between the limiting case PCT (205 I0F) and the PCT from the reanalysis of that limiting break size at 3413 MWt is the 58°F being reported. The 3413 MWt power level used in the reanalysis is acceptable because it bounds the Unit 2 3468 MWt steady state power limit in the operating license after adjusting for recapture of feedwater flow measurement and power calorimetric uncertainty.

4.

Margin allocation A.4 utilized a reduced SGTP of one percent.

Enclosure to AEP-NRC-2010-65 Page 5 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ = 2.45 FAH= 1.666 SGTP = 15%

3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 3250 MWt Reactor Power' LICENSING BASIS Analysis-of-Record, March 1992 MARGIN ALLOCATIONS (DELTA PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

Limiting NOTRUMP and Small Break LOCA analysis

2.

Burst and blockage / time in life

3.

Asymmetric HHSI Delivery

4.

NOTRUMP mixture level tracking / region depletion errors

5.

NOTRUMP Bubble Rise / Drift Flux Model Inconsistency Corrections B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1.

Artificial Leak-By C.

NEW 10 CFR 50.46 ASSESSMENTS D.

OTHER PCT = 19560F

-214OF

+95 0F

+50°F

+130F

+35 0F

+120F 0°F 0°F PCT = 1947°F LICENSING BASIS PCT + MARGIN ALLOCATIONS

1. Unit 2 is licensed to a 3468 MWt steady-state power level. However, 3304 MWt is assumed for the small break LOCA analysis with the safety injection (SI) system cross-tie valves closed. This is because Unit 2 Technical Specification 3.5.2 limits thermal power to 3304 MWt with a SI cross-tie valve closed. The 3250 MWt power level used in the reanalysis is acceptable because it bounds the Unit 2 3304 MWt steady state power limit in the operating license after adjusting for recapture of feedwater flow measurement and power calorimetric uncertainty.

Enclosure to AEP-NRC-2010-65 Page 6 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ = 2.32 FAH= 1.62 SGTP = 15%

4" cold leg break Operational Parameters: SI System Cross-Tie Valves Open, 3588 MWt Reactor Power LICENSING BASIS Analysis-of-Record, August 1992 PCT = 1531°F MARGIN ALLOCATIONS (DELTA PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

Effect of SI in Broken Loop

2.

Effect of Improved Condensation Model

3.

Drift Flux Flow Regime Errors

4.

LUCIFER Error Corrections

5.

Containment Spray During Small Break LOCA

6.

Boiling Heat Transfer Correlation Error

7.

Steam Line Isolation Logic Error

8.

Axial Nodalization, and Small Break LOCA correction

9.

NOTRUMP Specific Enthalpy Error

10.

Small Break LOCA Fuel Rod Initialization Error

11.

Loop Seal Elevation Error

12.

NOTRUMP Mixture Level Tracking / Region Depletion Errors

13.

NOTRUMP Bubble Rise / Drift Flux Model Inconsistency Corrections B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1.

Artificial Leak-By C.

NEW 10 CFR 50.46 ASSESSMENTS D.

OTHER

+150°F

-150OF

-130F

-160F

+20°F

-60F

+180 F

+30F

+20°F

+10°F

-38OF

+130F

+350F

+120F 0OF 0OF PCT = 1589OF LICENSING BASIS PCT + MARGIN ALLOCATIONS