AEP-NRC-2011-54, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes

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

ML11252B081
Person / Time
Site:	Cook
Issue date:	08/30/2011
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2011-54
Download: ML11252B081 (9)

Text

INDIANA MICHIGAN A unit of American Electric Power Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, MI 49106 AEPcom August 30, 2011 AEP-NRC-2011-54 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

Reference:

1)

Letter from P. S. Tam, U. S. Nuclear Regulatory Commission (NRC), to M. W. Rencheck, Indiana Michigan Power Company (I&M), "Donald C. Cook Nuclear Plant, Unit 1 - Issuance of Amendment to Renewed Facility Operating License Regarding Use of the Westinghouse ASTRUM Large Break Loss-Of-Coolant Accident Analysis Methodology (TAC No. MD7556),"

dated October 17, 2008 (ADAMS Accession No. ML082670351).

2)

Letter from P. S. Tam, NRC, to L. J. Weber, I&M, "Donald C. Cook Nuclear Plant, Unit 2 (CNP-2) - Issuance of Amendment to Adopt a New Large Break Loss-Of-Coolant Accident Analysis (TAC No.

ME1017),"

dated March 31, 2011, (ADAMS Accession No. ML110730783).

3)

Letter from J. P. Gebbie, I&M, to U. S. NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2, Docket No. 50-316, Response to Second Request for Additional Information Regarding a License Amendment Request Associated With the Large-Break Loss-Of-Coolant Accident Analysis Methodology (TAC No. ME1017)," dated February 24, 2011, AEP-NRC-2011-15 (ADAMS Accession Number ML110680210).

4)

Letter from M. H. Carlson, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2, Docket No. 50-316, Errors in Containment Backpressure Calculation in Large Break Loss-of-Coolant Accident Analysis," dated June 16, 2011, AEP-NRC-2011-35 (ADAMS Accession No. ML11171A655).

U,S. Nuclear Regulatory Commission AEP-NRC-2011-54 Page 2 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. 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 for calendar year 2010.

During the second half of 2010, I&M was notified of two errors involving the Unit 1 and Unit 2 large break LOCA analyses that had been performed using a plant-specific adaptation of the Westinghouse ASTRUM methodology.

The Unit 1 analysis affected by the errors had been approved by the NRC in October 2008 (Reference 1) and was therefore the analysis of record in 2010. The Unit 2 analysis affected by the errors was approved in 2011 (Reference 2) and was therefore not the analysis of record in 2010. The NRC was informed of the errors in the Unit 2 analysis via Reference 3.

The initial method for resolving the errors consisted of changing the limiting single failure assumed in the analyses. This method resulted in no impact on the calculated PCT. Errors that do not impact PCT are not reported under 10 CFR 50.46. Therefore, the errors are not included in the enclosed Unit 1 large break LOCA PCT summary for 2010. Since the Unit 2 analysis using the plant-specific adaptation of the ASTRUM methodology had not been approved by the NRC in 2010, the two errors are not applicable to the analysis addressed in the enclosed Unit 2 large break LOCA PCT summary for 2010. In 2011, the method for resolving these errors for both units was revised to be as described for Unit 2 in Reference 4. Although the planned method of resolving the errors was revised, there will continue to be no impact on the calculated PCT.

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/jen 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-2011-54 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)

IFM intermediate flow mixer LOCA loss of coolant accident MWt megawatts - thermal PCT peak cladding temperature RHR Residual Heat Removal SGTP steam generator tube plugging

Enclosure to AEP-NRC-2011-54 Page 1 CNP UNIT 1 LARGE BREAK LOCA Evaluation Model: ASTRUM (2004)

FQ= 2.15 F* = 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 2128OF 00F 00F 00F 00F PCT 2128OF

Enclosure to AEP-NRC-2011-54 Page 2 CNP UNIT 1 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ=2.32 FAr=l.55 SGTP=30%

3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 3250 MWt Reactor Power1 Notes: ZIRLO 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

+109 0F

+11 10F 0OF 0OF PCT= 1990°F LICENSING BASIS PCT+ MARGIN ALLOCATIONS 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-2011-54 Page 3 CNP UNIT 2 LARGE BREAK LOCA Scenario 1 Evaluation Model: BASH FQ = 2.335 Fan = 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 = 2051IF

+80F

+580F

+25 0F

-50°F

+470F OOF O°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 I°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-2011-54 Page 4 CNP UNIT 2 LARGE BREAK LOCA Scenario 2 Evaluation Model: BASH FQ= 2.335 FAH = 1.644 SGTP = 15%4 Break Size: Cd = 0.6 Operational Parameters: RHR System Cross-Tie Valves Closed, 3413 MWt Reactor Power3 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 Evaluation4 B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1. Cycle 13 ZIRLO Fuel Evaluation C.

NEW 10 CFR 50.46 ASSESSMENTS PCT = 2051 OF

+80F

+58 0F

+140F

+270F

-50OF 0OF 0OF D.

OTHER LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2108°F

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 (2051'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.

4.

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

Enclosure to AEP-NRC-2011-54 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

-214°F

+95 0F

+50°F

+13 0F

+35 0F

+12 0F 0°F 0°F PCT = 19470F 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-2011-54 Page 6 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ= 2.32 FH= 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 = 153 1°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

-150°F

-130F

-160F

+20°F

-6°F

+18°F

+30F

+20°F

+10 0F

-38OF

+13 0F

+350F

+120F 0°F 0°F PCT= 15890F LICENSING BASIS PCT + MARGIN ALLOCATIONS