AEP-NRC-2008-27, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes

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

ML082550483
Person / Time
Site:	Cook
Issue date:	08/29/2008
From:	Jensen J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2008-27
Download: ML082550483 (10)

Text

Indiana Michigan Power Company Nuclear Generation Group One Cook Place MEHIA Bridgman, MI 49106 POWE aep.com August 29, 2008 AEP-NRC-2008-27 10 CFR 50.46 Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 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 M. A. Peifer, 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, Thirty-Day Report For Loss-Of-Coolant Accident Evaluation Model Changes," AEP:NRC:7046-01, dated June 15, 2007.

2. Letter from M. A. Peifer, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Small Break Loss-of-Coolant Accident Evaluation Model Reanalysis," AEP:NRC:7046, dated March 29, 2007.

3. Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Completion of Commitment Regarding Small Break Loss-of-Coolant Accident Analysis 8.75-inch Case (TAC No.

MD5297),"

AEP-NRC-2008-1 1, dated July 24, 2008.

4. Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," AEP:NRC:7565-01, dated December 27, 2007.

5. Letter from M. A. Peifer, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Thirty-Day Report for Loss-Of-Coolant Accident Evaluation Model Changes," AEP:NRC:7046-03, dated October 16, 2007.

6. Letter from S.

D.

Simpson, I&M, to 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:7046-02, dated August 31, 2007.

U. S. Nuclear Regulatory Commission AEP-NRC-2008-27 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) model changes affecting the peak cladding temperature (PCT) for CNP Units 1 and 2. The attachment to this letter provides the Unit 1 and Unit 2 large break and small break LOCA analyses of record PCT values and error assessments.

By Reference 1, I&M submitted a schedule for reanalysis of the Unit 2 small and large break LOCA analyses of record. The Unit 1 small break LOCA analysis has been submitted by Reference 2, as supplemented by Reference 3, and the Unit 1 large break LOCA analysis was submitted for Nuclear Regulatory Commission (NRC) review and approval by Reference 4. There was one 30-day report, Reference 5, submitted since the last annual report, Reference 6.

The NRC approved a revised LOCBART Transient Extension Method (TEM) in a Safety Evaluation (SE) dated September 17, 2007, ADAMS Accession Number ML072490348. Consistent with the vendor limitations and conditions described in the subject SE and Reference 10 of the subject SE, CNP Units 1 and 2 are transitioning to realistic large break LOCA analysis methods and not the revised LOCBART TEM. The Unit 1 analysis, as described above, has been submitted to the NRC for review and approval by Reference 4. The reanalysis schedule for the Unit 2 large break LOCA analyses is provided in Reference 1 and remains unchanged.

This letter contains no new or revised commitments.

Should you have any questions, please contact John A. Zwolinski, Manager of Regulatory Affairs, at (269) 466-2478.

Sincerely, 4Josep

. Jensen Si pport Services Vice President RSP/rdw Attachment c:

T. A. Beltz - NRC Washington, DC J. L. Caldwell - NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachment J. T. King - MPSC, w/o attachment MDEQ - WHMD/RPS NRC Resident Inspector

ATTACHMENT TO AEP-NRC-2008-27 DONALD C. COOK NUCLEAR PLANT (CNP) UNITS 1 AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE

SUMMARY

Attachment to AEP-NRC-2008-27 Page 1 CNP UNIT 1 LARGE BREAK LOCA Scenario 1 Evaluation Model: BASH FQ= 2.15 FH = 1.55 SGTP = 15%

Break Size: Cd = 0.4 Operational Parameters: RHR System Cross-Tie Valves Closed, 3250 MWt Reactor Power' LICENSING BASIS Analysis-of-Record, December 2000 MARGIN ALLOCATIONS (Delta PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

LOCBART Cladding Emissivity Errors

2.

Rebaseline Using PAD 4.0

3.

LOCBART Pellet Volumetric Heat Generation Rate Error B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1.

Reduced Containment Spray Temperature

2.

15xI5 Upgrade Fuel C.

New 10 CFR 50.46 ASSESSMENTS D.

OTHER E.

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2038OF

-110 F

+570F

+230F

-590F 0OF 0OF PCT = 2059°F

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.

Attachment to AEP-NRC-2008-27 Page 2 CNP UNIT I LARGE BREAK LOCA Scenario 2 Evaluation Model: BASH FQ= 2.15 FAH = 1.55 SGTP = 15%2 Break Size: Ca = 0.4 Operational Parameters: RHR System Cross-Tie Valves Closed, 3250 MWt Reactor Power3 LICENSING BASIS Analysis-of-Record, December 2000 MARGIN ALLOCATIONS (Delta PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

LOCBART Cladding Emissivity Errors

2.

Rebaseline Using PAD 4.0

3.

LOCBART Pellet Volumetric Heat Generation Rate Error

4.

Increased Accumulator Water Temperature Evaluation 2 B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1.

15x15 Upgrade Fuel C.

New 10 CFR 50.46 ASSESSMENTS D.

OTHER E.

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2038°F

-11 0F

+57 0 F

+11 0F

-16 0F

-59 0 F 0°F 0OF PCT = 2020°F

2.

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

3.

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.

Attachment to AEP-NRC-2008-27 Page 3 CNP UNIT 1 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ = 2.32 FAH = 1.55 SGTP = 10%

3.25" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 3304 MWt Reactor Power LICENSING BASIS Analysis-of-Record, March 2007 PCT =17250 F MARGIN ALLOCATIONS (DELTA PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS

1.

None 0OF B.

PLANNED PLANT MODIFICATION EVALUATIONS

1.

None 0°F C.

NEW 10 CFR 50.46 ASSESSMENTS

1.

None 0°F D.

OTHER

1.

None 0°F E.

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT - 1725°F

Attachment to AEP-NRC-2008-27 Page 4 CNP UNIT 2 LARGE BREAK LOCA Scenario 1 Evaluation Model: BASH FQ= 2.335 FAn4 1.644 SGTP = 15%

Break Size: Cd = 0.6 Operational Parameters: RHR System Cross-Tie Valves Closed, 3413 MWt Reactor Power4 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 Error5 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

+47 0F O0F 0°F E.

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2139°F

4.

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.

5.

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

Attachment to AEP-NRC-2008-27 Page 5 CNP UNIT 2 LARGE BREAK LOCA Scenario 2 Evaluation Model: BASH FQ = 2.335 FH = 1.644 SGTP = 15%6 Break Size: Cd = 0.6 Operational Parameters: RHR System Cross-Tie Valves Closed, 3413 MWt Reactor Power 7 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 6 B.

PLANNED 50.59 PLANT CHANGE EVALUATIONS

1.

Cycle 13 ZIRLO Fuel Evaluation C.

New 10 CFR 50.46 ASSESSMENTS PCT = 2051 "F

+8 0F

+58°F

+140F

+27 0 F

-50OF 0OF D.

OTHER O°F E.

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2108°F

6.

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

7.

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.

Attachment to AEP-NRC-2008-27 Page 6 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 Power8 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 I.

None D.

OTHER

1.

None PCT = 1956 0F

-214OF

+95 0F

+50 0 F

+130F

+35 0 F

+120F 00 F 00F E.

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 1947 0F

8.

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 an 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.

Attachment to AEP-NRC-2008-27 Page 7 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 MARGIN ALLOCATIONS (DELTA PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS I.

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

1.

None D.

OTHER

1.

None PCT= 1531°F

+150°F

-150OF

-130F

-160F

+20°F

• -60F

+18°F

+30F

+20°F

+10°F

-38°F

+130F

+350F

+120F 00 F 0°F E.

LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 15890F