ML032510877

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

ML032510877
Person / Time
Site:	Cook
Issue date:	08/27/2003
From:	Zwolinski J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:3046
Download: ML032510877 (11)

Text

Indiana Michigan Power Company 500 Circle Drive Buchanan, Ml 49107 1395 INDIANA MICHIGAN POWER August 27, 2003 AEP:NRC:3046 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, D. C. 20555-0001 Donald C.- Cook Nuclear Plant Units 1 and 2 ANNUAL REPORT OF LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL CHANGES

Reference:

Letter from S. A. Greenlee, Indiana Michigan Power Company, to U. S. Nuclear Regulatory Commission Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2 Annual Report of Loss-Of-Coolant Accident Evaluation Model Changes," dated August 31, 2001 Pursuant to 10 CFR 50.46, Indiana Michigan Power Company (I&M), the licensee for the 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. Attachment 1 to this letter describes the current assessments against the large break and small break LOCA analyses of record. provides the large break and small break LOCA analyses of record PCT values and error assessments. Attachment 2 also demonstrates that all PCT values remain within the 2200 degree Fahrenheit PCT limit specified in 10 CFR 50.46(b)(1).

There are no new PCT assessments against the Unit 1 limiting large break, Unit 1 limiting small break, Unit 2 limiting large break, and Unit 2 limiting small break LOCA analyses of record. Therefore, no assessments are classified as significant in accordance with 10 CFR 50.46(a)(3)(i).

Because the previously reported changes to the Unit 1 limiting small break, the Unit 2 limiting large break, and Unit 2 limiting small break analyses of record were classified as significant, I&M submitted a schedule for performing new analyses in the referenced letter. I&M originally planned to complete the Unit 2 small break LOCA and large break LOCA analyses to support the CNP power uprate. I&M has subsequently contracted with a different fuel vendor, and will 401j

U. S. Nuclear Regulatory Commission Page 2 AEP:NRC:3046 complete the power uprate analyses with the new vendor. A revised schedule for reanalysis of each of these events is provided in Attachment 3.

Should you have any questions, please contact Mr. Brian A. McIntyre, Manager of Regulatory Affairs, at (269) 697-5806.

Sincerely, of Design Engineering and Regulatory Affairs DB/rdw Attachments c:

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

U. S. Nuclear Regulatory Commission AEP:NRC:3046 Page 3 bc:

T. E. Anderson M. J. Finissi D. W. Jenkins, w/o attachments J. A. Kobyra, w/o attachments R. J. Kohrt B. A. McIntyre, w/o attachments J. E. Newmiller D. J. Poupard R. P. Powers M. W. Rencheck M. K. Scarpello, w/o attachments T. K. Woods, w/o attachments J. A. Zwolinski, w/o attachments

ATTACHMENT 1 TO AEP:NRC:3046 ASSESSMENTS AGAINST THE LOSS-OF-COOLANT ACCIDENT (LOCA)

ANALYSES OF RECORD Indiana Michigan Power Company (I&M) previously submitted the annual 10 CFR 50.46 report for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2 in the referenced letter. The reported analysis of record peak cladding temperature (PCT) values in Attachment 2 of this letter remain the same as reported in the referenced letter, as no new loss-of-coolant accident (LOCA) analyses have been performed. There are no new PCT assessments against the CNP Units 1 and 2 large break LOCA or small break LOCA analyses of record. Current assessments against the analysis of record are reflected in the PCT accounting in Attachment 2. This transmittal satisfies the annual reporting requirement of 10 CFR 50.46(a)(3)(ii).

Reference:

Letter from S. A. Greenlee, Indiana Michigan Power Company to U. S. Nuclear Regulatory Commission Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes," submittal AEP:NRC:2046-01, dated August 15, 2002

ATTACHMENT 2 TO AEP:NRC:3046 DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE

SUMMARY

to AEP:NRC:3046 Page I TABLE 1 CNP UNIT 1 LARGE BREAK LOCA Evaluation Model:

BASH FQ=2.15 Fmf-1.55 SGTP=15%

Break Size: Cd=0.4 Operational Parameters: RHR System Cross-Tie Valves Closed, 32501 MWt Reactor Power Notes: ZIRLO clad, IFM grids LICENSING BASIS Analysis-of-Record, December 2000 MARGIN ALLOCATIONS (A PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS 2

1.

LOCBART Cladding Emissivity Errors3

2.

Reduced Containment Spray Temperature B.

NEW 10 CFR 50.46 ASSESSMENTS C.

OTHER

1.

Transition Core Penalty4 PCT= 2038 0F

-11F

+23 0F 0F

+31 0F PCT= 2081OF D.

LICENSING BASIS PCT+ MARGIN ALLOCATIONS l The 3250 MWt power level used in the reanalysis bounds the Unit 1 3304 MWt steady state power limit in the operating license crediting recapture of feedwater flow measurement and power calorimetric uncertainty.

2 ECCS model assessments are no longer being listed by year of occurrence. Instead the errors are being identified by error type. This is consistent with Westinghouse reporting methods and does not change the overall PCT.

3 This is a revised assessment. The prior generic assessment of +60F has been changed to -11 0F based on plant specific information.

4This penalty will be dropped once all fuel assemblies include the Intermediate Flow Mixing (IFM) Grids.

to AEP:NRC:3046 Page 2 TABLE 2 CNP UNIT 1 SMALL BREAK LOCA Evaluation Model:

NOTRUMP FQ-=2.32 Fw=1.55 SGTP=30%

3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 32501 MWt Reactor Power Notes: ZIRLO clad, IFM grids LICENSING BASIS Analysis-of-Record, December 2000 MARGIN ALLOCATIONS (A PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS 2 PCT= 17200F 1.

2.

Asymmetric HHSI Delivery Reduction in Turbine Driven Auxiliary Feedwater Flow B.

C.

3.

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

+500F

+1090F

+1 I IF O0F O0F D.

LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1990 0F l The 3250 MWt power level used in the reanalysis bounds the Unit 1 3304 MWt steady state power limit in the operating license crediting recapture of feedwater flow measurement and power calorimetric uncertainty.

2 ECCS model assessments are no longer being listed by year of occurrence. Instead the errors are being identified by error type. This is consistent with Westinghouse reporting methods and does not change the overall PCT.

to AEP:NRC:3046 Page 3 TABLE 3 CNP UNIT 2 LARGE BREAK LOCA Evaluation Model:

BASH Fq=2.335 F2ff 1.64 SGTP415%

Break Size: Cd=0.6 Operational Parameters: RHR System Cross-Tie Valves Closed, 3413 MWt Reactor Power' LICENSING BASIS Analysis-of-Record, December 19951 PCT= 205 1OF MARGIN ALLOCATIONS (A PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS 2

1.

ECCS double disk valve leakage

+80F

2.

BASH current limiting break size reanalysis to incorporate

+58 0F LOCBART spacer grid single phase heat transfer and LOCBART zirc-water oxidation errort

3.

Cycle 13 ZIRLO Fuel Evaluation3

-50 0F

4.

Reduced Containment Spray Temperature

+47 0F B.

NEW 10 CFR 50.46 ASSESSMENTS 0F C.

OTHER 0F D.

LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 2114OF

' 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 IF) and the PCT from the reanalysis of that limiting break size at 3413 MWt is the 58 0F being reported. The 3413 MWt power level used in the reanalysis bounds the Unit 2 3468 MWt steady state power limit in the operating license crediting recapture of feedwater flow measurement and power calorimetric uncertainty.

2 ECCS model assessments are no longer being listed by year of occurrence. Instead the errors are being identified by error type. This is consistent with Westinghouse reporting methods and does not change the overall PCT.

3 The ZIRLO fuel evaluation used a version of LOCBART that corrected for the Vapor Film Flow Regime Heat Transfer and Cladding Emissivity Errors. As reported in previous reports, these errors were -15'F and -100F respectively. Thus, since this reanalysis incorporates the errors previously reported, the errors are no longer being reported individually.

to AEP:NRC:3046 Page 4 TABLE 4 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model:

NOTRUMP FQ=2.45 Fm=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 PCT= 19560F MARGIN ALLOCATIONS (A PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS 2

1.

Limiting NOTRUMP and SBLOCA analysis3

-2140F

2.

Burst and blockage / time in life

+600F

3.

Asymmetric HHSI Delivery

+50 0F

4.

NOTRUMP mixture level tracking/region depletion errors

+13 0 F B.

NEW 10 CFR 50.46 ASSESSMENTS 0F C.

OTHER 0F D.

LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 18650F 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 SI system cross-tie valves closed after adjusting for recapture of feedwater flow measurement and power calorimetric uncertainty.

This is because Unit 2 Technical Specification 3.5.2 limits thermal power to 3304 MWt with a safety injection cross-tie valve closed. The 3250 MWt power level used in the reanalysis bounds the Unit 2 3304 MWt steady state power limit in the operating license crediting recapture of feedwater flow measurement and power calorimetric uncertainty.

2 ECCS model assessments are no longer being listed by year of occurrence. Instead the errors are being identified by error type. This is consistent with Westinghouse reporting methods and does not change the overall PCT.

3 This reanalysis is considered an evaluation because a full spectrum of break sizes was not analyzed.

This reanalysis incorporated the errors previously reported (Letter from M. W. Rencheck, Indiana Michigan Power Company to Nuclear Regulatory Commission Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2 Annual Report of Loss-of-Coolant Accident Evaluation Model Changes," submittal C1000-07, dated October 27,2000) in the individual years in which they occurred. The difference between the analysis-of-record limiting break size PCT and the reanalysis PCT is -214'F. Thus, since this reanalysis incorporates the errors previously reported, the errors are no longer being reported individually. Note that this does not impact the resulting PCT as it remains at 1865°F. It is only an accounting change.

to AEP:NRC:3046 Page 5 TABLE 5 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model:

NOTRUMP FO=2.32 Faff1.62 SGTP=15%

4" cold leg break Operational Parameters: SI System Cross-Tie Valves Open, 35881 MWt Reactor Power LICENSING BASIS Analysis-of-Record, August 1992 MARGIN ALLOCATIONS (A PCT)

A.

PREVIOUS 10 CFR 50.46 ASSESSMENTS2

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 SBLOCTA correction

9.

NOTRUMP Specific Enthalpy Error

10.

SBLOCTA Fuel Rod Initialization Error

11.

Loop Seal Elevation Error

12.

NOTRUMP Mixture Level Tracking / Region Depletion Errors B.

NEW 10 CFR 50.46 ASSESSMENTS C.

OTHER D.

LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1531OF

+1500F

-1500F

-130 F

-160 F

+200 F

-60F

+18 0F

+30F

+200F

+10F

-380 F

+130F 0F 0F PCT= 15420 F l The 3588 MWt power level used in the reanalysis bounds the Unit 2 3468 MWt steady state power limit in the operating license crediting recapture of feedwater flow measurement and power calorimetric uncertainty.

2 ECCS model assessments are no longer being listed by year of occurrence. Instead the errors are being identified by error type. This is consistent with Westinghouse reporting methods and does not change the overall PCT.

V-r ATTACHMENT 3 TO AEP:NRC:3046 COMMITMENTS The following table identifies those actions committed to by Indiana Michigan Power Company (I&M) in this document. Any other actions discussed in this submittal represent intended or planned actions by I&M. They are described to the Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.

Commitment Date A new Unit 2 SBLOCA analysis of both the safety injection cross-tie Startup for Unit 2 valves closed case and the safety injection cross-tie valves open case will Cycle 16.

be submitted.

A new Unit 2 LBLOCA analysis will be submitted.

Startup for Unit 2 Cycle 16.