ML062340193

Annual Report and Thirty-Day Report of Loss-of-Coolant Accident Evaluation Model Changes

ML062340193
Person / Time
Site:	Cook
Issue date:	08/11/2006
From:	Jensen J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP:NRC:6046
Download: ML062340193 (9)

Text

z Indiana Michigan Power INDIANA Cook Nuclear Plant MICHIGAN One Cook Place Bridgman, MI 49106 AEP.com A unit of American Electric Power August 11, 2006 AEP:NRC:6046 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 I and 2 ANNUAL REPORT AND THIRTY-DAY REPORT OF LOSS-OF-COOLANT ACCIDENT EVALUATION MODEL CHANGES

References:

1. Letter from Joseph N. Jensen, Indiana Micthigan Power Company (I&M), to U.S. Nuclear Regulatory Commission (NRC) Document Control Desk,

'Donald C. Cook Nuclear Plant Unit 1, Thirty-Day Report of Loss-of-Coolant Accident Evaluation Model Changes," AEP:NRC:5046, dated April 29, 2005.

2. Letter from Joseph N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, 10 CFR 50.46 Loss-of-Coolant Accident Reanalysis Schedule," AEP:NRC:4046-01, dated December 28, 2004.

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 and a 30-day report of PCT calculation code changes affecting the calculated PCT for the CNP Unit 1 large break LOCA (LBLOCA) analysis. Attachment 1 to this letter contains the 30-day report data, which describes the recent assessment against the Unit 1 LBLOCA analysis of record. Attachment 2 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 I LBLOCA analysis of record.

By Reference 2, I&M submitted a schedule for reanalysis of the Unit 1 and Unit 2 small break LOCA and the Unit 2 LBLOCA analyses of record. These schedules remain unchanged.

U. S. Nuclear Regulatory Commission AEP:NRC:6046 Page 2 There are no new commitments in this submittal. Should you have any questions, please contact Ms. Susan D. Simpson, Regulatory Affairs Manager, at (269) 466-2428.

Sincerely, qasaplN. Jensen Site Support Services Vice President DB/rdw Attachments c: J. L. Caldwell, NRC Region III K. D. Curry - AEP Ft. Wayne, w/o attachments J. T. King - MPSC, w/o attachments MDEQ - WHMD/RPMWS NRC Resident Inspector P. S. Tam - NRC Washington, DC

ATTACHMENT 1 TO AEP:NRC:6046 ASSESSMENT AGAINST THE UNIT 1 LARGE BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS OF RECORD Indiana Michigan Power Company is submitting a 30-day report of peak clad temperature (PCT) calculation code changes affecting the calculated PCT for the Donald C. Cook Nuclear Plant (CNP) Unit 1 large break loss of coolant accident (LBLOCA) analysis. The calculations for 15 x 15 fuel, using the PAD version 4.0 code, show a 57 degree Fahrenheit ('F) increase in PCT.

These calculations have been performed as part of a separate evaluation for future changes. , Table 1, demonstrates that the PCT value remains within the 2200'F PCT limit specified in 10 CFR 50.46(b)(1).

Assessment Against the Unit 1 LBLOCA Analysis of Record Rebaseline Using PAD 4.0

Background

A 57°F penalty was identified when the rebaseline analysis, with the BASH evaluation model using PAD 4.0 data, was performed for the CNP Unit 1 LBLOCA analysis.

Affected Evaluation Models 1981 Westinghouse LBLOCA Evaluation Model with BASH using the PAD version 4.0 code.

Estimated Effect The calculated PCT with assessments for the Unit 1 LBLOCA is 2175°F and remains below the maximum limit value of 2200'F. The impact on PCT was estimated using a plant-specific LOCBART calculation. As indicated in the PCT accounting in Attachment 2, the effect of the rebaseline using PAD 4.0 data is a 57'F penalty.

Conclusion This transmittal satisfies the annual reporting requirement and 30-day reporting requirement of 10 CFR 50.46(a)(3)(ii). Attachment 2 demonstrates that the PCT value remains within the 2200°F PCT limit specified in 10 CFR 50.46(b)(1).

ATTACHMENT 2 TO AEP:NRC:6046 DONALD C. COOK NUCLEAR PLANT (CNP) UNITS 1 AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE

SUMMARY

Attachment 2 to AEP:NRC:6046 Page I TABLE 1 CNP UNIT 1 LARGE BREAK LOCA Evaluation Model: BASH FQ=2.15 FArH=l.55 SGTP=15% Break Size: C,=0.4 Operational Parameters: RHR System Cross-Tie Valves Closed, 32501 MWt Reactor Power Notes: ZIRLO clad, IFM grids LICENSING BASIS Analysis-of-Record, November 2000 PCT = 2038°F MARGIN ALLOCATIONS (Delta PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. LOCBART Cladding Emissivity Errors -11F

2. Spacer Grid Blocked Area Ratio/Open Area Fraction +370F B. PLANNED 50.59 PLANT CHANGE EVALUATIONS

1. Reduced Containment Spray Temperature +23°F C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER 2

1. Transition Core Penalty +31 0 F

2. Rebaseline Using PAD 4.0 +57 0 F E. LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT = 2175°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.

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

Attachment 2 to AEP:NRC:6046 Page 2 TABLE 2 CNP UNIT 1 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ=2.32 FAH-=l.55 SGTP=30% 3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 32503 MWt Reactor Power Notes: ZIRLO clad, IFM grids LICENSING BASIS Analysis-of-Record, December 2000 PCT= 1720'F MARGIN ALLOCATIONS (A PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. Asymmetric HHSI Delivery +500 F

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

3. Burst and Blockage / Time in Life +111 0F B. NEW 10 CFR 50.46 ASSESSMENTS 0°F C. OTHER 0°F D. LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT= 1990'F 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.

to AEP:NRC:6046 Page 3 TABLE 3 CNP UNIT 2 LARGE BREAK LOCA Evaluation Model: BASH FQ=2.335 FAHI=l.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 PCT = 2051'F MARGIN ALLOCATIONS (A PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. ECCS double disk valve leakage +8°F

2. BASH current limiting break size reanalysis to incorporate +58'F LOCBART spacer grid single phase beat transfer and LOCBART zirc-water oxidation error

3. Cycle 13 ZIRLO Fuel Evaluation -50°F B. PLANNED 50.59 PLANT CHANGE EVALUATIONS

1. Reduced Containment Spray Temperature +470F C. NEW 10 CFR 50.46 ASSESSMENTS 00 F D. OTHER OWF E. LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT =2114*F 4 Power level used as basis for PCT acceptance is 3413 IWt 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 (20511F) 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 2 to AEP:NRC:6046 Page 4 TABLE 4 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ;=2.45 FAIr=l.666 SGTP=15% 3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 3250 MWt Reactor Power 5 LICENSING BASIS Analysis-of-Record, March 1992 PCT= 1956°F MARGIN ALLOCATIONS (A PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. Limiting NOTRUMP and SBLOCA analysis -214°F

2. Burst and blockage/time in life +95 0 F

3. Asymmetric HHSI Delivery +50 0 F

4. NOTRUMP mixture level tracking/region depletion errors +130 F

5. NOTRUMP Bubble Rise/Drift Flux Model Inconsistency +35°F Corrections B. PLANNED 50.59 PLANT CHANGE EVALUATIONS

1. Artificial Leak-By +120F C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER O°F E. LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT = 1947°F 5 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.

to AEP:NRC:6046 Page 5 TABLE 5 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ=2.32 FA1I=l. 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 0 F MARGIN ALLOCATIONS (A PCT)

A. PREVIOUS 10 CFR 50.46 ASSESSMENTS

1. Effect of S I in Broken Loop +150 0 F

2. Effect of Improved Condensation Model -150°F

3. Drift Flux Flow Regime Errors -130F

4. LUCIFER Error Corrections -160F

5. Containment Spray During SBLOCA +20WF

6. Boiling Heat Transfer Correlation Error -60F

7. Steam Line Isolation Logic Error +180F

8. Axial Nodalization, and SBLOCA correction +30 F

9. NOTRUMP Specific Enthalpy Error +20 0F

10. SBLOCA Fuel Rod Initialization Error +10WF

11. Loop Seal Elevation Error -38OF

12. NOTRUMP Mixture Level Tracking/Region Depletion Errors +13 01F

13. NOTRUMP Bubble Rise/Drift Flux Model Inconsistency +35°F Corrections B. PLANNED 50.59 PLANT CHANGE EVALUATIONS

1. Artificial Leak-By +12°F C. NEW 10 CFR 50.46 ASSESSMENTS 0°F D. OTHER 0°F E. LICENSING BASIS PCT+ MARGIN ALLOCATIONS PCT = 1589°F