ML072540683
| ML072540683 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 08/31/2007 |
| From: | Simpson S Indiana Michigan Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:7046-01, AEP:NRC:7046-02 | |
| Download: ML072540683 (8) | |
Text
INDIANA MICHIGAN POWERO A unit of American Electric Power August 31, 2007 Indiana Michigan Power Cook Nuclear Plant One Cook Place Bridgman, MI 49106 AERcom AEP:NRC:7046-02 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
Reference:
Letter from M. A. Peifer, Indiana Michigan Power Company, to U. S. Nuclear Regulatory Commission Document Control Desk, "Donald C. Cook Nuclear Plant Units I and 2, Thirty-Day Report For Loss-Of-Coolant Accident Evaluation Model.
Changes," AEP:NRC:7046-01, dated June 15, 2007.
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 the referenced letter, I&M submitted a schedule for reanalysis of the Unit I and Unit 2 small break LOCA and the Unit 2 large break LOCA analyses of record.
These schedules remain unchanged.
There are no new commitments in this submittal. Should you have any questions, please contact me at (269) 466-2428.
Sincerely, Susan D. Simpson Regulatory Affairs Manager RSP/rdw Attachment (5 pages) poý
U. S. Nuclear Regulatory Commission AEP:NRC:7046-02 Page 2 c:
J. L. Caldwell, NRC Region III K. D. Curry - AEP Ft. Wayne., w/o attaclmient J. T. King - MPSC, w/o attachment MDEQ - WHMD/RPMWS NRC Resident Inspector P. S. Tam - NRC Washington, DC
ATTACHMENT TO AEP:NRC:7046-02 DONALD C. COOK NUCLEAR PLANT (CNP) UNITS I AND 2 LARGE AND SMALL BREAK LOSS-OF-COOLANT ACCIDENT PEAK CLAD TEMPERATURE
SUMMARY
Attachment to AEP:NRC:7046-02 Page 1 TABLE 1 CNP UNIT 1 LARGE BREAK LOCA Evaluation Model: BASH FQ = 2.15 Fa1, = 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.
15X15 Upgrade Fuel C.
New 10 CFR 50.46 ASSESSMENTS
- 1. None D.
OTHER
- 1. None E.
LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2038OF
-11OF
+570 F
+110F
+23 0 F
-590F 0OF 0°F PCT = 2059°F 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:7046-02 Page 2 TABLE 2 CNP UNIT I 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= 1725 0F MARGIN ALLOCATIONS (DELTA PCT)
A.
PREVIOUS 10 CFR 50.46 ASSESSMENTS
- 1.
None 0°F B.
PLANNED PLANT MODIFICATION EVALUATIONS
- 1.
None 0°F C.
NEW 10 CFR 50.46 ASSESSMENTS
- 1.
None 00F D.
OTHER I.
None 0°F E.
LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 17250F
Attactmuent to AEP:NRC:7046-02 Page 3 TABLE 3 CNP UNIT 2 LARGE BREAK LOCA Evaluation Model: BASH FQ= 2.335 F,,, = 1.644 SGTP = 15%
Break Size: Cd = 0.6 Operational Parameters: RHR System Cross-Tie Valves Closed, 3413 MWt Reactor Power2 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.
Rebaseline of Limiting LOCBART Calculation 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
- 1. None D.
OTHER
- 1.
None PCT = 2051 °F
+8°F
+58 0 F
+ 160 F
+9°F
-50°F
+47°F 0°F 0OF E.
LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 2139OF 2 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 IF) 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:7046-02 Page 4 TABLE 4 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ = 2.45 FH = 1.666 SGTP = 15%
3" cold leg break Operational Parameters: SI System Cross-Tie Valves Closed, 3250 MWt Reactor Power3 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 PCT = 1956 0 F 3.
4.
Asynmmetric HHSI Delivery NOTRUMP mixture level tracking / region depletion errors
-214°F
+95 0 F
+50°F
+130F
+350 F
+120 F
- 5.
NOTRUMP Bubble Rise / Drift Flux Model Inconsistency Corrections B.
PLANNED 50.59 PLANT CHANGE EVALUATIONS I.
Artificial Leak-By C.
NEW 10 CFR 50.46 ASSESSMENTS
- 1.
None D.
OTHER
- 1.
None 0OF 00 F E.
LICENSING BASIS PCT + MARGIN ALLOCATIONS PCT = 19470F 3 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.
Attachment to AEP:NRC:7046-02 Page 5 TABLE 5 CNP UNIT 2 SMALL BREAK LOCA Evaluation Model: NOTRUMP FQ = 2.32 FAHl = 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
- 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
- 1.
None D.
OTHER
- 1.
None PCT= 1531°F
+150°F
-150°F
-130F
-160F
+20°F
-60F
+18°F
+3 0F
+20°F
+10°F
-38OF
+130F
- +350F
+120F O°F OF E.