3F1209-13, 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analyses
| ML100040044 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/30/2009 |
| From: | Cahill S Florida Power Corp, Progress Energy Florida |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 3F1209-13 | |
| Download: ML100040044 (3) | |
Text
Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10CFR50.46 December 30, 2009 3F1209-13 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Crystal River Unit 3 - 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss Of Coolant Accident Analyses
Reference:
Crystal River Unit 3 to NRC letter, dated December 16, 2009, "Crystal River Unit 3 - 10 CFR 50.46 Loss-of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report"
Dear Sir:
Pursuant to 10 CFR 50.46(a)(3), Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., hereby provides notification of a change in peak clad temperature (PCT) of greater than 50 degrees Fahrenheit (OF) in the Crystal River Unit 3 (CR-3) Small Break Loss of Coolant Accident (SBLOCA) analysis.
During Refueling Outage 16, the CR-3 Once Through Steam Generators (OTSGs) were replaced.
The new model OTSGs change the amount of Emergency Feedwater (EFW) reaching the OTSG tubes, reducing the cooling capacity during a SBLOCA. This increases the SBLOCA PCT by less than 620F, with a new PCT of approximately 131 0°F. The OTSG replacement has no impact on the Large Break Loss of Coolant Accident (LBLOCA) PCT. The attachment to this letter provides additional details. The resulting maximum PCT remains within the 2200°F limit of 10 CFR 50.46.
No new regulatory commitments are made in this letter.
If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.
Sincerely, Stephen J. Cahill Manager Engineering Crystal River Nuclear Plant SJC/pdk
Attachment:
Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis xc:
NRR Project Manager Regional Administrator, Region II Senior Resident Inspector Progress Energy Florida, Inc.
Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428
Progress Energy Florida, INC.
Crystal River Unit 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 Attachment Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis
U. S. Nuclear Regulatory Commission 3F1209-13 Attachment Page 1 of 1 Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., is providing the following information pursuant to 10 CFR 50.46(a)(3). Changes to the Evaluation Model or Peak Cladding Temperature (PCT) have occurred since FPC provided the last report by letter dated December 16, 2009. Current PCT results for Small Break (SB) and Large Break (LB)
Loss of Coolant Accidents (LOCAs) are provided in the following Tables.
CR-3 LB LOCA PCT Change Summary Cycle 17 Full Core of Mark-B-HTP Assemblies Delta PCT PCT Previously Reported PCT N/A 19940F (2009 Change Report dated December 16, 2009)
Analysis changes due to the 0
19940F replacement of steam generators Cumulative Change 0
Sum of absolute magnitude of changes 0
CR-3 SB LOCA PCT Change Summary Cycle 17 Full Core of Mark-B-HTP Assemblies Delta PCT PCT Previously Reported PCT N/A 12480F (2009 Change Report dated December 16, 2009)
Analysis changes due to the 62 0F 13 1 0°F replacement of steam generators Cumulative Change 620F Sum of absolute magnitude of changes 620F