3F0312-04, 10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis

10 CFR 50.46 Notification of Change in Peak Cladding Temperature for Large Break Loss of Coolant Accident Analysis

ML12081A278
Person / Time
Site:	Crystal River
Issue date:	03/19/2012
From:	Wunderly B Progress Energy Florida
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0312-04
Download: ML12081A278 (4)

Text

Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10CFR50.46 March 19, 2012 3F0312-04 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 Large Break Loss of Coolant Accident Analysis

Dear Sir:

Pursuant to 10 CFR 50.46(a)(3)(ii), 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) Large Break Loss of Coolant Accident (LBLOCA) analysis.

This is a notification of two estimated adjustments in the LBLOCA analysis that result in a net change in the PCT of zero for the Cycle 17 LBLOCA analysis. The first LBLOCA adjustment is due to an error in the blow down model control variable that affected the Emergency Core Cooling System bypass calculation. The second LBLOCA adjustment is due to changes in core cooling when the upper plenum column weldments are explicitly modeled.

Each adjustment resulted in a change of 801F to the PCT for the LBLOCA analysis. The PCT decreased by 80'F due to the first LBLOCA adjustment and the PCT increased by 807F due to the second LBLOCA adjustment. The two LOCA model input changes had no effect on the Small Break Loss of Coolant Accident analysis PCT. The attachment to this letter provides additional details.

The resulting maximum PCT for both analyses remains within the 2200'F limit of 10 CFR 50.46. Since there is no net change in the LOCA PCT as a result of these adjustments and the resulting PCT remains below 2200'F, there are no plans for future reanalysis.

This correspondence contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.

Sincerely, ****

Blair P. Wunderly .)

Director-Engineering-Nuclear Crystal River Nuclear Plant BPW/par

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. A oo -

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 Attachment 3F0312-04 Page 1 of 2 Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis Pursuant to 10 CFR 50.46(a)(3)(ii), 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 ('F) in the Crystal River Unit 3 (CR-3) Large Break Loss of Coolant Accident (LBLOCA) analysis.

Reference 1 describes two errors, identified by AREVA, which resulted in changes to the PCT for the Cycle 17 LBLOCA analysis. The LBLOCA errors were discovered during sensitivity studies being completed for the Bellefonte plant restart efforts. The extent of condition for the two errors identified resulted in changes to the LBLOCA analysis for CR-3.

The first error was discovered in the blow down model control variables that calculate the time for total end of Emergency Core Cooling System (ECCS) bypass. AREVA discovered that the end of ECCS bypass calculation was impacted by steam energy changes that result from core flood tank (CFT) injected liquid condensation efficiency and steam reaching the upper down comer region. The control variables incorrectly calculated the steam energy flowing into the upper down comer region. When the control variables were corrected, the end of the bypass time was predicted approximately 2 seconds earlier, resulting in a shorter lower plenum refill period with a quicker onset of lower core quench and therefore lower PCT. AREVA corrected the control variable error and performed a new limiting LBLOCA analysis. The correction shortened the lower plenum refill and decreased the ruptured segment cladding temperature by 80'F. The limiting temperature for unruptured segment cladding decreased by 40'F. The ECCS bypass is not used for a small break loss of coolant accident so these analyses are not affected by this error.

The second error was discovered when a new upper plenum column weldment model was applied which resulted in changes to core cooling. The revised modeling reflects a more detailed noding arrangement in the reactor vessel upper plenum that was used and approved for application in the AREVA BWNT LOCA Evaluation Model (BAW-10192P-A, Rev. 0). The scoping case, with the column weldment modeled over the top of the hot channel, resulted in reduced cooling during portions of the blow down phase. As a result, the end of the blow down fuel temperatures increased and these changes translate into a 40'F increase in the unruptured segment PCT, while the ruptured segment is increased by 80'F. This modeling change was considered for small break loss of coolant accident (SBLOCA) and it was concluded that it will not affect the limiting results because the SBLOCA is a slower evolving transient with up flows in the core hot bundles, such that there is no net change from the presence of a column weldment in the upper plenum.

Each of the two errors resulted in estimated PCT changes for the LBLOCA of 80'F. Nuclear Condition Report 518902 was generated in the CR-3 Corrective Action Program to document the two LBLOCA errors. The two errors offset each other with no net change to the current PCT for CR-3. Therefore, Cycle 17 information provided in Reference 2 remains valid.

REFERENCES

1. Letter from AREVA to CR-3, AREVA-FAB12-110, dated February 21, 2012, "10CFR 50.46 LOCA Report of Two EM Error Corrections (AREVA CR 2012-165: ECCS

U. S. Nuclear Regulatory Commission Attachment 3F0312-04 Page 2 of 2 Bypass Mathematical Error and AREVA CR 2012-757: Upper Plenum Column Weldment EM Change)"

2. CR-3 to NRC letter, 3F1211-14, dated December 14, 2011, "Crystal River Unit 3 - 10 CFR 50.46 Loss-of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report" Cycle 17 Summary The following tables provide the corrected peak clad temperature (PCT) results for Small Break (SB) and Large Break (LB) Loss of Coolant Accidents (LOCAs) for Cycle 17.

CR-3 LBLOCA PCT Change Summary Cycle 17 Full Core of Mark-B-HTP Assemblies Delta PCT PCT Previously Reported PCT N/A 1994-F (10 CFR 50.46 Notification dated December 14, 2011)

AREVA FAB 12-110 and CR-3 NCR 00 F 1994 0 F 518902: Analysis changes due to two evaluation model corrections Cumulative Change 0°F Sum of absolute magnitude of changes I 160OF CR-3 SBLOCA PCT Change Summary Cycle 17 Full Core of Mark-B-HTP Assemblies Delta PCT PCT Previously Reported PCT N/A 1535 0 F (10 CFR 50.46 Notification dated December 14, 2011)

Cumulative Change 0OF Sum of absolute magnitude of changes [ 0F