ML051520369

10 CFR 50.46 Annual Report

ML051520369
Person / Time
Site:	Oyster Creek
Issue date:	05/25/2005
From:	Cowan P AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2130-05-20100
Download: ML051520369 (8)

Text

1 OCFR50.46 May 25,2005 2 1 30-05-201 00 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Oyster Creek Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

10 CFR 50.46 Annual Report

References:

1) Letter from Michael P. Gallagher (AmerGen Energy Company, LLC) to U. S. Nuclear Regulatory Commission, 10 CFR 50.46 Annual Report, dated June 9,2004

2) GE Letter, 10 CFR 50.46 Notification Letter, 2005-01, April 1, 2005 The purpose of this letter is to transmit the 10 CFR 50.46 reporting information for Oyster Creek Generating Station (OCGS). The previous 50.46 report for OCGS (Reference 1) provided the cumulative Peak Cladding Temperature (PCT) errors for the most recent fuel designs through June 9, 2004. Since the Reference 1 annual report was issued, GE reported that the representative exposure point at which the long duration SAFER run is performed to provide the boundary conditions for the CORCL evaluations may not be bounding and can have a non-conservative effect on the CORCL results (Reference 2). Long duration SAFER runs were performed for each analyzed exposure point to evaluate this condition. The PCT impact for the reported condition was determined to be 0°F for GE9 and GE11 fuel.

Two attachments are included with this letter that provide the current OCGS 10 CFR 50.46 status. Attachment 1, Peak Cladding Temperature Rack-Up Sheet, provides information regarding the PCT for the limiting Large Break Loss of Coolant Accident (LOCA) Analysis evaluations for OCGS. Attachment 2, Assessment Notes, contains a detailed description for each change or error reported.

US. Nuclear Regulatory Commission May 25,2005 Page 2 If you have any questions, please contact Tom Loomis at 610-765-5510.

Very truly yours, Pamela B. cowan Director - Licensing & Regulatory Affairs AmerGen Energy Company, LLC Attachments: 1) Peak Cladding Temperature Rack-Up Sheet

2) Assessment Notes cc: S. J. Collins, USNRC Administrator, Region I P. S. Tam, USNRC Senior Project Manager, OCGS R. J. Summers, USNRC Senior Resident Inspector, OCGS File No. 05033

ATTACHMENT 1 10 CFR 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of May 24,2005 Peak Cladding Temperature Rack-Up Sheet Oyster Creek Generating Station

PLANT NAME:

Ovster Creek REPORT REVISION DATE:

05/24/05 CURRENT OPERATING CYCLE:

20 ECCS EVALUATION MODEL:

SAFER/CORCUGESTR-LOCA Fuel:

Limiting Fuel Type:

Limiting Single Failure:

ANALYSIS OF RECORD Evaluation Model:

GE9, GE11 GE9/GEl1 (same)

ADS Valve

1.

2.

3.

4.

Reference Peak Cladding Temperature (PCT):

NEDC-23785-1 -PA, Rev. 1, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-Of-Coolant Accident Volume 11, SAFER - Long-Term Inventory Model for BWR Loss-Of-Coolant Analysis, October 1 984.

Discharge Pipe 21 50°F NEDC-30996P-A1 SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-jet Pump Plants, Volume I, SAFER - Long-Term Inventory Model for BWR Loss-of-Coolant Analysis, October 1987.

NEDC-32950P, Compilation of Improvements to GENES SAFER ECCS-LOCA Evaluation Model, January 2000. (Application Methodology Description)

NEDC-30996P-A, SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-jet Pump Plants, Volume 11, SAFER Application Methodology for Non-jet Pump Plants, October 1987. (Non-jet Pump Plant - SAFER/CORCL)

Calculations:

1. GE-NE-0000-0001-7486-01 P, Oyster Creek Generating Station Loss-of-Coolant Accident Evaluation for GEI 1, GE Nuclear Energy, dated July 2002.

2. GE-NE-0000-0006-3699-01 P-R1, ECCS-LOCA Evaluation for Oyster Creek with Improved GE9 LHGR Limits, GE Nuclear Energy, dated September 2002.

Page 1 of 2

MARGIN ALLOCATION NET PCT (GE9)

NET PCT (GE11)

A. PRIOR LOCA MODEL ASSESSMENTS 2175°F 21 75°F I New LOCA analyses were performed for both GE9 and I APCT = 0°F I Total PCT Change from Current Assessments Cumulative PCT Change from Current Assessments CAPCT. = 0°F c IAPCT~=O"F B. CURRENT LOCA MODEL ASSESSMENTS NET PCT (GE9) 2175°F 21 75°F

. NET PCT (GE11)

I CORCL Boundary Conditions (See Note 4)

I APCT = 0°F I Page 2 of 2

ATTACHMENT 2 10 CFR 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessment Notes Oyster Creek Generating Station

Prior LOCA Assessment

1.

2.

3.

New LOCA analyses were performed for both GE9 and G E l l fuel in support of operating cycle 19. These analyses supersede all prior LOCA assessments. These analyses incorporate all errors and changes known at that time (as of July 2002).

[

Reference:

GE-NE-0000-0006-3699-01 P-R1, ECCS-LOCA Evaluation for Oyster Creek with Improved GE9 LHGR Limits, GE Nuclear Energy, dated September 2002.1

[

Reference:

GE-NE-0000-0001-7486-01 P, Oyster Creek Generating Station Loss-of-Coolant Accident Evaluation for G E l l, GE Nuclear Energy, dated July 2002.1 From August 2002 until May 2004, GE notified Exelon of two errors applicable to Oyster Creek, identified below (Notes 2 and 3).

The most recent annual 50.46 Report for Oyster Creek erroneously reported no update to the LOCA model assessment for GE9 fuel and correctly reported the new LOCA analysis for the introduction of GE11 fuel. A Peak Clad Temperature of 21 83°F was erroneously reported for GE9 fuel (correct value was 2150°F).

[

Reference:

Letter from Michael P. Gallagher (Exelon) to U.S. NRC, 10 CFR 50.46 Reporting Requirements, 21 30-02-20349, dated December 18, 2002.7 Prior LOCA Assessment GE reported that an error was found in the WEVOL code, which affects the calculated vessel volume in the downcomer region. The free volume in the region of the shroud head is calculated incorrectly, resulting in the calcul d value to be underpredicted by 4 - 10 ft3.

[

Reference:

GE Letter, 10 CFR 50.46 Notification Letter, 2002-05, August 26, 2002.1 Prior LOCA Assessment GE reported that a new heat source term has been postulated. This heat source involves the recombination of hydrogen and oxygen within the fuel bundle during the core heatup.

The additional heat will raise the temperature of the steam heat sink in the bundle, resulting in a potential increase in the peak cladding temperature and local oxidation. This recombination is spontaneous at temperatures above approximately 900°F. The hydrogen is generated by the steam-zirconium reaction during heatup. The oxygen enters the vessel either as a dissolved gas in the ECCS water or through the break when the vessel fully depressurizes and draws the containment non-condensable gases back into the vessel.

The current LOCA evaluation models do not include this new heat source. Pending disposition of this phenomenon, a change notification was supplied to provide the impact of hydrogen-oxygen recombination on the cladding temperature and local oxidation.

[

Reference:

GE Letter, 10 CFR 50.46 Notification Letter, 2003-05, Mav 13, 2004.1 Page 1 of 2

4. Current LOCA Assessment GE reported that the representative exposure point at which the long duration SAFER run is performed to provide the boundary conditions for the CORCL evaluations may not be bounding and can have a non-conservative effect on the CORCL results. Short duration SAFER runs are performed at each analyzed exposure point to provide the fuel bundle initial conditions. Long duration SAFER runs were performed for each analyzed exposure point.

The PCT impact for the reported condition was determined to be 0°F for GE9 and GE11 fuel.

[

Reference:

GE Letter, I0 CFR 50.46 Notification Letter, 2005-01, April 01, 2005.1 Page 2 of 2