ML041760104

10. CFR 50.46 2004 Annual Report

ML041760104
Person / Time
Site:	Oyster Creek
Issue date:	06/09/2004
From:	Gallagher M AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2130-04-20129
Download: ML041760104 (8)

Text

AmerGen SM An Exelon Company AmerGen Energy Company, LLC www.exeloncorp.coM 200 Exelon Way Kennett Square, PA 19348 10 CFR 50.46 June 9, 2004 2130-04-20129 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Oyster Creek Generating Station Operating License No. DPR-16 Docket No.50-219

Subject:

10 CFR 50.46 Annual Report

Reference:

1. Letter from Michael P. Gallagher (AmerGen Energy Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Reporting Requirements," dated December 18, 2002 The purpose of this letter is to transmit additional 10 CFR 50.46 reporting information for Oyster Creek Generating Station (OCGS). The most recent annual 50.46 Report for Oyster Creek (Reference 1) erroneously reported no update to the LOCA mocel assessment for GE9 fuel, and correctly reported the new LOCA analysis for the introduction of GE1 1 fuel. A Peak Clad Temperature of 21830 F was erroneously reported for GE9 fuel. The correct value was 2150 0F.

In addition, GE Nuclear Energy has reported that a new heat source term has been postulated.

This heat source involves the recombination cf hydrogen and oxygen witriin 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. The current LOCA evaluation models do not include this new heat source. Pending disposition of this phenomenon, a change notification has been issued that identifies the impact of hydrogen-oxygen recombination on the cladding temperature and local oxidation.

Twvo attachments are included with this letter that provide the current GCGS 10 CFR 50.46 status. Attachment 1, "Peak Cladding Temperature Rack-Up Sheet," provides updated 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.

OD

U.S. Nuclear Regulatory Commission Page 2 June 9, 2004 If you have any questions, please contact Tom Loomis at 610-765-5510.

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

2) Assessment Notes cc: H. J. Miller, USNRC Administrator, Region I P. S. Tam, USNRC Senior Project Manager, OCGS R. J. Summers, USNRC Senior Resident Inspector, OCGS File No. 02088

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 June 9, 2004 Peak Cladding Temperature Rack-Up Sheet

PLANT NAME: Oyster Creek ECCS EVALUATION MODEL: SAFER/CORCLIGESTR-LOCA REPORT REVISION DATE: 06/09/04 CURRENT OPERATING CYCLE: 19 ANALYSIS OF RECORD Evaluation Model:

1. NEDC-23785-1 -PA, Rev. 1, "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-Of-Coolant Accident Volume II, SAFER -

Long Term Inventory Model for BWR Loss-Of-Coolant Analysis,"

October 1984.

2. NEDC-30996P-A, "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.

3. NEDC-32950P, "Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model," January 2000. (Application Methodology Description)

4. NEDC-30996P-A, "SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-jet Pump Plants, Volume II, 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 GEl 1", GE Nuclear Energy, dated July 2002.

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

Page 1 of 2

Fuel: GE9, GE11 Limiting Fuel Type: GE9/GE11 (same)

Limiting Single Failure: ADS Valve Limiting Break Size and Location: 4.66 ft2 Double-Ended Guillotine (DEG) in a Recirculation Discharge Pipe Reference Peak Cladding Temperature 21500 F (PCT):

MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS New LOCA analyses were performed for both GE9 and APCT = 0 0F GE1 1 fuel in support of operating cycle 19 (See Note 1)

NET PCT (GE9) 2150OF NET PCT (GE1l1) 2150 0F B. CURRENT LOCA MODEL ASSESSMENTS WEVOL Code Error (See Note 2) APCT = 0 0F Hydrogen-Oxygen Recombination (See Note 3) APCT = +25 0F Total PCT Change from Current Assessments XAPCT = +25 0F Cumulative PCT Change from Current Assessments X IAPCT I = 25 0F NET PCT (GE9) 2175 0F NET PCT (GE1l1) 21 75°F 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

1. Prior LOCA Assessment New LOCA analyses were performed for both GE9 and GE1 1 fuel in support of operating cycle 19. These analyses supersede all pri6r 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.]

[

Reference:

GE-NE-0000-0001 -7486-01 P, "Oyster Creek Generating Station Loss-of-Coolant Accident Evaluation for GE1 1", GE Nuclear Energy, dated July 2002.]

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 GE1 1 fuel. A Peak Clad Temperature of 21830 F was erroneously reported for GE9 fuel (correct value was 2150 0F).

[

Reference:

Letter from Michael P. Gallagher (AmerGen Energy Company, LLC) to U.S. NRC, "10 CFR 50.46 Reporting Requirements", 2130-02-20349, dated December 18, 2002.]

2. Current 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 calculated value to be underpredicted by 4- 10 ft3.

[

Reference:

GE Nuclear Energy Letter, "10 CFR 50.46 Notification Letter", 2002-05, August 26, 2002.]

3. Current 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 0F. 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 Page 1 of 2

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 Nuclear Energy Letter, "10 CFR 50.46 Notification Letter", 2003-05, May 13, 2004.]