RA-09-047, CFR 50.46 Annual Report

CFR 50.46 Annual Report

ML091590224
Person / Time
Site:	Oyster Creek
Issue date:	06/05/2009
From:	David Helker Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-09-047
Download: ML091590224 (8)

Text

VJwvJ.exeloncorp.com 10 CFR 50.46 RA-09-047 June 5,2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

10 CFR 50.46 Annual Report

Reference:

1) Letter from David P. Helker (AmerGen Energy Company, LLC) to U.S. Nuclear Regulatory Commission, "10 CFR 50.46 Annual Report,"

dated June 6, 2008 The purpose of this letter is to transmit the 10 CFR 50.46 reporting information for Oyster Creek Nuclear Generating Station (OCNGS). The previous 50.46 report for OCNGS (Reference 1) provided the cumulative Peak Cladding Temperature (PCT) errors for the most recent fuel designs through June 6, 2008.

Since the referenced annual report was issued, no vendor notifications of Emergency Core Cooling System (ECCS) model error/changes that are applicable to OCNGS have been issued. Also, no ECCS-related changes or modifications have occurred at OCNGS that affect the assumptions of the ECCS analyses.

Two attachments are included with this letter that provide the current OCNGS 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 OCNGS. Attachment 2, "Assessment Notes," contains a detailed description for each change or error reported.

There are no commitments contained in this letter. If you have any questions, please contact Tom Loomis at 610-765-5510.

Very truly yours, 9~?~

David P. Helker Manager - Licensing

U.S. Nuclear Regulatory Commission June 5,2009 Page 2 Attachments: 1) Peak Cladding Temperature Rack-Up Sheet

2) Assessment Notes cc: S. J. Collins, USNRC Administrator, Region I G. E. Miller, USNRC Project Manager, OCNGS M. S. Ferdas, USNRC Senior Resident Inspector, OCNGS

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 5, 2009 Peak Cladding Temperature Rack-Up Sheet Oyster Creek Nuclear Generating Station

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of June 5, 2009 Attachment 1 Peak Cladding Temperature Rack-Up Sheet, OCNGS Page 1 of 2 PLANT NAME: Oyster Creek ECCS EVALUATION MODEL: SAFER/CORCUGESTR-LOCA REPORT REVISION DATE: 06/05/09 CURRENT OPERATING CYCLE: 22 ANALYSIS OF RECORD Evaluation Model:

1. NEDC-23785-1-PA Rev. 1, "The GESTR-LOCA and SAFER Models for the Evaluation ot the Loss-Ot-Coolant Accident Volume II, SAFER - Long Term Inventory Model tor BWR Loss-Ot-Coolant Analysis," October 1984.

2. NEDC-30996P-A, "SAFER Model tor Evaluation ot Loss-ot-Coolant Accidents tor Jet Pump and Non-jet Pump Plants, Volume I, SAFER - Long Term Inventory Model tor BWR Loss-ot-Coolant Analysis," October 1987.

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

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

Calculations:

1. GE-NE-0000-0001-7486-01P, "Oyster Creek Generating Station Loss-of-Coolant Accident Evaluation tor GE11 ," 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.

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 Reterence Peak Cladding Temperature (PCT)

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of June 5,2009 Attachment 1 Peak Cladding Temperature Rack-Up Sheet, OCNGS Page 2 of 2 MARGIN ALLOCATION A. PRIOR lOCA MODEL ASSESSMENTS New lOCA analyses were performed for both GE9 and LiPCT =OaF GE11 fuel in support of operatinq cycle 19 (See Note 1)

WEVOl Code Error (See Note 2) LiPCT =OaF Hydrogen-Oxygen Recombination (See Note 3) LiPCT =+25°F CORCl Boundary Conditions (See Note 4) LiPCT =OaF Hydrogen-Oxygen Recombination (See Note 5) LiPCT = -25°F 2007 Annual Report (See Note 6) LiPCT =OaF 2008 Annual Report (See Note 7) LiPCT = OaF NET PCT (GE11) 2150°F B. CURRENT lOCA MODEL ASSESSMENTS None (See Note 8) LiPCT =OaF Total PCT Change from Current Assessments LLiPCT =OaF Cumulative PCT Chanqe from Current Assessments L ILiPCT I =OaF NET PCT (GE11) 2150°F

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 Assessments as of June 5, 2009 Assessment Notes Oyster Creek Nuclear Generating Station

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of June 5,2009 Attachment 2 Assessment Notes, OCNGS Page 1 of 2

1. Prior LOCA Assessment New LOCA analyses were performed for both GE9 and GE11 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.]

[

Reference:

GE-NE-0000-0001-7486-01 P, "Oyster Creek Generating Station Loss-of-Coolant Accident Evaluation for GE11 ," 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 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 2183°F was erroneously reported for GE9 fuel (correct value was 2150°F).

[

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. 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 calculated value to be underpredicted by 4 - 10 fe.

[

Reference:

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

3. 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, May 13, 2004.]

Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments as of June 5, 2009 Attachment 2 Assessment Notes, OCNGS Page 2 of 2

4. Prior 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 OaF for GE9 and GEll fuel.

[

Reference:

GE letter, "10 CFR 50.46 Notification letter," 2005-01, April 01, 2005.]

5. Prior lOCA Assessment In item 3 above, 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. GE has performed a detailed evaluation of this phenomenon and has determined that there is sufficient conservatism in the Appendix K analysis which bounds the Upper Bound PCT and Oxidation with hydrogen-oxygen recombination in both PCT limited and oxidation limited exposure ranges. Therefore, the current SAFER/CORCl application methodology for conformance of the Appendix K analysis 10 CFR50.46 limits remains applicable. The hydrogen-oxygen recombination phenomenon does not need to be considered in the Appendix K analysis.

[

References:

GE letter, "10 CFR 50.46 Notification letter," 2003-05, Rev. 2. April 27, 2006 and AmerGen letter 2130-06-20347, David P. Helker to U.S. NRC, "10 CFR 50.46 Annual Report,"

dated June 8, 2006.]

6. Prior lOCA Assessment The reference letter documented that no ECCS model errors/changes were reported since the prior Oyster Creek annual report was issued.

[

Reference:

letter from David P. Helker (AmerGen Energy Company, llC) to U.S. Nuclear Regulatory Commission, 2130-07-20499, "10 CFR 50.46 Annual Report," dated June 07, 2007.]

7. Prior lOCA Assessment The reference letter documented that no ECCS model errors/changes were reported since the prior Oyster Creek annual report was issued.

[

Reference:

letter from David P. Helker (AmerGen Energy Company, llC) to U.S. Nuclear Regulatory Commission, RA-08-050, "10 CFR 50.46 Annual Report," dated June 06, 2008.]

8. Current lOCA Assessment Since the last annual report (see Note 7), no vendor notifications of Emergency Gore Cooling System (ECCS) model error/changes that are applicable to Oyster Creek have been issued. Also, no EGGS-related changes or modifications have occurred at Oyster Greek that affect the assumptions of the ECGS analyses.