RS-05-056, Transmittal of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for light-water Nuclear Power Reactors, Annual Report

Transmittal of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for light-water Nuclear Power Reactors, Annual Report

ML051250593
Person / Time
Site:	Quad Cities
Issue date:	05/05/2005
From:	Simpson P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-05-056
Download: ML051250593 (5)

Text

Exelon Generation W~W.~Xelo~COT~.Co~

4300 Winfieid Road Warrenville, IL 60555 10 CFR 50.46 RS-05-056 May 5,2005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Transmittal of I 0 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Annual Report for Quad Cities Nuclear Power Station, Units 1 and 2

Reference:

Letter from Patrick R. Simpson (Exelon Generation Company, LLC) to U. S. NRC, "Transmittal of 10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,'

Annual Report for Quad Cities Nuclear Power Station, Units 1 and 2,"

RS-04-066, dated May 5, 2004 The purpose of this letter is to provide the annual report required by 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," for Quad Cities Nuclear Power Station, Units 1 and 2. The attachments describe the changes in accumulated Peak Cladding Temperature (PCT) since the previous annual submittal (Reference).

Should you have any questions concerning this letter, please contact Mr. Dave Gullott at (630) 657-2819.

Respectfully, Manager - Licensing Attachments: Attachment A: Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report Attachment B: Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report Attachment C: Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes cc: Regional Administrator - NRC Region Ill NRC S Cit ea Illinois - of

Attachment A Quad Cities Nuclear Power Station Unit 1 I 0 CFR 50.46 Report PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: SAFEWGESTR-LOCA REPORT REVISION DATE: 05/05/05 CURRENT OPERATING CYCLE: 19 ANALYSIS OF RECORD Evaluation Model:

The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume Ill, SAFEWGESTR Application Methodology, NEDE-23785-1-PA, General Electric Company, Revision 1, October 1984.

Calculations:

"SAFEWGESTR-LOCA Loss-of-Coolant Accident Analysis for Dresden Nuclear Station 2 and 3 and Quad Cities Nuclear Station Units 1 and 2," NEDC-32990P, Revision 2, GE Nuclear Energy, September 2003.

Fuel Analyzed in Calculation: GE9/10, ATRIUM-9B and GE14 Limiting Fuel Type: GE14 Limiting Single Failure: Diesel Generator Limiting Break Size and Location: 1.O Double-Ended Guillotine in a Recirculation Suction Pipe Reference Peak Cladding Temperature (PCT) PCT = 21 10°F MARGlN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS B. CURRENT LOCA MODEL ASSESSMENTS GE LOCA Model Change due to New Heat Source (See Note 6) APCT = 0°F GE14 Fuel Reload (See Note 7) APCT = 0°F Total PCT change from current assessments CAPCT = 0°F Cumulative PCT change from current assessments C IAPCT = 0°F 2110 "F Net PCT

Attachment B Quad Cities Nuclear Power Station Unit 2 10 CFR 50.46 Report PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: SAFEWGESTR-LOCA REPORT REVISION DATE: 05105/05 CURRENT OPERATING CYCLE: -

18 ANALYSIS OF RECORD Evaluation Model:

The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume Ill, SAFERIGESTR Application Methodology, NEDE-23785-1-PA, General Electric Company, Revision 1, October 1984.

Calculations:

"SAFEWGESTR-LOCA Loss-of-Coolant Accident Analysis for Dresden Nuclear Station 2 and 3 and Quad Cities Nuclear Station Units 1 and 2," NEDC-32990P, Revision 2, GE Nuclear Energy, September 2003.

Fuel Analyzed in Calculation: GE9/10, ATRIUM-96 and GE14 Limiting Fuel Type: GE14 Limiting Single Failure: Diesel Generator Limiting Break Size and Location: 1.O Double-Ended Guillotine in a Recirculation Suction Pipe Reference Peak Cladding Temperature (PCT) PCT = 2110°F MARGl N ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS

. I 0 CFR 50.46 Report dated May 9,2002 (See Note 3) APCT = 0°F 10 CFR 50.46 Report dated May 8,2003 (See Note 4) APCT = 0°F 10 CFR 50.46 Report dated May 5,2004 (See Note 5) APCT = 0°F I Net PCT I 2110°F I B. CURRENT LOCA MODEL ASSESSMENTS GE LOCA Model Change due to New Heat Source (See Note 6) APCT = 0°F Total PCT change from current assessments CAPCT = 0°F Cumulative PCT change from current assessments c (APCT] = OOF

, Net PCT 21 10°F

Attachment C Quad Cities Nuclear Power Station Units 1 and 2 10 CFR 50.46 Report Assessment Notes

1. Prior LOCA Model Assessment The 50.46 letter dated March 28, 2002 reported a new LOCA analysis to support extended power uprate (EPU) and transition to GE14 fuel for Quad Cities Unit 2.

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Reference:

Letter from Timothy J. Tulon (Exelon) to US. NRC, 10 CFR 50.46, 30-Day Report for Quad Cities Unit 2, SVP-02-025, dated March 28, 2002.1

2. Prior LOCA Assessment A new LOCA analysis was performed to support EPU and transition to GE14 fuel for Quad Cities Unit 1. In the referenced letter, the impact of CS and LPCl leakage, GE LOCA error in the WEVOL code and change in DG start time requirement were reported. There is no assessment penalty.

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Reference:

Letter from Timothy J. Tulon (Exelon) to US. NRC, 10 CFR 50.46, 30-Day Report for Quad Cities Nuclear Power Station, Unit 1, SVP-02-104, dated December 6,2002.1

3. Prior LOCA Assessment In the referenced letter, no LOCA model assessment was reported for Unit 2 PCT.

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Reference:

Letter from Timothy J. Tulon (Exelon) to US. NRC, Transmittal of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light water nuclear power reactors, Annual Report for Quad Cities Units 1 and 2, SVP-02-039, dated May 9, 2002.]

4. Prior LOCA Assessment The referenced letter provided the annual 50.46 report for Units 1 and 2. This letter reported no LOCA model assessment for Unit 1 whereas it reported the impact of GE LOCA error in the WEVOL code and change in DG start time requirement for Unit 2. The PCT impact for these errors was determined to be OF.

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Reference:

Letter from Timothy J. Tulon (Exelon) to U.S. NRC, Transmittal of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light water nuclear power reactors, Annual Report for Quad Cities Nuclear Power Station, Units 1 and 2, SVP-03-063, dated May 8, 2003.1

5. Prior LOCA Assessment The referenced letter provided the annual 50.46 report for Units 1 and 2. This letter reported GE LOCA errors related to SAFER levellvolume table and Steam

Attachment C Quad Cities Nuclear Power Station Units 1 and 2 10 CFR 50.46 Report Assessment Notes Separator pressure drop and mid-cycle reload of GE14 fuel for Unit 1 (Cycle 18A). For Unit 2, this letter reported the same GE LOCA errors and second reload of GE14 fuel in Cycle 18 core. The PCT impact for these errors and reloads of GE14 fuel was determined to be O°F.

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Reference:

Letter from Patrick R. Simpson (Exelon) to US. NRC, Transmittal of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors, Annual Report for Quad Cities Nuclear Power Station, Units 1 and 2, RS-04-066, dated May 5, 2004.1 Current LOCA Assessment GE has postulated a new heat source applicable to the LOCA event. This heat source is due to recombination of hydrogen and excess oxygen drawn into the vessel from containment during core 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-condensible gases back into the vessel. The current LOCA evaluation model does not account for the effect of this heat source, which has potential to raise the steam temperature while leading to an increase in PCT and local oxidation. GE has evaluated the effect of this additional heat source for the jet pump plants like Quad Cities and determined that the impact is insignificant. This is because of the fact that oxygen from containment enters the vessel after the core is reflooded for the jet pump plants. Therefore, the PCT impact for all fuel types is zero and the effect on local oxidation is negligible.

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Reference:

General Electric 10 CFR 50.46 Notification Letter 2003-05, May 13, 2004.1 Current LOCA Assessment Quad Cities Unit 1 Cycle 19 is to start up in April 2005 with a new reload of GE14 fuel. The impact of this reload was evaluated by GE and reported to be negligible. GE determined that there is no PCT impact because of the change due to the new reload of GE14 fuel.

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Reference:

Supplemental Reload Licensing Report for Quad Cities 1 Reload 18 Cvcle 19. 0000-0028-1626-SRLR, Revision 0, January 2005.1