RS-04-066, 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

ML041320411
Person / Time
Site:	Quad Cities
Issue date:	05/05/2004
From:	Simpson P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-04-066
Download: ML041320411 (5)

Text

Exel o n.

Exelon Generation 4300 Winfield Road www.exeloncorp.com Nuclear Warrenville, IL 60555 10 CFR 50.46 RS-04-066 May 5, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

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

Reference:

Letter from T. J. Tulon (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," SVP-03-063, dated May 8, 2003 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. Thomas G. Roddey at (630) 657-2811.

Rspectfully, Patrick R. Simpson 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 IlIl NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety k00

Attachment A Quad Cities Nuclear Power Station Unit I 10 CFR 50.46 Report PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: SAFER/GESTR-LOCA REPORT REVISION DATE: 05/05/04 CURRENT OPERATING CYCLE: 18A ANALYSIS OF RECORD Evaluation Model:

The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Volume l1l, SAFER/GESTR Application Methodology, NEDE-23785-1-PA, General Electric Company, Revision 1, October 1984.

Calculations:

"SAFERIGESTR-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: 1.0 Double-Ended Guillotine in a Recirculation Suction Pipe Reference Peak Cladding Temperature (PCT) PCT = 2110 0 F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated December 6, 2002 (See Note 2) APCT = 0F 10 CFR 50.46 Report dated May 8, 2003 (See Note 4) APCT = 00 F Net PCT 2110 0F B. CURRENT LOCA MODEL ASSESSMENTS SAFER Level/olume Table Error (See Note 5) APCT =O0 F Steam Separator Pressure Drop Error (See Note 6) APCT = 0F Mid-Cycle Reload of GE14 Fuel (See Note 7) APCT = 0°F Total PCT change from current assessments l ZAPCT = 0F Cumulative PCT change from current assessments l APCT l= 00F Net PCT 2110 OF

Attachment B Quad Cities Nuclear Power Station Unit 2 10 CFR 50.46 Report PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: SAFER/GESTR-LOCA REPORT REVISION DATE: 05/05/04 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 1II,SAFERIGESTR Application Methodology, NEDE-23785-1-PA, General Electric Company, Revision 1, October 1984.

Calculations:

"SAFER/GESTR-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: 1.0 Double-Ended Guillotine in a Recirculation Suction Pipe Reference Peak Cladding Temperature (PCT) PCT = 21100 F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated March 28, 2002 (See Note 1) APCT = 00 F 10 CFR 50.46 Report dated May 9, 2002 (See Note 3) APCT = 00 F 10 CFR 50.46 Report dated May 8, 2003 (See Note 4) APCT = 00 F Net PCT 21100 F B. CURRENT LOCA MODEL ASSESSMENTS SAFER LevelNolume Table Error (See Note 5) APCT = 00F Steam Separator Pressure Drop Error (See Note 6) APCT = 00F Second Reload of GE14 in Cycle 18 Core (See Note 8) APCT = 00F Total PCT change from current assessments ZAPCT = 0F Cumulative PCT change from current assessments l APCT I = 0°F Net PCT 2110°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.

[

Reference:

Letter from Timothy J. Tulon (Exelon) to U.S. NRC, "10 CFR 50.46, 30-Day Report for Quad Cities Unit 2," SVP-02-025, dated March 28, 2002.]

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 LPCI leakage, GE LOCA error in the WEVOL code and change in DG start time requirement were reported. There is no assessment penalty.

[

Reference:

Letter from Timothy J. Tulon (Exelon) to U.S. NRC, "10 CFR 50.46, 30-Day Report for Quad Cities Nuclear Power Station, Unit 1," SVP-02-104, dated December 6, 2002.]

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

[

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 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 0F.

[

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.]

5. Current LOCA Assessment GE reported that an error was found in the initial level/volume table for SAFER. The level/volume tables were generated with incorrect initial water levels. This resulted in an incorrect volume split in the nodes above and below the water surface, and incorrect initial liquid mass. GE determined that the PCT impact of this error on all fuel types to be negligible.

[

Reference:

GE Letter, "10 CFR 50.46 Notification Letter, " 2003-01, May 6, 2003.]

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

6. Current LOCA Assessment GE reported that an error was found in the initial steam pressure drop input to the SAFER model. The calculation of this value for some plant/fuel types applied the wrong loss coefficient or erroneously included a term to account for the hydrostatic pressure. These errors resulted in a higher initial steam separator pressure drop and overly restricted the flow through the separator during the LOCA event. GE determined that the PCT impact of this error on all fuel types to be negligible.

[

Reference:

GE Letter, "10 CFR 50.46 Notification Letter," 2003-03, May 6, 2003.]

7. Current LOCA Assessment Additional 233 new GE14 fuel bundles were introduced into the Quad Cities Unit 1 Cycle 18A core to replace the same number of ATRIUM-9B fuel bundles. GE evaluated this change and determined that the impact on the licensing basis PCT to be 0F.

[

Reference:

Supplemental Reload Licensing Report for Quad Cities 1 Ql Ml 6 Cycle 18A, 0000-0014-8357-SRLR, Revision 0, May 2003.]

8. Current LOCA Assessment A second reload of GE14 fuel was introduced into the Quad Cities Unit 2 Cycle 18 core. GE evaluated this change and determined that the impact on the licensing basis PCT to be 00F.

[

Reference:

Supplemental Reload Licensing Report for Quad Cities Unit 2 Reload 17 Cycle 18, 0000-0024-0751-SRLR, Revision 0, January 2004.]

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