RS-08-064, CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors Annual Report

CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors Annual Report

ML081280785
Person / Time
Site:	Quad Cities
Issue date:	05/07/2008
From:	Hansen J Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-08-064
Download: ML081280785 (9)

Text

Exelon.

Nuclear Exelon Nuclear www,exeloncorp .com 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.46 RS-08-064 May 7, 2008 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:

10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Annual Report

Reference:

Letter from Patrick R . Simpson (Exelon Generation Company, LLC) to U. S . NRC, "10 CFR 50 .46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,' Annual Report,"

dated May 7, 2007 This letter provides 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 (QCNPS), 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 . John Schrage at (630) 657-2821 .

Jeffrey. H'dnsen 4x,_~

Manager - Licensing Attachments:

Attachment 1 : Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (GE Fuel)

Attachment 2: Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (GE Fuel)

Attachment 3: Quad Cities Nuclear Power Station Unit 1, 10 CFR 50.46 Report (Westinghouse Fuel)

Attachment 4: Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel)

Attachment 5: Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes

Attachment 1 Quad Cities Nuclear Power Station Unit 1 10 CFR 50.46 Report (GE Fuel)

PLANT NAME : Quad Cities Unit 1 ECCS EVALUATION MODEL: SAFE R/GESTR-LOCH REPORT REVISION DATE: 03/28/08 CURRENT OPERATING CYCLE: 20 ANALYSIS OF RECORD Evaluation Model:

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

Calculations :

"SAFE R/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 = 2110°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated December 6, 2002 See Note 2 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 10 CFR 50.46 Report dated May 5, 2005 See Note 6 APCT = 0°F 10 CFR 50.46 Report dated May 5, 2006 See Note 7 APCT = 0°F 10 CFR 50.46 Report dated May 7, 2007 See Note 8 APCT = 0°F Net PCT 2110 OF B. CURRENT LOCA MODEL ASSESSMENTS None APCT = 0°F Total PCT change from current assessments JAPCT = 0°F Cumulative PCT change from current assessments APCT =0°F Net PCT 2110 °F

Attachment 2 Quad Cities Nuclear Power Station Unit 2 10 CFR 50.46 Report (GE Fuel)

PLANT NAME : Quad Cities Unit 2 ECCS EVALUATION MODEL: SAFER/GESTR-LOCH REPORT REVISION DATE : 03/28/08 CURRENT OPERATING CYCLE: 20 ANALYSIS OF RECORD Evaluation Model :

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

Calculations :

"SAFE R/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 = 2110°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 Report dated March 28, 2002 (See Note 1) APCT = 0°F 10 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 10 CFR 50.46 Report dated May 5, 2005 See Note 6 APCT = 0°F 10 CFR 50.46 Report dated May 5, 2006 See Note 7 APCT = 0°F 10 CFR 50.46 Report dated May 7, 2007 See Note 8 APCT = 0°F Net PCT 2110°F B. CURRENT LOCA MODEL ASSESSMENTS None APCT = 0°F Total PCT change from current assessments Y-OPCT = 0°F Cumulative PCT change from current assessments Y APCT = 0°F Net PCT 211 0°F

Attachment 3 Quad Cities Nuclear Power Station Unit 1 10 CFR 50.46 Report (Westinghouse Fuel)

PLANT NAME : Quad Cities Unit 1 ECCS EVALUATION MODEL : USA5 REPORT REVISION DATE : 03/28/08 CURRENT OPERATING CYCLE: 20 ANALYSIS OF RECORD Evaluation Model:

"Westinghouse BWR ECCS Evaluation Model : Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004 .

Calculations :

"Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021QC-LOCA, Revision 3, Westinghouse Electric Company, LLC., April 2007.

Fuel Analyzed in Calculation : SVEA-96 Optima2 Limiting Fuel Type : SVEA-96 Optima2 Limiting Single Failure : LPCI injection valve Limiting Break Size and Location: 1 .0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT) PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None - new Analysis for Optima2 fuel transition APCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS New analysis (See Note 9) APCT = 0°F Total PCT change from current assessments Y-OPCT = 0°F Cumulative PCT change from current assessments APCT = 0°F Net PCT 2150°F

Attachment 4 Quad Cities Nuclear Power Station Unit 2 10 CFR 50 .46 Report (Westinghouse Fuel)

PLANT NAME : Quad Cities Unit 2 ECCS EVALUATION MODEL: USA5 REPORT REVISION DATE: 03/28/08 CURRENT OPERATING CYCLE : 20 ANALYSIS OF RECORD Evaluation Model:

"Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," WCAP-16078-P-A, November 2004 .

Calculations :

"Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021QC-LOCA, Revision 3, Westinghouse Electric Company, LLC., April 2007 .

Fuel Analyzed in Calculation: SVEA-96 Optima2 Limiting Fuel Type: SVEA-96 Optima2 Limiting Single Failure : LPCI injection valve Limiting Break Size and Location : 1 .0 double-ended guillotine break in the recirculation pump suction line Reference Peak Cladding Temperature (PCT) PCT = 2150°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS None - new Analysis for OQtima2 fuel APCT = 0°F Net PCT 2150°F B. CURRENT LOCA MODEL ASSESSMENTS New analysis (See Note 9) APCT = 0°F New Reload of SVEA-96 Optima2 Fuel (See Note 10) APCT = 0°F Total PCT change from current assessments J1OPCT = 0°F Cumulative PCT change from current assessments APCT =0-F Net PCT 2150°F

Attachment 5 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 LOCH 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 LOCH 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 0°F.

[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. 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 level/volume table and Steam 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 Page 1 of 4

Attachment 5 Quad Cities Nuclear Power Station Units 1 and 2 10 CFR 50 .46 Report Assessment Notes reload of GE14 fuel in Cycle 18 core. The PCT impact for these errors and reloads of GE14 fuel was determined to be 0°F.

[Reference : Letter from Patrick R. Simpson (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, RS-04-066, dated May 5, 2004 .]

6. Prior LOCH Assessment The referenced letter provided the annual 50 .46 report for Units 1 and 2. This letter reported GE LOCA error due to new heat source of Units 1 & 2 and Quad Cities Unit 1 Cycle 19 with a new reload of GE14 fuel .

[Reference : Letter from Patrick R. Simpson (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, RS-05-056, dated May 5, 2005 .]

7. Prior LOCA Assessment The referenced letter provided the annual 50 .46 report for Units 1 and 2. This letter reported LOCA evaluations for installation of new steam dryers during mid-cycle outages for 01 C19 and Q2C18, respectively . Also, the letter reported Q2C19 startup in April 2006 with the first reload of Westinghouse Optima2 fuel and implementation of the Westinghouse LOCA analysis . Additionally, LOCA evaluations by both GE and Westinghouse were reported for Unit 2 modification to the inlet configuration of the 6" inlet standpipe of eight main steam safety valves and four Electromatic relief valves, which replaced the previously installed inlet pipe and flange with a 6" Tee, flange and an Acoustic Side Branch (ASB).

The PCT impact due to the plant modifications was determined to be 0°F.

(Reference : Letter from Patrick R . Simpson (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, RS-06-064, dated May 5, 2006 .]

8 . Prior LOCA Assessment The referenced letter provided the annual 50 .46 report for Units 1 and 2 . This letter reported GE LOCA evaluation for installation of a modification to the inlet configuration of the 6" inlet standpipe of eight main steam safety valves and four Electromatic relief valves for Quad Cities Unit 1 . Also, this letter reported evaluation of a change in the GE small break analysis assumption for axial power shape and Westinghouse LOCA analysis Hgap correlation input error. The PCT impact due to these changes was determined to be 0° F.

Attachment 5 Quad Cities Nuclear Power Station Units 1 and 2 10 CFR 50.46 Report Assessment Notes

[Reference : Letter from Patrick R. Simpson (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, RS-07-070, dated May 7, 2007 .]

9. Current LOCA Assessment With Quad Cities Power Station Unit 2 Cycle 19 startup in April 2006, Unit 2 implemented a Westinghouse LOCA analysis supporting the transition to Optima2 fuel . A common LOCA analysis was first performed to apply to four units of Dresden and Quad Cities plants . Subsequently, Westinghouse performed a new plant-specific LOCA Analysis for Quad Cities Nuclear Power Station . This new analysis applies to operation of the Westinghouse Optima2 fuel in the Quad Cities Unit 1 and 2 reactors . This analysis applies specific inputs and assumptions in the LOCA calculation approved in the licensed Westinghouse methodology. Included are :

a. Containment back pressure - the amount of containment overpressure credited in accordance with acceptance letter issued by the NRC,

b. Proportional leakage,

c. ECCS temperature reduction,

d. Quad Cities specific ECCS parameters including the ECCS flow and leakages,

e. Diesel generator sequencing specific to Quad Cities, f . Two channel model,

g. Improved definition of end of lower plenum flashing used to terminate non-zero heat transfer coefficient.

The above changes as implemented in the Quad Cities specific LOCH analysis are in compliance with the Westinghouse LOCH methodology. These changes result in the same PCT at less restrictive MAPLHGR limits compared to the original common LOCA analysis . There is no prior or current assessment penalty for the Quad Cities specific LOCA analysis . With the introduction of Optima2 fuel, the limiting PCT for Optima2 as analyzed under the Westinghouse LOCH method is 2150 °F whereas the limiting PCT for GE14 as analyzed under GE LOCA method is 2110 °F.

[References :

(1) "Dresden 2 & 3 and Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021QC-LOCA, Revision 2, September 2006 .

(2) "Quad Cities 1 & 2 LOCH Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TR021 QC-LOCA, Revision 3, Westinghouse Electric Company, LLC. April 2007 .]

Attachment 5 Quad Cities Nuclear Power Station Units 1 and 2 10 CFR 50.46 Report Assessment Notes

10. A second reload of SVEA-96 Optima2 fuel was introduced into the Quad Cities Unit 2 Cycle 20 core . Westinghouse evaluated this change and determined that the impact on the licensing basis PCT to be 0°F.

[Reference : "Quad Cities Nuclear Power Station Unit 2 Cycle 20 MAPLHGR Report," NF-BEX-07-198-NP, Revision 0, January 2008.]

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