RS-07-070, Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report

Quad Cities Nuclear Power Station, Units 1 and 2, 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, Annual Report

ML071270677
Person / Time
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Issue date:	05/07/2007
From:	Simpson P R Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, NRC/NRR/ADRO
References
RS-07-070
Download: ML071270677 (7)

Text

RS-07-070 May 7, 2007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

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 5, 2006 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. David Gullott at (630) 657-2819. Respectfully, P6 Patrick R. Simpson Manager - Licensing 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 1 4w) VVVVw exel o? I CUT p. CC YY1 Exel6n Nuclear 10 CFR 50.46 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 (GE Fuel) Attachment C: Quad Cities Nuclear Power Station Unit 2, 10 CFR 50.46 Report (Westinghouse Fuel) Attachment D: Quad Cities Nuclear Power Station Units 1 and 2, 10 CFR 50.46 Report Assessment Notes ANALYSIS OF RECORD Evaluation Model: Attachment A Quad Cities Nuclear Power Station Unit 1 10 CFR 50.46 Report PLANT NAME: Quad Cities Unit 1 ECCS EVALUATION MODEL: SAFER/GESTR-LOCA REPORT REVISION DATE: 05/07/07 CURRENT OPERATING CYCLE: 19 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

"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 = 2110°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS B. CURRENT LOCA MODEL ASSESSMENTS Acoustic Side Branch See Note 8 APCT = 0°F Axial Power Shape Impact on Small Break LOCA See Note 9 APCT = 0°F Total PCT change from current assessments EAPCT = 0°F Cumulative PCT change from current assessments E APCT = 0°F Net PCT 2110 OF 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 Net PCT 2110 ° F PLANT NAME: Quad Cities Unit 2 ECCS EVALUATION MODEL: SAFER/GESTR-LOCA REPORT REVISION DATE: 05/07/07 CURRENT OPERATING CYCLE: 19 ANALYSIS OF RECORD Evaluation Model: Calculations

Attachment B Quad Cities Nuclear Power Station Unit 2 10 CFR 50.46 Report (GE Fuel) 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. "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 B. CURRENT 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) OPCT = 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 Net PCT 2110°F Axial Power Sha" z 4m u t act on Small Break LOCA See Note 9 APCT = 0°F Total PCT change from current assessments Y_APCT = 0°F Cumulative PCT change from current assessments Y APCT = 0°F Net PCT 2110°F PLANT NAME: ECCS EVALUATION MODEL: REPORT REVISION DATE: CURRENT OPERATING CYCLE: 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

"Task Report for TSD DOW04-21 Final LOCA Analysis for Quad Cities 1 & 2 and Dresden 2 3," NF-BEX-06-44-P, Westinghouse Electric Company, LLC. April 2006." 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 B. CURRENT LOCA MODEL ASSESSMENTS Attachment C Quad Cities Nuclear Power Station Unit 2 10 CFR 50.46 Report (Westinghouse Fuel) Quad Cities Unit 2 USA5 05/07/07 19 10 CFR 50.46 Report dated May 5, 2006 (See Note 7) Net PCT APCT =.0 0 F 2150O F Hgap Correlation Input Error (See Note 10) APCT = 0°F Total PCT change from current assessments Y-OPCT = 0°F Cumulative PCT change from current assessments E APCT = 0°F Net PCT 2150°F Attachment D 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," 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," 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," 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 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," 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 Attachment D Quad Cities Nuclear Power Station Units 1 and 2 10 CFR 50.46 Report Assessment Notes 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 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," dated May 5, 2004] 6. Prior LOCA 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," 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 Q1 C19 and Q2C18, respectively. Also, the letter reported Q2C19 startup in April 2006 with the first reload of Westinghouse Optimal 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," dated May 5, 2006] 8. Current LOCA Assessment During a mid-cycle outage, in May 2006, Quad Cities Unit 1 installed a modification to the inlet configuration of the 6" inlet standpipe of eight main steam safety valves and four Electromatic relief valves. The modification replaces the previously installed inlet pipe and flange with a 6" Tee, flange and an Acoustic Side Branch (ASB). The ASB consists of 6" pipe filled with metal screen material. GE performed an evaluation on the effects of this modification

the results showed that the licensing basis PCT was unaffected, thus the PCT impact was zero.

Attachment D Quad Cities Nuclear Power Station Units 1 and 2 10 CFR 50.46 Report Assessment Notes 9. Current LOCA Assessment Past GE small break analysis assumed a mid-peaked power shape consistent with the GE methodology. Recently, GE has determined that for small break cases, a top-peaked power shape can result in higher calculated PCT than a mid-peaked shape. GE has implemented a change in performing analysis of small break LOCA cases, which requires considering both mid-peaked as well as top-peaked power shapes. GE has evaluated effect of this change for Quad Cities LOCA analysis and determined the impact on the licensing basis PCT is zero. This is because the limiting PCT is based on the DBA large break and the revised small break PCT remains below the limiting large break. Therefore, there is no impact on the limiting large break PCT and as a result licensing basis PCT remains unaffected. 10. Current LOCA Assessment In the referenced letter, Westinghouse has identified Hgap correlation input error. Westinghouse has assessed the error and determined that the BWR LOCA evaluation model uses a bounding Hgap value. Thus, the effect of this error on peak cladding temperature (PCT) is determined to be zero.