RS-06-165, Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for November 4, 2005 Through November 3, 2006

Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for November 4, 2005 Through November 3, 2006

ML063070401
Person / Time
Site:	Clinton
Issue date:	11/03/2006
From:	Nicely K AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-06-165
Download: ML063070401 (5)

Text

AmerGen Energy Company, LLC www.exeloTicorp .com erGen An Exelon Company 4300 Winfield Road Warrenville, f L 60555 RS-06-165 November 3, 2006 10 CFR 50.46 U . S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, D.C . 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," paragraph (a)(3)(ii), AmerGen Energy Company, LLC (AmerGen) is submitting the annual report of the Emergency Core Cooling System (ECCS)

Evaluation Model changes and errors for Clinton Power Station (CPS), Unit 1 . This report covers the period from November 4, 2005 through November 3, 2006.

Should you have any questions concerning this letter, please contact Mr. Timothy A. Byam at (630) 657-2804 .

Respectfully, 911-1.

Kenneth M . Nicely Manager - Licensing Attachments:

1 . 10 CFR 50.46 Report 2 . 10 CFR 50.46 Report Assessment Notes

Attachment 1 Clinton Power Station Unit 1 10 CFR 50.46 Report Page 1 of 2 PLANT NAME: Clinton Power Station, Unit 1 ECCS EVALUATION MODEL: SAFER/GESTR - LOCA REPORT REVISION DATE : 11/03/06 CURRENT OPERATING CYCLE : 11 ANALYSIS OF RECORD Evaluation Model Methodology: The GESTR-LOCA and SAFER Models for the Evaluation of the Lass-of-Coolant Accident ;

Volume III, SAFER/GESTR Application Methodology, NEDC-23785-1-PA, Revision 1, General Electric Company, October 1984 .

Calculation : Clinton Power Station, SAFE R/GESTR-LOCA Analysis Basis Documentation, NEDC-32974P, GE Nuclear Energy, October 2000.

Fuel : GE 14 Limiting Fuel : GE 14 Limiting Single Failure: High Pressure Core Spray (HPCS) Diesel Generator Limiting Break Size and Location : 1 .0 Double Ended Guillotine of Recirculation Pump Suction Piping Reference Peak Cladding Temperature (PCT): 1550°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated November 13, 2000 (See Note 1) APCT = 0°F 10 CFR 50.46 report dated November 08, 2001 (See Note 2) APCT = 5°F 10 CFR 50.46 report dated November 05, 2002 (See Note 3) APCT = 35°F 10 CFR 50.46 report dated November 05, 2003 (See Note 4) APCT = 5°F 10 CFR 50 .46 report dated November 05, 2004 (See Note 5) APCT = 0°F 10 CFR 50.46 report dated November 04, 2005 (See Note 6) APCT = 0°F Net PCT 1595°F

Attachment 1 Clinton Power Station Unit 1 10 CFR 50.46 Report Page 2 of 2 B. CURRENT LOCA MODEL ASSESSMENTS impact of Top Peaked Power Shape for Small Break LOCA (Note 7) AFP(;T = O'F Cumulative PCT change from current assessments APCT I =OOF -

Net PCT 1595OF

Attachment 2 Clinton Power Station Unit 1 10 CFR 50.46 Report Assessment Notes Page 1 of 2 NOTES:

1 . Prior LOCA Model Assessments The referenced letter reported a new analysis of record for Clinton Power Station (CPS) .

[Reference : Letter from M. A. Reandeau (AmerGen Energy Company) to U .S.

NRC, "Report of a Change to the ECCS Evaluation Model Used for Clinton Power Station (CPS)," dated November 13, 2000 .]

2. Prior LOCA Model Assessments An inconsistent core exit steam flow was used in the pressure calculation in the SAFER code then there is a change in the two-phase level . The incorrect calculated pressure may result in premature termination of ECCS condensation and will impact the second peak clad temperature (PCT). GE evaluated the impact of this error and determined that the impact is an increase of 50 F in the PCT . This error was reported to the NRC in the referenced letter .

[Reference : Letter from K. A. Ringer (Exelon Generation Company) to U .S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station," dated November 8, 2001 .]

3. Prior LOCA Model Assessments In the referenced letter to the NRC, the impact of the Low Pressure Coolant Injection (LPCI) and Low Pressure Core Spray (LPCS) minimum flow valve flow diversion was reported and was found to have a 0"F impact . Also in the referenced letter GE LOCA errors were reported all of which had a O'F PCT increase except for a SAFER Core Spray sparger injection elevation error that resulted in a 150F increase in the PCT. The Extended Power Uprate (EPU) has resulted in an increase of 20OF in the PCT. The EPU was implemented in Cycle 9 Reload .

[Reference : Letter from Patrick R . Simpson (Exelon Generation Company) to U .S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station," dated November 5, 2002.]

4. Prior LOCA Model Assessments In the referenced letter to the NRC, the impact of an error found in the initial level/volume table for SAFER was reported . 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 . This error resulted in a 50F increase in the PCT for all fuel types (i .e., GE 10 & GE14) .

[Reference : Letter from Patrick R. Simpson (Exelon Generation Company) to U.S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station," dated November 5, 2003.]

Attachment 2 Clinton Power Station Unit 1 10 CFR 50.46 Report Assessment Notes Page 2 of 2

5. Prior LOCA Model Assessments In the referenced letter to the NRC, the impact of a GE postulated new heat source applicable to the LOCA event was reported . This heat source is due to recombination of hydrogen and excess oxygen drawn into the vessel from containment during core heatup. The PCT impact for all fuel types was OOF and the effect on local oxidation was negligible .

[Reference : Letter from Patrick R. Simpson (Exelon Generation Company) to U.S . NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station," dated November 5, 2004.]

6. Prior LOCA Model Assessments In the referenced letter to the NRC, the impact of the 24-month cycle operation was reported . The evaluation determined that the LOCA analysis of record was performed with bounding assumptions and hence is not impacted with the 24-month cycle. A zero degree PCT impact is assigned .

[Reference : Letter from Patrick R. Simpson (Exelon Generation Company) to U .S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station," dated November 4, 2005.]

7. Current LOCA Model Assessments GE reported that past small break ECCS-LOCA analyses have assumed a mid-peaked power shape, consistent with DBA break analyses . Recently GE has determined that for small break cases, a top peak axial power shape can result in higher calculated PCT.

Evaluations have been performed on representative plants spanning all BWR plant types. The impact on the licensing basis PCT is reported as OOF for GE 14 Fuel for CPS .

[Reference : Exelon Clinton Power Station, GE 10CFR 50.46 Notification Letter 2006-01, July 28, 2006.]