RS-10-162, Submittal of Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors

Submittal of Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors

ML103020228
Person / Time
Site:	Clinton
Issue date:	10/29/2010
From:	Hansen J Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-10-162
Download: ML103020228 (6)

Text

Exelon Generation www.exeloncorp.com Exelvna 4300 Winfield Road Warrenville, IL 60555 Nuclear RS-10-162 10 CFR 50.46 October 29, 2010 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 Change s

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), Exelon Generation Compa ny, LLC (EGC) is submitting the annual report of the Emergency Core Cooling System (ECCS)

Evaluation Model changes and errors for Clinton Power Station, Unit 1 (CPS).

This report covers the period from October 31, 2009 through October 29, 2010.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Mitchel A. Mathews at (630) 657-28 19.

Respectfully, JeffreiL. Hansen Manager - Licensing and Regulatory Affairs 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: 10/29/10 CURRENT OPERATING CYCLE: 13 ANALYSIS OF RECORD Evaluation Model Methodology: The GESTR -LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident; Volume III, SAFER/GESTR Application Methodology, NEDC-23785 PA, Revision 1, General Electric Company, October 1984.

Calculation: Clinton Power Station, SAFE R/G ESTR-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

Attachment 1 Clinton Power Station, Unit 1 10 CFR 50.46 Report Page 2 of 2 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 10 CFR 50.46 report dated November 03, 2006 (See Note 7)

LPCT = 0°F 10 CFR 50.46 report dated April 19, 2007 (See Note 8)

APCT = 6°F 10 CFR 50.46 report dated November 02, 2007 (See Note 9)

APCT = 0°F 10 CFR 50.46 report dated October 31, 2008 (See Note 10)

APCT = 0°F 10 CFR 50.46 report dated October 30, 2009 (See Note 11)

APCT = 0°F Net PCT 1601 OF B. CURRENT LOCA MODEL ASSESSMENTS None (Note 12)

APCT = 0°F Total PCT change from current assessments Y-APCT = 0°F Cumulative PCT change from current assessments APCT I = 0°F Net PCT 1601 °F

Attachment 2 Clinton Power Station, Unit 1 10 CFR 50.46 Report Assessment Notes Page 1 of 3 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 Powe r Station (CPS),"

dated November 13, 2000.]

2. Prior LOCA Model Assessments An inconsistent core exit steam flow was used in the pressure calcul ation in the SAFER code when there is a change in the two-phase level. The incorrect calcul ated 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 5°F in the PCT. This error was reported to the NRC in the referenced letter.

[

Reference:

Letter from K. A. Ainger (Exelon Generation Company) to U.S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Chang es 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 divers ion 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 0°F PCT increase except for a SAFER Core Spray sparge r injection elevation error that resulted in a 15°F increase in the PCT. The Extended Power Uprate (EPU) has resulted in an increase of 20°F in the PCT. The EPU was implemented in Cycle 9 Reload.

[

Reference:

Letter from Patrick R. Simpson (Exelon Generation Comp any) to U.S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Mode l 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 5°F increase in the PCT for all fuel types (i.e., GE 10 & GE14).

[

Reference:

Letter from Patrick R. Simpson (Exelon Generation Comp any) to U.S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Mode l 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 3

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 recom bination of hydrogen and excess oxygen drawn into the vessel from containment during core heatup. The PCT impact for all fuel types was 0°F and the effect on local oxidation was neglig ible.

[

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 0°F PCT impact was 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. Prior LOCA Model Assessments In the referenced letter to the NRC, the impact of the top peak axial power shape on the small break LOCA was reported. The impact of the top peak axial power shape on the licensing basis PCT was 0 °F for GE 14 Fuel for CPS.

[

Reference:

Letter from Kenneth M. Nicely (Exelon Generation Company) to U.S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Chang es and Errors for Clinton Power Station," dated November 3, 2006.]

8. Prior LOCA Model Assessments In the referenced letter, the impact of the core shroud repair on the PCT was reported to the NRC. The leakage flows through the repair holes result in slightl y increased time to core recovery, following core uncovery. The effect has been conservativ ely assessed to increase the PCT for the limiting LOCA by less than 6 °F.

[

Reference:

Letter from Patrick R. Simpson (Exelon Generation Company) to U.S. NRC, "Updated Annual Report of Emergency Core Cooling System Evalu ation Model Changes and Errors for Clinton Power Station," dated April 19, 2007.]

Attachment 2 Clinton Power Station, Unit 1 10 CFR 50.46 Report Assessment Notes Page 3 of 3

9. Prior LOCA Model Assessments In the referenced letter, Exelon submitted to the NRC the annual 10CFR 50.46 report for 2007.

There was no LOCA model assessment for the Clinton LOCA analys is.

[

Reference:

Letter from Jeffrey L. Hansen (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 2, 2007.]

10. Prior LOCA Model Assessments In the referenced letter, Exelon submitted to the NRC the annua l 10CFR 50.46 report for 2008.

There was no LOCA model assessment for the Clinton LOCA analys is.

[

Reference:

Letter from Jeffrey L. Hansen (Exelon Generation Company) to U.S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station," dated October 31, 2008.]

11. Prior LOCA Model Assessments In the referenced letter, Exelon submitted to the NRC the annual 10CFR 50.46 report for 2009.

There was no LOCA model assessment for the Clinton LOCA analys is.

[

Reference:

Letter from Jeffrey L. Hansen (Exelon Generation Company) to U.S. NRC, "Annual Report of Emergency Core Cooling System Evaluation Model Changes and Errors for Clinton Power Station," dated October 30, 2009.]

12. Current LOCA Model Assessments Since the last annual report (see Note 11), no vendor notifications of Emerg System (ECCS) model error/changes that are applicable to Clinton have ency Core Cooling been issued. No ECCS-related changes or modifications have occurred at Clinton that affects the assumptions of the ECCS analyses. Eight Isotope Test Assemblies (GE14i ITA) were loaded in to the cycle 13 core, all the GE14 10CFR50.46 errors are applicable to the GE14i ITA and the PCT for GE14 fuel remain applicable for GE14i ITAs.