ML041140033

Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report

ML041140033
Person / Time
Site:	LaSalle
Issue date:	03/09/2004
From:	Landahl S Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML041140033 (6)

Text

Exelon Generation Company, LLC www.exeloncorp.com NEuexaT LaSalle County Station 2601 North 21"Road Marseilles, IL 61341-9757 March 9, 2004 10 CFR 50.46 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Subject:

Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report

Reference:

Letter from S. R. Landahl (Exelon Generation Company, LLC) to U. S. NRC, "Plant Specific ECCS Evaluation Changes -

10 CFR 50.46 Report," dated June 9, 2003.

In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Exelon Generation Company (EGC), LLC, submits the enclosed attachments to fulfill the 30-day and annual reporting requirements for LaSalle County Station, Units 1 and 2. In the referenced letter, EGC reported the fuel peak cladding temperatures (PCTs) calculated based on an acceptable model to be 13010 F for General Electric (GE) fuel and 18070 F for Framatome ANP (FANP) fuel. The referenced letter also estimated the effects of changes that occurred since the PCT values had been calculated using an acceptable model. As earlier reported these changes resulted in an increase of 25*F to 1832 0F for FANP Fuel. The previous GE analysis supported a fuel type no longer in use for power operation at LaSalle County Station, Units 1 and 2, and GE14 fuel has been introduced into Unit 1. As a result, a new analysis was performed for the GE fuel in Unit 1. Based on the new analysis, the PCT for GE fuel increased to a value of 1380'F. This is a change of 790F from the last evaluation using an acceptable model.

Unit 1 employs a mixed core design containing co-resident GE and FANP fuel.

Unit 2 employs a core design containing only FANP fuel. The Loss of Coolant Accident (LOCA) analyses of record for both GE and FANP fuel are within all of the acceptance criteria set forth in 10 CFR 50.46.

Attachments 1 and 2 provide PCT information for the limiting LOCA evaluations for LaSalle County Station, Units 1 and 2, including all assessments as of February 20, 2004. The assessment notes are contained in Attachment 3 and provide a detailed description for each change or error reported.

fNbeo\

U. S. Nuclear Regulatory Commission March 11, 2004 Page 2 Should you have any questions concerning this letter, please contact Mr. Glen Kaegi, Regulatory Assurance Manager, at (815) 415-2800.

Respectfully, Susan R. Landahl Plant Manager LaSalle County Station Attachments cc: Regional Administrator - NRC Region IlIl NRC Senior Resident Inspector - LaSalle County Station

Attachment I LaSalle Units 1 10 CFR 50.46 Report (GE Fuel)

PLANT NAME: LaSalle Unit 1 ECCS EVALUATION MODEL: SAFER/GESTR LOCA REPORT REVISION DATE: February 20, 2004 CURRENT OPERATING CYCLES: LIC11 ANALYSIS OF RECORD Evaluation Model Methodology. NEDE-23785-1-PA, Rev. 1, "GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident (Volume l1l), SAFER/GESTR Application Methodology " ,

October 1984.

Calculation: Project Task Report, Exelon LaSalle Unit I SAFERIGESTR Loss-of-Coolant Accident Analysis for GE 14 Fuel," GE report number GE-NE-0000-0022-8684-RO, dated December 2003.

Fuel: GE14 Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Double Ended Guillotine of Recirculation Pump Suction Location: Piping Reference PCT: 1380"F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS Complete Break spectrum analysis was performed In December 2003 for the APCT = 0 'F introduction of GE 14 and all errors were addressed. .

Net PCT 1380 "F B. CURRENT LOCA MODEL ASSESSMENTS Unit I Jet Pump Riser Leakage, LPCS and HPCS Leakge and Displacement APCT = 0 OF of Water (Note 1)

Total PCT Change from Current Assessments APCT= O "F Cumulative PCT Change from Current Assessments E l APCT I = 0 ¶F Net PCT 1380 OF

Attachment 2 LaSalle Units 1 and 2 10 CFR 50.46 Report (FANP Fuel)

PLANT NAME: LaSalle Units I and 2 ECCS EVALUATION MODEL: EXEM BWR Evaluation Model REPORT REVISION DATE: February 20, 2004 CURRENT OPERATING CYCLE: LIC10 and L2C10 ANALYSIS OF RECORD Evaluation Model Methodology: Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF 048(P)(A), January 1993.

BWR Jet Pump Model Revision for RELAX, ANF 048(P)(A), Supplement I and Supplement 2, Siemens Power Corporation, October 1997.

Calculation: 1. LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM Th -9B Fuel, EMF-2175(P), March 1999.

2. LOCA Break Spectrum Analysis for LaSalle Units I and 2, EMF-2174(P), March 1999.

3. LaSalle Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUMTm-10 Fuel, EMF-2641(P), November 2001.

4. LaSalle Units I and 2 LOCA Break Spectrum Analysis for ATRIUMT'-10 Fuel, EMF-2639(P), November 2001.

Fuel: ATRIUM T'-9Band ATRIUM"m-10 Limiting Fuel ATRIUMnm-9B Limiting Single Failure: HPCS Diesel Generator Limiting Break Size and Location: 1.1 ft2 Recirculation Pump Discharge Side Line Break Reference PCT: 1807 OF MARGIN ALLOCATION A 001nl2 I AflA Mfl~Fl~i AC=QR=1-10 CFR 50.46 report dated May 7, 1999 (See Note 2) APCT = 0 °F 10 CFR 50.46 report dated February 9, 2000 (See Note 3) APCT =18 °F 10 CFR 50.46 report dated June 12, 2000 (See Note 4) APCT = 0 OF 10 CFR 50.46 report dated June 8, 2001 (See Note 5) APCT = 0 OF 10 CFR 50.46 report dated June 8, 2002 (See Note 6) APCT = 2 °F 10 CFR 50.46 report dated June 9, 2003 (See Note 7) APCT = 5 °F Net PCT 1832 OF B. CURRENT LOCA MODEL ASSESSMENTS Data Transfer from PREHUXY to HUXY (8) APCT = 0OF Unit 1 Jet Pump Riser Leakage, LPCS and HPCS Leakge and Displacement APCT = 0 OF of Water (Note 1)

Total PCT Change from Current Assessments FAPCT = 0 OF Cumulative PCT Change from Current Assessments I APCT = 0OF Net PCT 1832 OF

Attachment 3 LaSalle Units I and 2 10 CFR 50.46 Report Assessment Notes

1. Jet Pump Riser Leakage, LPCS/HPCS Leakage and Displacement of Water for Unit I During the startup of LaSalle Unit 1 Cycle 11 several evaluations were performed as documented In the Exelon Engineering Evaluation (EC) process. The net results of these evaluations were that there was a zero degree PCT Impact.

[

References:

(i) EC 347217, Evaluate the Impact of the Jet Pump Riser Leakage for the LaSalle Unit 1 GE and the Framatome LOCA Analyses. (ii) EC 346839, Impact of the LPCS and HPCS Leakage on the LaSalle Unit I GE and Framatome LOCA Analyses.

(iii) EC 346969, Evalaute the Impact on the Safety Analyses due to 10 Gallons of Water Displaced at LaSalle Unit I due to the Installation of Wedges and Clamps to Repair the Jet Pumps.]

2. Prior LOCA Model Assessment for FANP fuel The May 1999 LOCA model assessment was a new analysis of record for Framatome (Formerly Siemens) due to the Introduction of ATRIUM-9B fuel Into the Unit 2 Cycle 8 core. Therefore, there is no PCT change. Analysis was performed for a core power of 3722 MWt that bounds the current uprated power of 3489 MWL

[

Reference:

Letter from J. A. Benjamin (ComEd) to U.S. NRC, 'Report of Significant Change In Calculated Peak Cladding Temperature (PCT) - IOCFR 50.46 Report," dated May 7, 1999.]

3. Prior LOCA Model Assessment for FANP fuel The February 2000 50.46 report assessed the impact of errors In the LOCA evaluation model.

[

Reference:

Letter from J. A. Benjamin (ComEd) to U.S. NRC, uPlant Specific ECCS Evaluation Changes - 10CFR 50.46 Report, dated February 9, 2000.]

4. Prior LOCA Model Assessment for FANP fuel The June 2000 10 CFR 50.46 report does not have any PCT assessment for ATRIUM-9B fuel.

[

Reference:

Letter from C. G. Pardee (ComEd) to U.S. NRC, 'Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report, dated June 12, 2000.]

5. Prior LOCA model assessment for FANP fuel The reference letter assessed Impact of Unit 2 LPCS riser leakage, errors in FANP LOCA analysis model and Unit 2 Cycle 9 reload fuel.

[

Reference:

Letter from M. A Schiavoni (Exelon) to U.S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 8, 2001.]

Attachment 3 LaSalle Units I and 2 10 CFR 50.46 Report Assessment Notes

6. Prior LOCA model assessment for FANP fuel The referenced letter assessed Impact of errors In FANP LOCA analysis model, Unit I Cycle 10 reload fuel and ATRIUM-9B exposure extension.

[

Reference:

Letter from M. A Schiavoni (Exelon) to U.S. NRC, nPIant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 8, 2002.]

7. Prior LOCA model assessment for FANP fuel The June 2003 50.46 report assessed the impact of errors in the LOCA evaluation, Unit 2 jet pump leakage, Unit 2 Cycle 10 reload Fuel and the Unit 1 mid-cycle reload.

[

Reference:

Letter from Susan R Landahl (Exelon) to U.S. NRC, OPIant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated June 9, 2003.]

8. PREHUXY to HUXY data Error FANP reported that a problem may occur In the transfer of RELAX coolant temperature data from PREHUXY to HUXY at the time of core spray. This can cause an error in the application of the HUXY quenching model. FANP determined that the Impact on the limiting break spectrums results is zero degree.

[

Reference:

10 CFR 50.46 PCT Reporting for LaSalle Units and Transmittal of CR 11130, Letter AWW:04:003, Issued by A. W. Will, January 9, 2004.]