ML073110137

Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report

ML073110137
Person / Time
Site:	Dresden
Issue date:	10/31/2007
From:	Bost D Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-0049
Download: ML073110137 (8)

Text

Exelkn.

Exelon Generation Company, LLC www.exeloncorp.com Nuclear Dresden Nuclear Power Station 6500 North Dresden Road Morris, IL 60450-9765 10 CFR 50.46(a)(3)(ii)

October 31, 2007 SVPLTR: #07-0049 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249

Subject:

Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report

Reference:

Letter from D. Bost (Exelon Generation Company, LLC) to U. S. NRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," dated November 9, 2006 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 Company LLC, is submitting this letter and its attachment to meet the annual reporting requirements.

Dresden Nuclear Power Station (DNPS) has maintained the same emergency core cooling (ECCS) model as reported in the referenced letter for Unit 2 and GE14 fuel in Unit 3. For Unit 3, the Westinghouse Loss of Coolant Accident (LOCA) model has been implemented to support the transition to Optima2 fuel. No vendor 10 CFR 50.46 LOCA model change/error notifications were received since the last annual report. The attachment provides the PCT value for each unit and the "rack-up" sheets for the LOCA analyses, along with assessment note summaries.

If there are any questions concerning this letter, please contact Mr. James Ellis at (815) 416-2800.

Respectfully, Danny Bo t Site Vi President Dres en Nuclear Power Station

Attachment:

Dresden Nuclear Power Station Units 2 and 3 - 10 CFR 50.46 Report cc:

Regional Administrator - NRC Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station

DRESDEN NUCLEAR POWER STATION UNITS 2 AND 3 10 CFR 50.46 REPORT

Dresden Nuclear Power Station Units 2 and 3 10 CFR 50.46 Report Unit 2 PLANT NAME:

ECCS EVALUATION MODEL:

REPORT REVISION DATE:

CURRENT OPERATING CYCLE:

Dresden Nuclear Power Station, Unit 2 SAFER/GESTR-LOCA 09/26/2007 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:

"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: 9x9-2, 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, 2001 (See Note 1)

APCT = 0°F 10 CFR 50.46 report dated November 25, 2002 (See Note 2)

APCT = 0°F 10 CFR 50.46 report dated November 25, 2003 (See Note 3)

APCT = 0°F 10 CFR 50.46 report dated November 24, 2004 (See Note 4)

APCT = 0°F 10 CFR 50.46 report dated November 16, 2005 (See Note 5)

APCT = 00F 10 CFR 50.46 report dated November 9, 2006 (See Note 6)

APCT = 0°F Net PCT 2110 OF B. CURRENT LOCA MODEL ASSESSMENTS None APCT = 0°F Total PCT change from current assessments Y-APCT = 0°F Cumulative PCT change from current assessments YZ I APCT I = 0°F Net PCT 2110 OF

Dresden Nuclear Power Station Units 2 and 3 10 bFR 50.46 Report Unit 3 GE Fuel PLANT NAME:

ECCS EVALUATION MODEL:

REPORT REVISION DATE:

CURRENT OPERATING CYCLE:

Dresden Nuclear Power Station, Unit 3 SAFER/GESTR-LOCA 09/26/2007 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:

"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: 9x9-2, 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 November 25, 2002 (See Note 2)

APCT = 0°F 10 CFR 50.46 report dated November 25, 2003 (See Note 3)

APCT = 0°F 10 CFR 50.46 report dated November 24, 2004 (See Note 4)

APCT = 0°F 10 CFR 50.46 report dated November 16, 2005 (See Note 5)

APCT = 0°F 10 CFR 50.46 report dated November 9, 2006 (See Note 6)

APCT = 0°F Net PCT 2110 OF B. CURRENT LOCA MODEL ASSESSMENTS None APCT = 0°F Total PCT change from current assessments Y-APCT = 0°F Cumulative PCT change from current assessments YZ I APCT I= 0-F Net PCT 2110 OF

Dresden Nuclear Power Station Units 2 and 3 10 CFR 50.46 Report Unit 3 Westinghouse Fuel PLANT NAME:.

'Dresden Nuclear Power Station, Unit 3 ECCS EVALUATION MODEL:

1USA5 REPORT REVISION DATE:

09/26/2007 CURRENT OPERATING CYCLE:

20 ANALYSIS OF RECORD Evaluation Model:

Calculations:

'Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Quallfication and Application to SVEA-96 Optima2 Fuel,"

WCAP-1 6078-P-A, November 2004.

"Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TRO21 DR-LOCA, Revision 2, Westinghouse Electric Company LLC, June 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)

MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS PCT = 2150'F None - new LOCA analysis APCT = 0OF]

PCT 2150OFI B. CURRENT LOCA MODEL ASSESSMENTS None - new analysis (See Note 7)

APCT = 0°F Total PCT change from current assessments ZAPCT = 0°F Cumulative PCT change from current assessments Y-APCT I = 0°F PCT 2150°F

Dresden Nuclear Power Station Units 2 and 3 10 CFR 50.46 Report Assessment Notes

1. Prior LOCA Model Assessment The 10 CFR 50.46 letter dated December 6, 2001 reported a new LOCA analysis to support extended power uprate (EPU) and transition to GE14 fuel for Dresden Unit 2 Cycle 18. The same report assessed impact of errors in Framatome ANP LOCA analysis model for Dresden Unit 3 Cycle 17 at pre-EPU power level.

[

Reference:

Letter from Preston Swafford (PSLTR: #01-0122) (Exelon) to USNRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," December 6, 2001.]

2. Prior LOCA Model Assessment!

Unit 3 implemented GE LOCA analysis and GE14 fuel with Dresden Unit 3 Cycle 18 startup on October 25, 2002. Therefore, both Dresden Units 2 and 3 are being maintained under the same LOCA analysis. In the referenced letter, the impact of GE LOCA error in the WEVOL code was reported foe Dresden Units 2 and 3 and determined to be negligible.

[

Reference:

Letter from Robert J. Hovey (RHLTR: #02-0083) (Exelon) to USNRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," November 25, 2002.]

3. Prior LOCA Model Assessment The annual 10 CFR 50.46 report provided information on the LOCA model assessments for SAFER Level/Volume table error and Steam Separator pressure drop error. In the referenced letter, the impact of these two GE LOCA errors were reported to be negligible.

I

[

Reference:

Letter from Robert J. Hovey (RHLTR: #03-0077) (Exelon) to USNRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," November 25, 2003.]

4. Prior LOCA Model Assessment The referenced annual 10 CFR 50.46 report provided information on reload of GE14 fuel for Dresden Unit 2 Cycle 19 and impact of postulated hydrogen-oxygen recombination on PCT. GE determined that there is no PCT impact because of the change due to the new reload of GE14 fuel and the postulated hydrogen -oxygen recombination.

[

Reference:

Letter from Danny Bost (SVPLTR: #04-0075) (Exelon) to USNRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," November 24, 2004.]

Dresden Nuclear Power Station Units 2 and 3 10 CFR 50.46 Report Assessment Notes

5. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for Units 2 and 3. The letter reported the PCT impact of reload of GE14 fuel for D3C19 starting on December 8, 2004.

Also, the letter reported the GE L,OCA evaluation for Unit 3, which implemented the lower sectional replacement and T-box-clamp repairs. GE determined that there is no PCT impact because of the change due to the new reload of GE14 fuel and the lower sectional replacement and T-box clamp repairs.

[

Reference:

Letter from Danny Bost (SVPLTR: #05-0044) (Exelon) to USNRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," November 16, 2005.]

6. Prior LOCA Model Assessment The referenced letter provided the annual 10 CFR 50.46 report for Units 2 and 3. The letter reported the PCT impact of the reload of GE14 fuel for D2C20. The letter also reported an evaluation of increased leakage of less than 5 gpm at runout condition in core spray line flow due to crack growth identified during D2R19 outage.. Additionally, a GE evaluation of the small break for impact due to top-peak axial power shape was reported in this letter.

The impact due to these changes on the licensing basis PCT was reported as zero.

[

Reference:

"Letter from Danny Bost (SVPLTR: #06-0054) (Exelon) to USNRC, "Plant Specific ECCS Evaluation Changes - 10 CFR 50.46 Report," November 9, 2006.]

7.

Current LOCA Model Assessment With Dresden Power Station Unit 3 Cycle 20 startup in November 2006, Unit 3 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 Dresden Nuclear Power Station. This new analysis applies to operation of the Westinghouse Optima2 fuel in the Dresden reactor. 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 ECCS leakage,

c. ECCS temperature reduction,

d. Plant-specific ECCS parameters including the ECCS flow and leakages specific to

Dresden,

e. Emergency Diesel Generator load sequencing time delays specific to Dresden,

f.

Two channel model,

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

Dresden Nuclear Power Station Units 2 and 3 10 CFR 50.46 Report Assessment Notes The above changes as implemented in the Dresden specific LOCA analysis are in compliance with the Westinghouse LOCA 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 Dresden specific LOCA analysis.

With the introduction of Optima2,fuel, the limiting PCT for Optima2 as analyzed under the Westinghouse LOCA method is 2150 OF whereas the limiting PCT for GE14 as analyzed under GE LOCA method is 2110 OF.

[

References:

(1) "Dresden 2 & 3 and Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel,"

OPTIMA2-TRO21 DR-LOCA, Revision 1, September 2006.

(2) "Dresden 2 & 3 LOCA Analysis for SVEA-96 Optima2 Fuel," OPTIMA2-TRO21 DR-LOCA, Revision 2, Westinghouse Electric Company LLC, June 2007.]