RS-07-154, Plant Specific ECCS Evaluation Change - 10 CFR 50.46 Report

Plant Specific ECCS Evaluation Change - 10 CFR 50.46 Report

ML073230758
Person / Time
Site:	Braidwood
Issue date:	11/19/2007
From:	Simpson P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-07-154
Download: ML073230758 (6)

Text

RS-07-154 10 CFR 50.46 November 19, 2007 U. S. Nuclear Regulatory Commission ATTN : Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No. NPF-72 NRC Docket No. STN 50-456

Subject:

Plant Specific ECCS Evaluation Change - 10 CFR 50 .46 Report References : 1 . Letter from P. R. Simpson (Exelon Generation Company, LLC) to U.S. NRC, "Annual Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Required by 10 CFR 50.46, 'Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors,"'

dated April 13, 2007

2. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U . S. NRC, "ECCS Evaluation Model Error - 10 CFR 50.46 Report " dated June 22, 2007 In accordance with 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," Exelon Generation Company, LLC (EGC) is submitting the attached information to fulfill the 30-day reporting requirement for Braidwood Station Unit 1 .

In the Reference 1 letter, EGC reported the Braidwood Station Unit 1 fuel peak cladding temperatures (PCTs), calculated based on an acceptable evaluation model, were 1659.0°F and 2141 .OOF for the small break (SBLOCA) and large break loss-of-coolant accidents (LBLOCA),

respectively .

The Reference 2 letter reported an error in the Emergency Core Cooling System (ECCS) evaluation model (i.e., the HOTSPOT computer code) for LBLOCA that resulted in an increase to the calculated PCT for Braidwood Station Unit 1 and 2 and Byron Station Unit 1 and 2 . This change resulted in an increase in PCT of 20OF establishing a new Braidwood Station Unit 1 LBLOCA PCT value of 2161 .0'F . The SBLOCA PCT remained at 1659 .OOF.

November 19, 2007 U . S. Nuclear Regulatory Commission Page 2 Subsequent to the Reference 2 submittal, EGG evaluated the PCT changes for the SIBLOCA as a result of Generic Safety Issue 191, "Assessment of Debris Accumulation on PWR Sump Performance," modifications that were made during the recent Fall 2007 Braidwood Station Unit 1 refueling outage. The modifications involved replacement of Safety Injection throttle valves, of which caused EGGS flow changes. As a result these modifications, the calculated PCT for the Braidwood Station Unit 1 SIBLOCA increased 90OF and is now 1749 .OOF . However, this value remains within the acceptance criteria set forth in 10 CFR 50.46. Therefore, additional is not required . provides updated information regarding the PCT values for the limiting small break and large break loss-of-coolant accident evaluations for Braidwood Station Unit 1 . contains a detailed description for each change or error reported . Note 9 of describes the change that was made in support of the GSI-1 91 modifications discussed above.

There are no regulatory commitments contained in this letter . If you have any questions concerning this letter, please contact Mr. David Chrzanowski at (630) 657-2816 .

Patrick R. Simpson Manager - Licensing Attachments:

1 . Braidwood Station Unit 1 10 CFR 50.46 Report - Peak Cladding Temperature Rack-up Sheets 2 . Braidwood Station Unit 1 10 CFR 50.46 Report - Assessment Notes

Attachment 1 Braidwood Unit 1 10CFR50.46 Report Peak Cladding Temperature Rack-up Sheets PLANT NAME : Braidwood Station Unit 1 ECCS EVALUATION MODEL: Small Break Loss of Coolant Accident (SBLOCA)

REPORT REVISION DATE : 10108/07 CURRENT OPERATING CYCLE: 14 ANALYSIS OF RECORD (AOR)

Evaluation Model : NOTRUMP Calculation : Westinghouse CN-LIS-00-208, December 2000 Fuel : VANTAGE+ 17 x 17 Limiting Fuel Type : VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Heat Flux Hot Channel Factor (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 5%

Limiting Break Size and Location : 2" Low Tavg in the Bottom of the Cold Leg Notes: Zr-4/ZIRLO Clad Fuel Reference Peak Cladding Temperature (PCT) PCT = 1624 .0°F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10CFR50.46 report dated June 11, 2001 (see note 1) APCT = 0°F 10CFR50.46 report dated April 18, 2002 (see note 2) APCT = 0°F 10CFR50.46 report dated April 14, 2003 (see note 3) APCT = 0°F 10CFR50.46 report dated April 14, 2004 (see note 4) APCT = 35°F 10CFR50.46 report dated April 14, 2005 (see note 5) APCT = 0°F 10CFR50 .46 report dated April 14, 2006 (see note 6) APCT = 0°F 10CFR50 .46 report dated April 13, 2007 (see note 7) APCT = 0°F 10CFR50.46 report dated June 22, 2007 (see note 8) APCT = 0°F NET PCT PCT = 1659.0°F B. CURRENT LOCA MODEL ASSESSMENTS ECCS Flow Change Evaluation (see note 9) APCT = 90°F NET PCT PCT = 1749.0°F

Attachment 1 Braidwood Unit 1 10CFR50.46 Report Peak Cladding Temperature Rack-up Sheets PLANT NAME : B raidwood Station Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)

REPORT REVISION DATE : 10/08/07 CURRENT OPERATING CYCLE : 14 ANALYSIS OF RECORD Evaluation Model: COD (1996)

Calculation: Westinghouse CN-LIS-00-7, September 2000 Fuel : VANTAGE+ 17 x 17 Limiting Fuel Type : VANTAGE+ 17 x 17 Limiting Single Failure: Loss of one train of ECCS flow Heat Flux Hot Channel Factor (FO) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) =1 .70 Steam Generator Tube Plugging (SGTP) = 5%

Limiting Break Size and Location : Guillotine break in the Cold Leg Break Notes: Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2044.0°F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10CFR50.46 report dated June 11, 2001 (see note 1) APCT = 12°F 10CFR50.46 report dated April 18, 2002 (see note 2) APCT = 0°F 10CFR50.46 report dated April 14, 2003 (see note 3) APCT = 0°F 10CFR50.46 report dated April 14, 2004 (see note 4) APCT = 0°F 10CFR50.46 report dated April 14, 2005 (see note 5) APCT = 5°F 10CFR50.46 report dated April 14, 2006 (see note 6) APCT = 0°F 10CFR50.46 report dated April 13, 2007 (see note 7) APCT = 80°F 10CFR50 .46 report dated June 22, 2007 (see note 8) APCT = 20°F NET PCT PCT = 2161 .0°F B. CURRENT LOCA MODEL ASSESSMENTS None APCT = 0°F NET PCT PCT = 2161 .0°F

Attachment 2 Braidwood Unit 1 10CFR50.46 Report Assessment Notes 1 . Prior LOCA Model Assessment The 10CFR50 .46 report dated June 11, 2001 reported new large break loss of coolant accident (LBLOCA) and small break loss of coolant accident (SBLOCA) analyses to support operations at uprated power conditions . The same report assessed the impact from decay heat uncertainty error in Monte Carlo calculations on LBLOCA analysis and the impact from annular axial blankets on SBLOCA analysis . Evaluations for plant conditions and LBLOCA and SBLOCA model changes which resulted in OOF PCT change were reported . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .

2. Prior LOCA Model Assessment The 10CFR50 .46 report dated April 18, 2002 reported evaluations for LBLOCA and SBLOCA model changes which resulted in OOF PCT change. Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .

3. Prior LOCA Model Assessment The 10CFR50.46 report dated April 14, 2003 reported evaluations for LBLOCA and SBLOCA model changes which resulted in OOF PCT change. Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .

4. Prior LOCA Model Assessment The 10CFR50.46 report dated April 14, 2004 reported evaluations for LBLOCA model changes which resulted in OOF PCT change. A SBLOCA assessment related to NOTRUMP bubble rise/drift flux model inconsistency corrections, which resulted in 351F PCT assessment, was reported . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .

5. Prior LOCA Model Assessment The 10CFR50.46 report dated April 14, 2005 reported evaluations for LBLOCA model changes, which resulted in a 5'F PCT change due to the Revised Slowdown Heatup Uncertainty Calculation. Assembly N10S was reconstituted with two stainless steel filler rods during Braidwood Unit 1 Refueling Outage 11 . This assembly is reloaded into the core and is in use during Braidwood Unit 1 Cycle 12 operation . The introduction of up to five stainless steel filler rods has been evaluated and shown to have no impact on LBLOCA and SBLOCA analyses .

The estimated PCT effect is OOF. This assembly will be discharged during Reload 12 and will not be resident in the core for Braidwood Unit 1 Cycle 13 . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .

6. Prior LOCA Model Assessment The 10CFR50.46 report dated April 14, 2006 reported evaluations for LBLOCA model changes, which resulted in a OOF PCT change due to the Revised Iteration Algorithm for Calculating the Average Fuel Temperature, Pellet Radial Profile, Improved Automation of End of Blowdown Time, Thermodynamic Properties from THERMO, Vessel Unheated Conductor Noding and Containment Relative Humidity Assumption .

Attachment 2 Braidwood Unit 1 10CFR50 .46 Report Assessment Notes NOTRUMP General Code Maintenance resulted in OOF change for SBLOCA . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles.

7. Prior LOCA Model Assessment The 10CFR50 .46 report dated April 13, 2007 reported evaluations LOCA model changes and errors . The report documented general code maintenance for NOTRUMP, containment heat sink addition evaluation and NOTRUMP refined break spectrum, which resulted in OOF PCT impact . Cycle specific evaluations related to axial power shape distribution envelope violation were reported for the applicable operating cycles .

8. Prior LOCA Model Assessment A 30-day report was submitted to the NRC dated June 22, 2007, to report an error in the HOTSPOT code which resulted in an increase of 550F in the LBLOCA PCT for Braidwood Unit 2 and Byron Unit 2 and an 20'F LBLOCA PCT increase for Braidwood Unit 1 and Byron Unit 1 .

In addition, the report also reported minor errors with the reactor vessel data collections that potentially affected the vessel inlet and outlet fluid volume, metal mass, and surface area. The corrected values were evaluated for impact, and a OOF penalty was assessed for Byron Units 1 and 2, and Braidwood Units 1 and 2, SBLOCA analysis . The LBLOCA analyses were not affected by these minor errors .

9. ECCS Flow Change Evaluation Applicable to Braidwood Unit 1, the ECCS assessment evaluated changes in ECCS flow during the recirculation phase due to GSI-1 91 related safety injection Sl throttle valve replacements .

The evaluation of recirculation phase ECCS flow changes relative to impact on the current Analysis of Record (AOR) was performed for the following areas : SBLOCA and LBLOCA.

Large break loss of coolant accident analyses consider a fuel rod temperature transient that is over long before the transfer to ECCS sump recirculation . Therefore, there is no impact on the LBLOCA PCT calculations.

NOTRUMP and SBLOCTA calculations were performed for Braidwood Unit 1 ; the only input changes consisted of the revised ECCS flows during recirculation . Additional refined break sizes, including non-integer breaks, were selected and evaluated based on PCT results for each initial case in order to capture the worst PCT . Since beginning-of-life (BOL) PCT results approximately equal to or greater than 1700OF were calculated, burnup studies were performed for the limiting BOL break size. Based on the NOTRUMP and SBLOCTA calculations performed for Braidwood Unit 1, a conservative, bounding PCT assessment of +90OF was applied to the current Braidwood Unit 1 SBLOCA PCT .