RS-05-046, 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.
| ML051050037 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 04/14/2005 |
| From: | Bauer J Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-05-046 | |
| Download: ML051050037 (14) | |
Text
www.exeloncorp.com 10 CFR 50 .46 RS-05-046 April 14,2005 States United Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D .C. 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
Subject:
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" 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 (EGC), is submitting the annual report of the Emergency Core Cooling System (ECCS)
Evaluation Model changes and errors for Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2. This annual report is to be submitted to the NRC by April 14, 2005 . , "Peak Cladding Temperature Rack-Up Sheets," provides updated information regarding the peak cladding temperature (PCT) for the limiting small break and large break loss-of-coolant accident (LOCA) analyses evaluations for the Byron and Braidwood Stations . , "Assessment Notes," contains a detailed description for each change or error reported .
Please contact C. W. Szabo at (630) 6570821 should you have any questions concerning this report .
Respectfully, Joseph A. Bauer Manager - Licensing : Peak Cladding Temperature Rack-Up Sheets : Assessment Notes
Attachment 1 BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos. 50-456 and 50-457 License Nos. NPF-72 and NPF-77 and BYRON STATION UNITS 1 AND 2 Docket Nos. 50-454 and 50-455 License Nos. NPF-37 and NPF-66 10 CFR 50 .46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments Peak Cladding Temperature Rack-Up Sheets
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: 4/4/05 CURRENT OPERATING CYCLE : 12 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 Emergency Core Cooling System (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 : 2" Low Tavg 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 10 CFR 50 .46 report dated June 11, 2001 (see note 1) APCT = O'F 10 CFR 50 .46 report dated April 18, 2002 (see note 2) APCT = 01F 10 CFR 50 .46 report dated April 14, 2003 (see note 3) PCT_= 01F 10 CFR 50 .46 report dated April 14, 2004 (see note 4) APCT = 351F NET PCT PCT = 1659 .0'F B. CURRENT LOCA MODEL ASSESSMENTS Reconstituted Fuel (see note 5) APCT := OOF General Code Maintenance (N TRUMP) (see note 7) A PCT = OOF NET PCT PCT = 1659 .0*F
Peak Cladding Temperature Rack-up Sheets PLANT NAME : Braidwood Station, Unit 1 ECCS EVALUATION MODEL: Large Break Loss of Coolant Accident (LBLOCA)
REPORT REVISION DATE : 04/04/05 CURRENT OPERATING CYCLE : 12 AOR Evaluation Model : WCOBRA/TRAC 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 (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 5%
Limiting Break Size: Guillotine Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2044 .0'F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) APCT 12'F 10 CFR 50.46 report dated April 18, 2002 (see note 2) APCT 01 F 10 CFR 50.46 report dated April 14, 2003 (see note 3) APCT OOF 10 CFR 50 .46 report dated April 14, 2004 (see note 4) APCT OOF NET PCT PCT = 2056.0*F B. CURRENT LOCH MODEL ASSESSMENTS Reconstituted Fuel (see note 5) A PCT 011F
&K Power Shape Distribution Violation (see note 6) APCT 80'F Implementation of ASTRUM Capability in HOTSPOT (see note 8) APCT OOF Improved Automation of End of Blowdown Time (see note 9) APCT OOF General Code Maintenance (WC/T) (see note 10) APCT OOF Revised Blowdown Heatup Uncertainty Calculation (see note 11) APCT 51 F NET PCT PCT = 2141 .OOF Page 2 of 8
Peak Cladding Temperature Rack-up Sheets PLANT NAME : Braidwood Station, Unit 2 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: 04/04/05 CURRENT OPERATING CYCLE : 12*
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) = 10%
Limiting Break: 2" Low Tavg Notes: Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 1627 .0'F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50 .46 report dated June 11, 2001 (see note 1) APCT = 30F 10 CFR 50 .46 report dated April 18, 2002 (see note 2) APCT = O'F 10 CFR 50 .46 report dated April 14, 2003 (see note 3) APCT = O'F 10 CFR 50 .46 report dated April 14, 2004 (see note 4) APCT = 35'F NET PCT PCT = 1665 .0*F B. CURRENT LOCA MODEL ASSESSMENTS
)de Maintenance (NOTRUMP) (see note 7) APCT = 0-F-]
- Note that Braidwood Station, Unit 2, Cycle 12 projected startup date is early May 2005.
Peak Cladding Temperature Rack-up Sheets PLANT NAME : Braidwood Station, Unit 2 ECCS EVALUATION MODEL : LBLOCA REPORT REVISION DATE: 04/04/05 CURRENT OPERATING CYCLE : 12*
AOR Evaluation Model : WCOBRA/TRAC 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 (FQ) = 2 .60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 10%
Limiting Break Size: Guillotine Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2088.0'F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) APCT = 121F 10 CFR 50.46 report dated April 18, 2002 (see note 2) APCT = 011F 10 CFR 50.46 report dated April 14, 2003 (see note 3) A PCT = 01F 10 CFR 50.46 report dated April 14, 2004 (see note 4) APCT = 01F NET PCT PCT = 2100.0*F B. CURRENT LOCA MODEL ASSESSMENTS Axial Power Shape Distribution Violation (see note 6) APCT = 81 F Implementation of ASTRUM Capability in HOTSPOT (see note 8) A PCT = OOF Improved Automation of End of Blowdown Time (see note 9) APCT = OOF General Code Maintenance (WC/T) (see note 10) APCT = 01F Revised Blowdown Heatup Uncertainty Calculation (see note 11) APCT = 51F NET PCT PCT = 2113 .0*F
- Note that Braidwood Station, Unit 2, Cycle 12 projected startup date is early May 2005.
Peak Cladding Temperature Rack-up Sheets PLANT NAME: Byron Station, Unit 1 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: 04/04/05 CURRENT OPERATING CYCLE : 14 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: 2" Low Tavg Notes: Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 16210'F MARGIN ALLOCATION A. PRIOR LOSS OF COOLANT ACCIDENT (LOCA) MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) APCT = 01F 1 () CFR 50.46 report dated April 18, 2002 (see note 2) APCT = OOF 10 CFR 50.46 report dated April 14, 2003 (see note 3) APCT = OOF 10 CFR 50.46 report dated April 14, 2004 (see note 4) APCT = 350F NET PCT PCT = 1659 .0'F B. CURRENT LOCH MODEL ASSESSMENTS I General Code Maintenance (NOTRUMP) (see note 7) 1 A PCT_Oor NET PCT PCT = 1659.0*F
Peak Cladding Temperature Rack-up Sheets PLANT NAME : Byron Station, Unit 1 ECCS EVALUATION MODEL : LBLOCA REPORT REVISION DATE : 04/04/05 CURRENT OPERATING CYCLE: 14 A0FR Evaluation Model: WCOBRA/TRAC 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 Steam Generator Tube Plugging (SGTP) = 5%
Heat Flux Hot Channel Factor (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Limiting Break Size : Guillotine Notes: Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2044.0'F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50 .46 report dated June 11, 2001 (see note 1) APCT 121F 10 CFR 50 .46 report dated April 18, 2002 (see note 2) APCT 01F 10 CFR 50 .46 report dated April 14, 2003 (see note 3) APCT 01F 10 CFR 50 .46 report dated April 14, 2004 (see note 4) APCT 01F NET PCT PCT = 2056.0*F B. CURRENT LOCH MODEL ASSESSMENTS Axial Power Shape Distribution Violation (see note 6) APCT 801F Implementation of ASTRUM Capability in HOTSPOT (see note 8) APCT 01F Improved Automation of End of Blowdown Time (see note 9) A PCT 01F General Code Maintenance (WC/T) (see note 10) APCT 01F Revised Blowdown Heatup Uncertainty Calculation (see note 11) APCT 51F NET PCT PCT = 2141 .0*F
Peak Cladding Temperature Rack-up Sheets PLANT NAME : Byron Station, Unit 2 ECCS EVALUATION MODEL: SBLOCA REPORT REVISION DATE: 04/04/05 CURRENT OPERATING CYCLE: 12 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) = 10%
Limiting Break: 2" Low Tavg Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 1627 .0'F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50 .46 report dated June 11, 2001 (see note 1) APCT = 30F 10 CFR 50.46 report dated April 18, 2002 (see note 2) APCT = O'F 10 CFR 50.46 report dated April 14, 2003 (see note 3) A PCT = OOF 10 CFR 50.46 report dated April 14, 2004 (see note 4) APCT = 350F NET PCT PCT = 1665 .0'F B. CURRENT LOCH MODEL ASSESSMENTS I G Dde Maintenance (NOTRUMP) (see note 7) 1 APCT = OOF NET PCT PCT = 1665 .0*F
Peak Cladding Temperature Rack-up Sheets PLANT NAME : Boron Station, Unit 2 ECCS EVALUATION MODEL : LBLOCA REPORT REVISION DATE: 04/04/05 CURRENT OPERATING CYCLE: 12 AOR Evaluation Model : WCOBRA/TRAC 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 (FQ) = 2.60 Nuclear Enthalpy Rise Hot Channel Factor (FNAH) = 1 .70 Steam Generator Tube Plugging (SGTP) = 10%
Limiting Break Size : Guillotine Notes : Zr-4/ZIRLO Clad Fuel Reference PCT PCT = 2088 .0'F MARGIN ALLOCATION A. PRIOR LOCA MODEL ASSESSMENTS 10 CFR 50.46 report dated June 11, 2001 (see note 1) APCT 120 F 10 CFR 50.46 report dated April 18, 2002 (see note 2) APCT 01 F 10 CFR 50.46 report dated April 14, 2003 (see note 3) APCT 01 F 10 CFR 50.46 report dated April 14, 2004 (see note 4) APCT 0 IF Axial Power Shape Distribution Violation (see note 6) APCT 81F NET PCT PCT = 2108.OOF B. CURRENT LOCH MODEL ASSESSMENTS Implementation of ASTRUM Capability in HOTSPOT (see note 8) A PCT = O'F Improved Automation of End of Blowdown Time (see note 9) APCT = O'F General Code Maintenance (WC/T) (see note 10) APCT = 01F Revised Blowdown Heatup Uncertainty Calculation (see note 11) APCT = 51F NET PCT PCT = 2113.0'F
Attachment 2 BRAIDWOOD STATION UNITS 1 AND 2 Docket Nos . 50-456 and 50-457 License Nos . NPF-72 and NPF-77 and BYRON STATION UNITS 1 AND 2 Docket Nos. 50-454 and 50-455 License Nos. NPF-37 and NPF-66 10 CFR 50 .46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors" Report of the Emergency Core Cooling System Evaluation Model Changes and Errors Assessments Assessment Notes
Assessment Notes 1 . Prior Loss-of-Coolant Accident (LOCH} Model Assessment report The 10 CFR 50.46 dated June 11, 2001 reported new large-break LOCA (LBLOCA) and small-break LOCA (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 0 OF peak cladding temperature (PCT) change were reported .
Cycle specific evaluations related to axial power shape distribution envelope violation was reported for the applicable operating cycles .
2 . Prior LOCH Model Assessment The 10 CFR 50.46 report dated April 18, 2002 reported evaluations for LBLOCA and SBLOCA model changes which resulted in 01F PCT change. Cycle specific evaluations related to axial power shape distribution envelope violation was reported for the applicable operating cycles .
3 . Prior LOCH Model Assessment The 10 CFR 50 .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 was reported for the applicable operating cycles .
- 4. Prior LOCH Model Assessment The 10 CFR 50 .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. Reconstituted Fuel Assembly N 1 OS was reconstituted with two stainless steel filler rods during Braidwood Station, Unit 1, refueling outage 11 . This assembly is reloaded into the core and is in use during Braidwood Station, 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 LBOCA and SBLOCA analyses . The estimated PCT effect is OOF.
- 6. Axial power Shape Distribution Envelope Violation (PMID, PBOT)
The LBLOCA analysis is performed based on assuming an axial power shape distribution envelope (PMID, PBOT), where PMID is the power in the middle one-third of the core ; and PBOT is the power in the lower one-third of the core . The envelope is pertinent to the BELOCA analysis and is presented as Figure 11-1 of WCAP-1 5585, "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Byron /Braidwood Units 1 and 2 Nuclear Plant," November 2000. For every reload cycle Westinghouse verifies that the envelope remains limiting . If there is a violation then a PCT penalty is calculated .
For Braidwood Station, Unit 1, Cycle 12 there was a violation and a PCT penalty of 80°F was calculated.
Assessment Notes For Braidwood Station, Unit 2, Cycle 12 there was a violation and a PCT penalty of 80 F was calculated .
For Byron Station, Unit 1, Cycle 14 there was a violation and a PCT penalty of 801F was calculated .
For Byron Station, Unit 2, Cycle 12 there was a violation and a PCT penalty of 80F was calculated . This penalty has been reported in 10 CFR 50 .46 report dated April 14, 2004.
For Braidwood Station, Unit 2, Cycle 12 and Byron Station, Unit 1, Cycle 14, Westinghouse found two types of violation. For the violations outside of the sampling range shown in Figure 11-1 of WCAP-1 5585 but inside the response surface shown in Figure 9.2-1 of VVCAF41 5585, a PCT penalty of 8'F and 80OF was calculated for Braidwood Station, Unit 2, Cycle 12 and Byron Station, Unit 1, Cycle 14 respectively . This is the same type of violation as the cycles discussed above.
The second type of violation is for power shapes slightly outside the response surface shown in Figure 9 .2-1 of WCAP-1 5585. These violations were determined to be non-limiting power shapes and were evaluated by extrapolating the power distribution response surface in order to predict PCT . In all cases, the predicted PCT was non-limiting as compared to the LBLOCA PCT reported in the Attachment 1 LBLOCA PCT sheet. The Westinghouse reload methodology approved by the NRC, WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology, March 1978," allows evaluation of key parameters slightly out of bounds using conservative quantitative evaluation. The extrapolation methodology used in this evaluation is consistent with the methodology for extrapolating FQ and FAH described in WCAP-12945-P-A.
7 . General Code Maintenance (NOTRUMP)
Various changes in code input and output format have been made to enhance usability and help preclude errors in analyses . This includes both input changes (e.g ., more relevant input variables defined and more common input values used as defaults) and input diagnostics designed to preclude unreasonable values from being used, as well as various changes to code output which have no effect on calculated results. In addition, various updates were made to eliminate inactive coding, improve active coding, and enhance commenting, both for enhanced usability and to facilitate code debugging when necessary. These changes represent Discretionary Changes that will be implemented on a forward-fit basis, in accordance with Section 4.1 .1 of WCAP-1 3451 .
The nature of these changes leads to an estimated PCT impact of OOF.
8 . Implementation of ASTRUM Capability in HOTSPOT The HOTSPOT computer code was modified to be compatible with the Automated Statistical Treatment of Uncertainty Methodology (ASTRUM) . An option is used to trigger the ASTRUM HOTSPOT technique or the Monte Cado made used in the previous Best Estimate LBLOCA evaluation models . These changes were determined to be Discretionary changes in accordance with Section 4.1 .1 of WCAP-1345 1 .
None of these changes affect the results of design basis analysis performed using these evaluation models . Therefore, the estimated effect is OOF .
Assessment Notes
- 9. Improved Automation of End of Slowdown Time Heat transfer multipliers are considered in the uncertainty methodology as a function of time period in the transient. The blowdown cooling heat transfer multipliers are applied during the time period following turnaround of the blowdown heatup through the end of blowdown. For simplicity, the end of blowdown was originally defined as the time when the system pressure dropped below 40 psia. In cases where the pressure did not drop below 40 psia, the analyst would manually redefine the end of blowdown based on time of minimum system pressure The automated selection of the end of blowdown time was improved by replacing the 40-psia criterion with a selection based on the time at which the system pressure stops decreasing .
All of these changes are considered to be Discretionary changes in accordance with Section 4.1 .1 of WCAP-1 3451 .
The correct end of blowdown time was selected in the prior analyses . Therefore, the estimated effect is 0-F .
10 . General Code Maintenance (WC/T)
A number of coding changes were made as part of normal code maintenance . Examples include correction of debug plots not used in design analysis and improved consistency between the HOTSPOT nominal PCT (not used in uncertainty analysis) and ACOBRA/TRAC PCT. All of these changes are considered to be Discretionary changes in accordance with Section 4.1 .1 of WCAP- 1345 1 .
None of these changes affect the results of design basis analyses . Therefore, the estimated effect is OOF.
11 . Revised Slowdown Heatup Uncertainty Distribution Correction of modeling inconsistencies and input errors in the LOFT input decks have resulted in a change in the predicted peak cladding temperature transients . Revised analyses of the LOFT and ORNL tests were performed using the current version of ACOBRA/TRAC. As a result of this re-analysis, revised blowdown heatup heat transfer coefficients were developed and the revised cumulative distribution function (CDF) was programmed into a new version of HOTSPOT. The revised CDF was previously reported to the NRC by Westinghouse in LTR-NRC-04-1 1 . The overall code uncertainty for blowdown was also recalculated and programmed into a new version of MONTECF . The overall code uncertainty for reflood was not affected . The corrections were determined to be Non-Discretionary changes in accordance with Section 4 .1 .2 of WCAP-13451 .
An estimate of the PCT effect of the revised blowdown heatup CDF was performed by calculating the impact on the reference transient for representative 2-, 3- and 4-loop plants .
The estimates bound all of the 95th percentile HOTSPOT results. Estimates of the effect of the revised overall code uncertainty for blowdown were made on a plant specific basis by repeating the MONTECF analysis .
The limiting PCT occurs during the reflood period for Byron Station and Braidwood Station.
A 50F PCT assessment is assigned.