ML20063L282

Application of Westinghouse Full Spectrum LOCA Evaluation Model

ML20063L282
Person / Time
Site:	Byron, Braidwood
Issue date:	02/28/2020
From:	Westinghouse
To:	Office of Nuclear Reactor Regulation
References
CAE-19-26/CCE-19-25, RS-20-010
Download: ML20063L282 (85)

Text

Attachment 4 Non-Proprietary Class 3 Version of the FSLOCA TM LAR Input Document, "APPLICATION OF WESTINGHOUSE FULL SPECTRUM' LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1 AND 2"

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page I of84 APPLICATION OF WESTINGHOUSE FULL SPECTRUM LOCA EVALUATION MODEL TO BRAIDWOOD STATION, UNITS 1 AND 2, AND BYRON STATION, UNITS 1AND2

1.0 INTRODUCTION

Analyses with the FULL SPECTRUMTM loss-of-coolant accident (FSLOCA') evaluation model (EM) have been completed for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2. This license amendment request (LAR) for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, requests approval to apply the Westinghouse FSLOCA EM.

The FSLOCA EM (Reference 1) was developed to address the full spectrum of loss-of-coolant accidents (LOCAs) which result from a postulated break in the reactor coolant system (RCS) of a pressurized water reactor (PWR). The break sizes considered in the Westinghouse FSLOCA EM include any break size in which break flow is beyond the capacity of the normal charging pumps, up to and including a double-ended guillotine (DEG) rupture of an RCS cold leg with a break flow area equal to two times the pipe area, including what traditionally are defined as Small and Large Break LOCAs.

The break size spectrum is divided into two regions. Region I includes breaks that are typically defined as Small Break LOCAs (SBLOCAs). Region II includes break sizes that are typically defined as Large Break LOCAs (LBLOCAs).

The FSLOCA EM explicitly considers the effects of fuel pellet thermal conductivity degradation (TCD) and other burnup-related effects by calibrating to fuel rod performance data input generated by the PADS code (Reference 2), which explicitly models TCD and is benchmarked to high burnup data in Reference

2. The fuel pellet thermal conductivity model in the WCOBRA/TRAC-TF2 code used in the FSLOCA EM explicitly accounts for pellet thermal conductivity degradation.

Three of the Title 10 of the Code of Federal Regulations (CFR) 50.46 criteria (peak cladding temperature (PCT), maximum local oxidation (MLO), and core-wide oxidation (CWO)) are considered directly in the FSLOCA EM. A high probability statement is developed for the PCT, MLO, and CWO that is needed to demonstrate compliance with 10 CFR 50.46 acceptance criteria (b)(l), (b)(2), and (b)(3) (Reference 3) via statistical methods. The MLO is defined as the sum of pre-transient corrosion and transient oxidation consistent with the position in Information Notice 98-29 (Reference 4). The coolable geometry acceptance criterion, 10 CFR 50.46 (b)(4), is assured by compliance with acceptance criteria (b)(l),

(b)(2), and (b)(3), and demonstrating that fuel assembly grid deformation due to combined seismic and LOCA loads does not extend to the in-board fuel assemblies such that a coolable geometry is maintained.

The FSLOCA EM has been generically approved by the Nuclear Regulatory Commission (NRC) for Westinghouse 3-loop and 4-loop plants with cold leg Emergency Core Cooling System (ECCS) injection (Reference 1). Since Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, are Westinghouse designed 4-loop plants with cold leg ECCS injection, the approved method is applicable.

Information required to address Limitations and Conditions 9 and 10 of the NRC's Safety Evaluation (SE) for Reference 1 was submitted in Reference 12 in support of application of the FSLOCA EM to Westinghouse 4-loop plants.

This report summarizes the application of the Westinghouse FSLOCA EM to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2. The application of the FSLOCA EM to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, is consistent with the NRC-approved methodology (Reference FULL SPECTRUM and FSLOCA are trademarks or registered trademarks of Westinghouse Electric Company LLC, its subsidiaries and/or affiliates in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 2 of 84 1), with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The application of the FSLOCA EM is consistent with the conditions and limitations as identified in the NRC's SE for Reference 1, and is also applicable for the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, plant designs and operating conditions.

Both Exelon and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses. These interface processes, along with Westinghouse internal processes for assessing EM changes and errors, are used to identify the need for LOCA analysis impact assessments.

The major plant parameters and analysis assumptions used in the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses with the FSLOCA EM are provided in Tables 1 through 7. Note that these parameters are consistent between all units.

2.0 METHOD OF ANALYSIS 2.1 FULL SPECTRUM LOCA Evaluation Model Development In 1988, the NRC amended the requirements of 10 CFR 50.46 (Reference 3 and Reference 6) and Appendix K, "ECCS Evaluation Models," to permit the use of a realistic EM to analyze the performance of the ECCS during a hypothetical LOCA. Westinghouse's previously approved best-estimate LBLOCA EM is discussed in Reference 8. The EM is referred to as the Automated Statistical Treatment of Uncertainty Method (ASTRUM), and was developed following Regulatory Guide (RG) 1.157 (Reference 7).

When the FSLOCA EM was being developed, the NRC issued RG 1.203 (Reference 9) which expands on the principles ofRG 1.157, while providing a more systematic approach to the development and assessment process of a PWR accident and safety analysis EM. Therefore, the development of the FSLOCA EM followed the Evaluation Model Development and Assessment Process (EMDAP), which is documented in RG 1.203. While RG 1.203 expands upon RG 1.157, there are certain aspects of RG 1.157 which are more detailed than RG 1.203; therefore, both RGs were used for the development of the FSLOCAEM.

Separate analyses were performed for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, due to the different steam generator designs in the Unit 1 plants versus Unit 2 plants. The Braidwood Station and Byron Station, Unit 2 plants have the original Westinghouse Model D5 steam generators installed. However, replacement steam generators supplied by Babcock and Wilcox International (BWI) were installed in Braidwood Station and Byron Station, Units 1. Because different steam generator designs would be expected to affect transient behavior and analysis results, separate analyses were completed. The units are otherwise functionally similar with respect to the major systems and components.

2.2 WCOBRA/TRAC-TF2 Computer Code The FSLOCA EM (Reference 1) uses the WCOBRA/TRAC-TF2 code to analyze the system thermal-hydraulic response for the full spectrum of break sizes. WCOBRA/TRAC-TF2 was created by combining a ID module (TRAC-P) with a 3D module (based on Westinghouse modified COBRA-TF). The lD and 3D modules include an explicit non-condensable gas transport equation. The use ofTRAC-P allows for the extension of a two-fluid, six-equation formulation of the two-phase flow to the lD loop components.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 3 of 84 This new code is WCOBRA/TRAC-TF2, where "TF2" is an identifier that reflects the use of a three-field (TF) formulation of the 3D module derived by COBRA-TF and a two-fluid (TF) formulation of the ID module based on TRAC-P.

This best-estimate computer code contains the following features:

1. Ability to model transient three-dimensional flows in different geometries inside the reactor vessel

2. Ability to model thermal and mechanical non-equilibrium between phases

3. Ability to mechanistically represent interfacial heat, mass, and momentum transfer in different flow regimes

4. Ability to represent important reactor components such as fuel rods, steam generators (SGs),

reactor coolant pumps (RCPs ), etc.

A detailed assessment of the computer code WCOBRA/TRAC-TF2 was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the ability of the code to predict key physical phenomena for a LOCA. Modeling of a LOCA introduces additional uncertainties which are identified and quantified in the plant-specific analysis. The reactor vessel and loop noding scheme used in the FSLOCA EM is consistent with the noding scheme used for the experiment simulations that form the validation basis for the physical models in the code. Such noding choices have been justified by assessing the model against large and full scale experiments.

2.3 Compliance with FSLOCA EM Limitations and Conditions The NRC's SE for Reference 1 contains 15 limitations and conditions on the NRC-approved FSLOCA EM. A summary of each limitation and condition and how it was met is provided below.

Limitation and Condition Number 1 Summary The FSLOCA EM is not approved to demonstrate compliance with 10 CFR 50.46 acceptance criterion (b)(5) related to the long-term cooling.

Compliance The analyses for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, with the FSLOCA EM are only being used to demonstrate compliance with 10 CFR 50.46 (b)(l) through (b)(4).

Limitation and Condition Number 2 Summary The FSLOCA EM is approved for the analysis of Westinghouse-designed 3-loop and 4-loop PWRs with cold-side injection. Analyses should be executed consistent with the approved method, or any deviations from the approved method should be described and justified.

Compliance Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, are Westinghouse-designed 4-loop PWRs with cold-side injection, so they are within the NRC-approved methodology. The analyses for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, utilize the NRC-approved FSLOCA

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 4 of84 methodology, except for the changes which were previously transmitted to the NRC pursuant to 10 CFR 50.46 in LTR-NRC-18-30 (Reference 5). After completion of the analyses for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, two errors were discovered in the WCOBRA/TRAC-TF2 code that can occur under certain conditions. These errors were found to have a negligible impact on analysis results with the FSLOCA EM as described in LTR-NRC-19-6 (Reference 13).

The treatment for the uncertainty in the gamma energy redistribution is discussed on pages 29-75 and 29-76 ofWCAP-16996-P-A, Revision 1 (Reference 1), and the equation for the assumed increase in hot rod and hot assembly relative power is presented on page 29-76. The power increase in the hot rod and hot assembly due to energy redistribution in the application of the FSLOCA EM to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, was calculat~d incorrectly. This error resulted in a 0% to 5%

deficiency in the modeled hot rod and hot assembly rod linear heat rates on a run-specific basis, depending on the as-sampled value for the uncertainty. The effect of the error correction was evaluated against the application of the FSLOCA EM to Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2.

The error correction has only a limited impact on the power modeled for a single assembly in the core.

As such, there is a negligible impact of the error correction on the system thermal-hydraulic response during the postulated LOCA.

For Region I, the primary impact of the error correction is on the rate of cladding heatup above the two-phase mixture level in the core during the boiloff phase'. The PCT impact was assessed using run-specific PCT versus linear heat rate relationships and the run-specific hot rod and hot assembly linear heat rate increase that would result from the error correction. Using this approach, the correction of the error was estimated to increase the Region I analysis PCT by l 7°F for Braidwood Station and Byron Station, Units 1, leading to a final result of 1181 °F for the Region I analysis. Using this approach, the correction of the error was estimated to increase the Region I analysis PCT by 1°F for Braidwood Station and Byron Station, Units 2, leading to a final result of 1169°F for the Region I analysis.

For Region II, parametric PWR sensitivity studies, derived from a subset of uncertainty analysis simulations covering various design features and fuel arrays, were examined to determine the sensitivity of the analysis results to the error correction. The PCT impact from the error correction was found to be different for the different transient phases (i.e., blowdown versus reflood) based on the PWR sensitivity studies and existing power distribution sensitivity studies. Based on the results from the PWR sensitivity studies, the correction of the error is estimated to increase the Region II analysis PCT by 31 °F for Braidwood Station and Byron Station, Units 1, leading to an analysis result of 1643°F for the Region II analysis assuming loss-of-offsite power (LOOP) and 1641 °F for the Region II analysis assuming offsite power available (OPA). Based on the results from the PWR sensitivity studies, the correction of the error is estimated to increase the Region II analysis PCT by 31 °F for Braidwood Station and Byron Station, Units 2, leading to an analysis result of 1711 °F for the Region II analysis assuming LOOP and l 752°F for the Region II analysis assuming OPA.

All of the analysis results including the error correction continue to maintain compliance with the 10 CFR 50.46 acceptance criteria.

Limitation and Condition Number 3 Summary For Region II, the containment pressure calculation will be executed in a manner consistent with the approved methodology (i.e., the COCO or LOTIC2 model will be based on appropriate plant-specific design parameters and conditions, and engineered safety features which can reduce pressure are

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 5 of84 modeled). This includes utilizing a plant-specific initial containment temperature, and only taking credit for containment coatings which are qualified and outside of the break zone-of-influence.

Compliance The containment pressure calculations for the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses were performed consistent with the NRC-approved methodology. Appropriate design parameters and conditions were modeled, as were the engineered safety features which can reduce the containment pressure. A minimum initial temperature associated with normal full-power operating conditions was modeled, and no coatings were credited on any of the containment structures.

Limitation and Condition Number 4 Summary The decay heat uncertainty multiplier will be [

]8'c The analysis simulations for the FSLOCA EM will not be executed for longer than 10,000 seconds following reactor trip unless the decay heat model is appropriately justified. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results will be provided in the analysis submittal in units of sigma and absolute units.

Compliance Consistent with the NRC-approved methodology, the decay heat uncertainty multiplier was [

r,c for the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses. The analysis simulations were all executed for no longer than 10,000 seconds following reactor trip. The sampled values of the decay heat uncertainty multiplier for the cases which produced the Region I and Region II analysis results have been provided in units of sigma and approximate absolute units in Tables 16 and 17.

Limitation and Condition Number S Summary The maximum assembly and rod length-average bumup is limited to [

]8'c respectively.

Compliance The maximum analyzed assembly and rod length-average bumup is less than or equal to [

r,c respectively, for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2.

Limitation and Condition Number 6 Summary The fuel performance data for analyses with the FSLOCA EM should be based on the PADS code (at present), which includes the effect of thermal conductivity degradation. The nominal fuel pellet average temperatures and rod internal pressures should be the maximum values, and the generation of all the PADS fuel performance data should adhere to the NRC-approved PADS methodology.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 6 of 84 Compliance PADS fuel performance data is utilized in the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses with the FSLOCA EM. The analyzed fuel pellet average temperatures bound the maximum values calculated in accordance with Section 7.5.1 of Reference 2, and the analyzed rod internal pressures were calculated in accordance with Section 7.5.2 of Reference 2.

Limitation and Condition Number 7 Summary The YDRAG uncertainty parameter should be [

Ja,c Compliance Consistent with the NRC-approved methodology, the YDRAG uncertainty parameter was [

]a,c for the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Region I analyses.

Limitation and Condition Number 8 Summary The [

Compliance Consistent with the NRC-approved methodology, the [

p,c for the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Region I analyses.

Limitation and Condition Number 9 /

Summary For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to confirm that the [

r,c for the plant design being analyzed. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

Compliance Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, are Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 12.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 7 of 84 Limitation and Condition Number 10 Summaty For PWR designs which are not Westinghouse 3-loop PWRs, a sensitivity study will be executed to: 1) demonstrate that no unexplained behavior occurs in the predicted safety criteria across the region boundary, and 2) ensure that the [

re must cover the equivalent 2 to 4-inch break range using RCS-volume scaling relative to the demonstration plant. This sensitivity study should be executed once, and then referenced in all applications to that particular plant class.

Additionally, the minimum sampled break area for the analysis of Region II should be 1 ft2.

Compliance Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, are Westinghouse-designed 4-loop PWRs. The requested sensitivity study was performed for a 4-loop Westinghouse-designed PWR and is discussed in Reference 12.

The minimum sampled break area for the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Region II analyses was 1 ft2.

Limitation and Condition Number 11 Summaty There are various aspects of this Limitation and Condition, which are summarized below:

1. The [ r,c the Region I and Region II analysis seeds, and the analysis inputs will be declared and documented prior to performing the Region I and Region II uncertainty analyses. The [

r,c and the Region I and Region II analysis seeds will not be changed throughout the remainder of the analysis once they have been declared and documented.

2. If the analysis inputs are changed after they have been declared and documented, for the intended purpose of demonstrating compliance with the applicable acceptance criteria, then the changes and associated rationale for the changes will be provided in the analysis submittal. Additionally, the preliminary values for peak cladding temperature (PCT), maximum local oxidation (MLO),

and core-wide oxidation (CWO) which caused the input changes will be provided. These preliminary values are not subject to Appendix B verification, and archival of the supporting information for these preliminary values is not required.

3. Plant operating ranges which are sampled within the uncertainty analysis will be provided in the analysis submittal for both regions.

Compliance This Limitation and Condition was met for the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses as follows:

1. The [ r,c the Region I and Region II analysis seeds, and the analysis inputs were declared and documented prior to performing the Region I and Region II uncertainty analyses. The [

r,c and the Region I and Region II analysis seeds were not changed once they were declared and documented.

2. The analysis inputs were not changed once they were declared and documented.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 8 of 84

3. The plant operating ranges which were sampled within the uncertainty analyses are provided for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, in Table 1. Note that these operating ranges are consistent between all units.

Limitation and Condition Number 12 Summary The plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves must be adequately accounted for in analysis with the FSLOCA EM.

Compliance A bounding plant-specific dynamic pressure loss from the steam generator secondary-side to the main steam safety valves was modeled in the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses.

Limitation and Condition Number 13 Summary In plant-specific models for analysis with the FSLOCA EM: 1) the [

t*c and 2) the

]a,c Compliance The [

]a,c in the analyses for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2. The [ r,c in the analyses.

Limitation and Condition Number 14 Summary For analyses with the FSLOCA EM to demonstrate compliance against the current 10 CFR 50.46 oxidation criterion, the transient time-at-temperature will be converted to an equivalent cladding reacted (ECR) using either the Baker-Just or the Cathcart-Pawel correlation. In either case, the pre-transient corrosion will be summed with the LOCA transient oxidation. If the Cathcart-Pawel correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 13 percent limit. If the Baker-Just correlation is used to calculate the LOCA transient ECR, then the result shall be compared to a 17 percent limit.

Compliance For the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses, the Baker-Just correlation was used to convert the LOCA transient time-at-temperature to an ECR. The resulting LOCA transient ECR was then summed with the pre-existing corrosion for comparison against the 10 CFR 50.46 local oxidation acceptance criterion of 17%.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 9 of 84 Limitation and Condition Number 15 Summary The Region II analysis will be executed twice; once assuming LOOP and once assuming OP A. The results from both analysis executions should be shown to be in compliance with the 10 CFR 50.46 acceptance criteria.

The [

Compliance For both the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, analyses, the Region II uncertainty analysis was performed twice; once assuming a LOOP and once assuming OPA. The results from both analyses that were performed are in compliance with the 10 CFR 50.46 acceptance criteria (see Section 5.0).

The [

r,c 3.0 REGION I ANALYSIS 3.1 Description of Representative Transient The small break LOCA transient can be divided into time periods in which specific phenomena are occurring, as discussed below.

Blowdown The rapid depressurization of the RCS coincides with subcooled liquid flow through the break. Following the reactor trip on the low pressurizer pressure setpoint, the pressurizer drains, and safety injection is initiated on the low pressurizer pressure SI setpoint. After reaching this setpoint and applying the safety injection delays, high pressure safety injection flow begins. Phase separation begins in the upper head and upper plenum near the end of this period until the entire RCS eventually reaches saturation, ending the rapid depressurization slightly above the steam generator secondary side pressure near the modeled main steam safety valve (MSSV) setpoint.

Natural Circulation This quasi-equilibrium phase persists while the RCS pressure remains slightly above the secondary side pressure. The system drains from the top down, and while significant mass is continually lost through the break, the vapor generated in the core is trapped in the upper regions by the liquid remaining in the crossover leg loop seals. Throughout this period, the core remains covered by a two-phase mixture and the fuel cladding temperatures remain at the saturation temperature level.

Loop Seal Clearance As the system drains, the liquid levels in the downhill side of the pump suction (crossover leg) become depressed all the way to the bottom elevations of the piping, allowing the steam trapped during the natural circulation phase to vent to the break (i.e., a process called loop seal clearance). The break flow and the flow through the RCS loops become primarily vapor. Relief of a static head imbalance allows for a quick but temporary recovery of liquid levels in the inner portion of the reactor vessel.

Boil-Off With a vapor vent path established after the loop seal clearance, the RCS depressurizes at a rate controlled by the critical flow, which continues to be a primarily high quality mixture of water and steam. The RCS

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 10 of84 pressure remains high enough such that safety injection flow cannot make up for the primary system fluid inventory lost through the break, leading to core uncovery and a fuel rod cladding temperature heatup.

Core Recovery The RCS pressure continues to decrease, and once it reaches that of the accumulator gas pressure, the introduction of additional ECCS water from the accumulators replenishes the reactor vessel inventory and recovers the core mixture level. The transient is considered over as the break flow is compensated by the injected flow.

3.2 Analysis Results The Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Region I analyses were performed in accordance with the NRC-approved methodology in Reference 1, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The transient that produced the analysis PCT result for Braidwood Station and Byron Station, Units 1, is a cold leg break with a break diameter of 3.8-inches. The transient that produced the analysis PCT result for Braidwood Station and Byron Station, Units 2, is a cold leg break with a break diameter of3.7-inches. The most limiting ECCS single failure of one ECCS train is assumed in the analyses as identified in Table 1. Control rod drop is modeled for breaks less than 1 square foot assuming a 2 second signal delay time and a 4 second rod drop time. RCP trip is modeled coincident with re<J.ctor trip on the low pressurizer pressure setpoint for LOOP transients.

LOOP is the limiting scenario due to manual operator action to trip the RCPs within 5 minutes for the OPA condition. When the low pressurizer pressure SI setpoint is reached, there is a 50 second delay to account for emergency diesel generator start-up, filling headers, etc., after which safety injection is initiated into the reactor coolant system.

The results of the Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, Region I uncertainty analyses are summarized in Tables 8 and 9, respectively. Tables 8 and 9 show the analysis-of-record PCT result, which is the sum of the uncertainty analysis result plus the impact of the energy redistribution uncertainty error correction. The figures presenting the analysis results correspond to the uncertainty analysis results. The MLO and CWO were confirmed to maintain compliance with the 10 CFR 50.46 acceptance criteria with the error correction. The uncertainty analyses are applicable to fuel with Optimized ZIRLO' cladding and ZIRLO cladding. Tables 8 and 9 separately list the 95/95 MLO results from the Region I uncertainty analyses for Optimized ZIRLO cladding (which applies to fuel assemblies with Optimized ZIRLO cladding) and ZIRLO cladding (which covers fuel assemblies with either ZIRLO cladding or Optimized ZIRLO cladding). The MLO values for the two different cladding materials are presented because the steady-state corrosion is different for ZIRLO cladding versus Optimized ZIRLO cladding. The sampled decay heat uncertainty multipliers for the Region I analysis cases are provided in Tables 16 and 17.

Tables 10 and 11 contain sequences of events for the transients that produced the Region I analysis PCT results for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, respectively. Figures 1 through 13 (Braidwood Station and Byron Station, Units 1) and 28 through 40 (Braidwood Station and Byron Station, Units 2) illustrate the calculated key transient response parameters for these transients.

Optimized ZIRLO and ZIRLO are trademarks or registered trademarks of Westinghouse Electric Company LLC;'its affiliates and/or its subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

Westinghouse Non-Proprietary Class 3 CAE-J 9-26/CCE-19-25 Attachment 3 Page 11 of 84 4.0 REGION II ANALYSIS 4.1 Description of Representative Transient A large-break LOCA transient can be divided into phases in which specific phenomena are occurring. A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the fuel assemblies undergo. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below.

Blowdown - Critical Heat Flux (CHF) Phase In this phase, the break flow is subcooled, the discharge rate of coolant from the break is high, the core flow reverses, the fuel rods go through departure from nucleate boiling (DNB), and the cladding rapidly heats up and the reactor is shut down due to the core voiding.

The regions of the RCS with the highest initial temperatures (upper core, upper plenum, and hot legs) begin to flash during this period. This phase is terminated when the water in the lower plenum and downcomer begins to flash. The mixture level swells and a saturated mixture is pushed into the core by the intact loop RCPs, still rotating in single-phase liquid. As the fluid in the cold leg reaches saturation conditions, the discharge flow rate at the break decreases significantly.

Blowdown - Upward Core Flow Phase Heat transfer is increased as the two-phase mixture is pushed into the core. The break discharge rate is reduced because the fluid becomes saturated at the break. This phase ends as the lower plenum mass is depleted, the fluid in the loops become two-phase, and the RCP head degrades.

Blowdown - Downward Core Flow Phase The break flow begins to dominate and pulls flow down through the core as the RCP head degrades due to increased voiding, while liquid and entrained liquid flows also provide core cooling. Heat transfer in this period may be enhanced by liquid flow from the upper head. Once the system has depressurized to less than the accumulator cover pressure, the accumulators begin to inject cold water into the cold legs.

During this period, due to steam upflow in the downcomer, a portion of the injected ECCS water is bypassed around the downcomer and sent out through the break. As the system pressure continues to decrease, the break flow and consequently the downward core flow are reduced. The system pressure approaches the containment pressure at the end of this last period of the blowdown phase.

During this phase, the core begins to heat up as the system approaches containment pressure, and the phase ends when the reactor vessel begins to refill with ECCS water.

Refill Phase The core continues to heat up as the lower plenum refills with ECCS water. This phase is characterized by a rapid increase in fuel cladding temperature at all elevations due to the lack of liquid and steam flow in the core region. The water completely refills the lower plenum and the refill phase ends. As ECCS water enters the core, the fuel rods in the lower core region begin to quench and liquid entrainment begins, resulting in increased fuel rod heat transfer.

Reflood Phase During the early reflood phase, the accumulators begin to empty and nitrogen is discharged into the RCS.

The nitrogen surge forces water into the core, which is then evaporated, causing system re-pressurization and a temporary reduction of pumped ECCS flow; this re-pressurization is illustrated by the increase in RCS pressure. During this time, core cooling may increase due to vapor generation and liquid entrainment, but conversely the early reflood pressure spike results in loss of mass out through the broken cold leg.

I Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page 12 of84 The pumped ECCS water aids in the filling of the down comer throughout the reflood period. As the quench front progresses further into the core, the PCT elevation moves increasingly higher in the fuel assembly.

As the transient progresses, continued injection of pumped ECCS water refloods the core, effectively removes the reactor vessel metal mass stored energy and core decay heat, and leads to an increase in the reactor vessel fluid mass. Eventually the core inventory increases enough that liquid entrainment is able to quench all the fuel assemblies in the core.

4.2 Analysis Results The Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Region II analyses were performed in accordance with the NRC-approved methodology in Reference 1, with exceptions identified under Limitation and Condition Number 2 in Section 2.3. The analyses were performed assuming both LOOP and OPA, and the results of both of the LOOP and OPA analyses for both Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, are compared to the 10 CFR 50.46 acceptance criteria.

The most limiting ECCS single failure of one ECCS train is assumed in the analyses as identified in Table

1. The results of the Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, Region II LOOP and OPA uncertainty analyses are summarized in Tables 8 and 9 for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, respectively. Tables 8 and 9 show the analysis-of-record PCT result, which is the sum of the uncertainty analysis result plus the impact of the energy redistribution uncertainty error correction. The figures presenting the analysis results correspond to the uncertainty analysis results. The MLO and CWO were confirmed to maintain compliance with the 10 CFR 50.46 acceptance criteria with the error correction. The uncertainty analyses are applicable to fuel with Optimized ZIRLO cladding and ZIRLO cladding. Tables 8 and 9 separately list the 95/95 MLO results from the Region II uncertainty analyses for Optimized ZIRLO cladding (which applies to fuel assemblies with Optimized ZIRLO cladding) and ZIRLO cladding (which covers fuel assemblies with either ZIRLO cladding or Optimized ZIRLO cladding). The MLO values for the two different cladding materials are presented because the steady-state corrosion is different for ZIRLO cladding versus Optimized ZIRLO cladding. The sampled decay heat uncertainty multipliers for the Region II analysis cases are provided in Tables 16 and 17.

Tables 12 and 13 contain the sequence of events for the Braidwood Station and Byron Station, Units 1, transients that produced the limiting analysis PCT result for the OPA and LOOP offsite power assumptions, respectively. Figures 14 through 27 illustrate the key response parameters for the limiting LOOP transient for Braidwood Station and Byron Station, Units 1.

Tables 14 and 15 contain the sequence of events for the Braidwood Station and Byron Station, Units 2, transients that produced the limiting analysis PCT result for the OPA and LOOP offsite power assumptions, respectively. Figures 41 through 54 illustrate the key response parameters for the limiting OPA transient for Braidwood Station and Byron Station, Units 2.

The containment pressure is calculated for each LOCA transient in the analyses using the COCO code (References 10 and 11). The COCO containment code is integrated into the WCOBRA/TRAC-TF2 thermal-hydraulic code. The transient-specific mass and energy releases calculated by the thermal-hydraulic code at the end of each timestep are transferred to COCO. COCO then calculates the containment pressure based on the containment model (the inputs are summarized in Tables 2, 3 and 4) and the mass and energy releases, and transfers the pressure back to the thermal-hydraulic code as a boundary condition at the break, consistent with the methodology in Reference 1. The containment model for COCO calculates a conservatively low containment pressure, including the effects of all the installed pressure reducing systems and processes such as assuming all trains of containment spray are operable and assuming fan cooler operation. The containment backpressure curves for the transients that produced

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-I 9-25 Attachment 3 Page 13 of 84 the analysis PCT results are provided in Figures 21 and 48 for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, respectively.

5.0 COMPLIANCE WITH 10 CFR 50.46 It must be demonstrated that there is a high level of probability that the following criteria in 10 CFR 50.46 are met for the analyses for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2:

(b)(l) For each analysis, the analysis PCT corresponds to a bounding estimate of the 95th percentile PCT at the 95-percent confidence level. Since the PCT results are less than 2,200°F, the analyses with the FSLOCA EM confirm that 10 CFR 50.46 acceptance criterion (b)(1 ), i.e.,

"Peak Cladding Temperature shall not exceed 2200°F," is demonstrated.

The results are shown in Tables 8 and 9 for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, respectively.

(b)(2) For each analysis, the analysis MLO corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the MLO results are less than 17 percent when converting the time-at-temperature to an equivalent cladding reacted using the Baker-Just correlation and adding the pre-transient corrosion, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Maximum Local Oxidation of the cladding shall not exceed 17 percent," is demonstrated.

The results are shown in Tables 8 and 9 for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, respectively.

(b)(3) For each analysis, the analysis CWO corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. Since the CWO results are less than 1 percent, the analyses confirm that 10 CFR 50.46 acceptance criterion (b)(3), i.e., "Core-Wide Oxidation shall not exceed 1 percent," is demonstrated.

The results are shown in Tables 8 and 9 for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, respectively.

(b)( 4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains in a coolab le geometry.

This criterion is met by demonstrating compliance with criteria (b)(l), (b)(2), and (b)(3), and by assuring that fuel assembly grid deformation due to combined LOCA and seismic loads is specifically addressed. Criteria (b )(1 ), (b )(2), and (b )(3) have been met for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, as shown in Tables 8 and 9.

It is discussed in Section 32.l of the NRC-approved FSLOCA EM (Reference 1) that the effects ofLOCA and seismic loads on the core geometry do not need to be considered unless fuel assembly grid deformation extends beyond the core periphery (i.e., deformation in a fuel assembly with no sides adjacent to the core baffle plates). Inboard grid deformation due to combined LOCA and seismic loads is not calculated to occur for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2. The FSLOCA EM analyses did not invalidate the existing seismic/LOCA analysis.

(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided following the successful initial operation of the ECCS.

--~--

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 14 of 84 Long-term cooling is dependent on the demon.strati on of the continued delivery of cooling water to the core. The actions that are currently in place to maintain long-term cooling are not impacted by the application of the NRC-approved FSLOCA EM (Reference 1).

Based on the analysis results for Region I and Region II presented in Tables 8 and 9 for Braidwood Station and Byron Station, Units 1, and Braidwood Station and Byron Station, Units 2, respectively, it is concluded that Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, comply with the criteria in 10 CFR 50A6.

6.0 REFERENCES

1. "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," WCAP-16996-P-A, Revision 1, November 2016.

2. "Westinghouse Performance Analysis and Design Model (PADS)," WCAP-17642-P-A, Revision 1, November 2017.

3. "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50, Federal Register, Volume 39, Number 3, January 1974.

4. "Information Notice 98-29: Predicted Increase in Fuel Rod Cladding Oxidation," USNRC, August 1998.

5. "U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2017," LTR-NRC-18-30, July 2018.

6. "Emergency Core Cooling Systems: Revisions to Acceptance Criteria," Federal Register, V53, N180, pp. 35996-36005, September 1988. *

7. "Best Estimate Calculations of Emergency Core Cooling System Performance," Regulatory Guide 1.157, USNRC, May 1989.

8. "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," WCAP-16009-P-A, January 2005.

9. "Transient and Accident Analysis Methods," Regulatory Guide 1.203, USNRC, December 2005.

10. "Westinghouse Emergency Core Cooling System Evaluation Model- Summary," WCAP-8339, June 1974.

11. "Containment Pressure Analysis Code (COCO)," WCAP-8327, June 1974.

12. '"Information to Satisfy the FULL SPECTRUM LOCA (FSLOCA) Evaluation Methodology Plant Type Limitations and Conditions for 4-loop Westinghouse Pressurized Water Reactors (PWRs)' (Proprietary/Non-Proprietary)," LTR-NRC-18-50, July 2018.

13. "U.S. Nuclear Regulatory Commission 10 CFR 50.46 Annual Notification and Reporting for 2018," LTR-NRC-19-6, February 2019.

~I Westingh9use Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 I I

Attachment 3 Page 15 of84 Table 1. Plant Operating Range Analyzed and Key Parameters for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 Parameter As-Analyzed Value or Range 1.0 Core Parameters a) Core power :S 3658 MWt +/- 0% Uncertainty b) Fuel type l 7xl 7 OFA, ZIRLO Clad with IFMs /

Optimized ZIRLO High Performance Cladding Material with IFMs, Integral Fuel Burnable Absorber (IFBA) Rods c) Maximum total core peaking factor (FQ), 2.6 including uncertainties d) Maximum hot channel enthalpy rise 1.7 peaking factor (Filli), including uncertainties e) Axial flux difference (AFD) band at 100% +12/-17%

power f) Maximum transient operation fraction 1.0 2.0 Reactor Coolant System Parameters a) Thermal design flow (TDF) 92,000 gpm/loop b) Vessel average temperature (TAvG) 565°F :S TAvG:S 598°F c) Pressurizer pressure (PRcs) 2207 psia :S PRcs :S 2293 psia d) Reactor coolant pump (RCP) model and Model 93A, 7000 hp power 3.0 Containment Parameters a) Containment modeling Region I: Constant pressure equal to initial containment pressure Region II: Calculated for each transient using transient-specific mass and energy releases and the information in Tables 2, 3, and 4 4.0 Steam Generator (SG) and Secondary Side Parameters a) Steam generator tube plugging level :S 15%

b) Main steam safety valve (MSSV) nominal Table 7 set pressures, uncertainty and accumulation

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 16 of 84 Table 1. Plant Operating Range Analyzed and Key Parameters for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 Parameter As-Analyzed Value or Range c) Main feedwater temperature Units 1:

A nominal value of 441.1 °P was modeled based on a range of 433-449.2°P.

Units 2:

A nominal value of 442.1 °P was modeled based on a range of 435-449.2°F.

d) Auxiliary feedwater temperature A nominal value of 113 .5°P was modeled.

e) Auxiliary feedwater flow rate 140 gpm/SG 5.0 Safety Injection (SI) Parameters a) Single failure configuration ECCS: Loss of one train of pumped ECCS Region II containment pressure: All containment spray trains are available b) Safety injection temperature (Tsr) 32°P:::; Tsr:::; 120°P c) Low pressurizer pressure safety injection 1715 psia safety analysis limit d) Initiation delay time from low pressurizer  :::; 37 seconds (OPA) or:::; 50 seconds pressure SI setpoint to full SI flow (LOOP) e) Safety injection flow Minimum flows in Table 5 (Region I) or Table 6 (Region II) 6.0 Accumulator Parameters a) Accumulator temperature (TAcc) 60°P :::; TAcc :::; 120°P 3 3 b) Accumulator water volume (VAce) 920 ft  :::; V ACC :::; 980 ft c) Accumulator pressure (P Ace) 586.3 psig:::; P ACC:::; 662.3 psig d) Accumulator boron concentration 2". 2200 ppm 7.0 Reactor Protection System Parameters a) Low pressurizer pressure reactor trip signal  :::; 2 seconds processing time

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 17 of84 Table 1. Plant Operating Range Analyzed and Key Parameters for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 Parameter As-Analyzed Value or Range b) Low pressurizer pressure reactor trip 1857 psia setpoint

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page 18 of84 Table 2. Containment Data Used for Region II Calculation of Containment Pressure for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 Parameter Value Maximum containment net free volume 3. lxl06 ft:3 Minimum initial containment temperature at full power operation 60°F Refueling water storage tank (RWST) temperature for containment spray 32°F :S RWST Temp :S 120°F Minimum RWST temperature for broken loop spilling SI 32°F Minimum containment outside air / ground temperature -25°F Minimum initial containment pressure at normal full power operation -0.5 psig Minimum containment spray pump initiation delay from containment ~ 0 seconds (OPA) or~ 0 high pressure signal time seconds (LOOP)

Maximum containment spray flow rate from all pumps 9600 gpm Maximum number of containment fan coolers in operation during LOCA 4 transient Minimum fan cooler initiation delay time ~ 0 seconds (OPA) or~ 15 seconds (LOOP)

Maximum heat removal rate per fan cooler as a function of containment Table 3 temperature Maximum number of containment venting lines (including purge lines, 2 pressure relieflines or any others) which can be OPEN at onset of transient at full power operation Maximum effective valve diameter of each containment venting line 8 inches Maximum containment pressure setpoint for venting valve closure 4.6 psig Maximum delay time between reaching containment pressure setpoint 2 sec and start of venting valve closure Maximum venting valve closure time at normal full power operation 5 sec Containment walls / heat sink properties Table 4 SI spilling flows 265.7 lbm/sec

• Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 19 of 84 Table 3. Fan Cooler Performance Data Used for Region II Calculation of Containment Pressure for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 Containment Temperature Heat Removal Rate Heat Removal Rate (OF) (MBTU/hr) (BTU/sec) 50 4.392 1220 100 14.52 4033 110 18.71 5197 120 23.75 6596 130 29.75 8264 160 54.33 15092 190 87.97 24437 220 124.9 34695 250 159.7 44360 271 182.3 50626 Table 4. Containment Heat Sink Data Used for Region II Calculation of Containment Pressure for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 Wall Area (ft2) Thickness (ft) Material 1 283617.0 0.025 Steel 2 1154.7 0.1325 Steel 3 29719.0 0.00714 Steel 4 20411.0 0.00390 Steel 5 892.5 0.20833 Steel 6 782.25 0.23958 Steel 7 1107.75 0.1250 Steel 8 906.09 0.1040 Steel 9 42144.9 0.04167 Steel 10 40000.0 0.01040 Steel 11 531.82 0.1667 Steel 12 131.67 0.1875 Steel 13 101391.04 0.50 Concrete 0.50 Concrete

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 20 of 84 14 14766.67 0.50 Concrete 0.50 Concrete 15 828.13 0.03362 Steel 1.0 Concrete 16 1850.2 0.02292 Steel 1.0 Concrete 17 10134.8 0.01563 Steel 1.0 Concrete 18 23489.55 0.04899 Steel 1.0 Concrete 19 3022.63 0.15276 Steel 1.0 Concrete 20 115872.75 0.02083 Steel 4.50 Concrete 21 35400.0 0.00390 Steel 22 714.0 0.00390 Steel 23 30000.0 0.01042 Steel Table 5. Safety Injection Flow Used for Region I Calculation for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 High Head Safety Intermediate Head Low Head Safety Pressure (psia) Injection (HHSI) Flow Safety Injection (IBSI) Injection (LHSI) Flow (gpm) Flow (gpm) (gpm) 14.7 279 403 0 34.7 277 401 0 54.7 276 398 0 74.7 275 395 0 94.7 273 393 0 106.9 272 391 0 120.0 276 390 0 214.7 263 374 0 314.7 251 353 0 414.7 238 335 0

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 21 of 84 514.7 226 317 0 614.7 213 297 0 714.7 200 277 0 814.7 186 256 0 914.7 172 229 0 1014.7 158 199 0 1114.7 143 161 0 1214.7 126 111 0 1235.0 124 96 0 1300.0 113 47 0 1300.01 113 0 0 1314.7 110 0 0 1399.7 96 0 0 1400.0 69 0 0 1414.7 67 0 0 1514.7 47 0 0 1614.7 25 0 0 1614.71 0 0 0 1710.0 0 0 0

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page 22 of84 Table 6. Safety Injection Flow Used for Region II Calculation for Braidwood Station, Units 1.and 2, and Byron Station, Units 1 and 2 Intermediate Head Plus Intermediate Head Plus High Head Safety Low Head Safety Injection Low Head Safety Injection Pressure (IBSI and LHSI) Flow (IBSI and LHSI) Flow Injection (IIlISI) Flow (psia)

(gpm) With Miniflow Open With Miniflow Closed (gpm) (gpm) 14.7 278.35 2725.55 2853.8 34.7 276.45 2188.8 2345.55 54.7 275.5 1645.4 1829.7 74.7 273.6 1101.05 1284.4 94.7 272.65 602.3

• 833.15 106.9 271.7 390.45 513.95 120.0 274.55 383.8 383.8 214.7 263.15 362.9 362.9 314.7 250.8 340.1 340.1 414.7 238.45 316.35 316.35 414.8 0.0 0.0 0.0 3000.0 0.0 0.0 0.0 Table 7. Steam Generator Main Steam Safety Valve Parameters for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2 Stage Set Pressure (psig) Uncertainty(%) Accumulation (psi) 1 1175 5 5 2 1190 5 5 3 1205 5 5 4 1220 5 5 5 1235 5 5

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 23 of84 Table 8. Braidwood Station and Byron Station, Units 1: Analysis Results with the FSLOCA EM Region I Value Region II Value Region II Value Outcome (OPA) (LOOP) 95/95 PCT 1164+17 = 1181°P 1610+31 = 1641 °P 1612+31 = 1643°P 95/95 MLO 6.7% 6.7% 6.7%

(Optimized ZIRLO cladding) 95/95 MLO 10.7% 10.6% 10.6%

(ZIRLO cladding) 95/95 cwo 0% 0.12% 0.12%

Table 9. Braidwood Station and Byron Station, Units 2: Analysis Results with the FSLOCA EM

~egion I Value Region II Value Region II Value Outcome (OPA) (LOOP) 95/95 PCT 1168+1 = 1169°P 1721+31 = 1752°P 1680+3 l = 1711 °P 95/95 MLO 6.5% 6.7% 6.7%

(Optimized ZIRLO cladding) 95/95 MLO 10.4% 10.6% 10.6%

(ZIRLO cladding) 95/95 cwo 0% 0.19% 0.15%

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 24 of84 Table 10. Braidwood Station and Byron Station, Units 1: Sequence of Events for Region I Analysis PCT Transient Event Time after Break (sec)

Start of Transient 0.0 Reactor Trip Signal 13.4 Safety Injection Signal 26.0 Safety Injection Begins 76.0 Loop Seal Clearing Occurs 520 Top of Core Uncovered 760 Accumulator Injection Begins 1200 PCT Occurs 1205 Top of Core Recovered 1515 Table 11. Braidwood Station and Byron Station, Units 2: Sequence of Events for Region I Analysis PCT Transient Event Time after Break (sec)

Start of Transient 0.0 Reactor Trip Signal 23.7 Safety Injection Signal 38.8 Safety Injection Begins 88.8 Loop Seal Clearing Occurs 560 Top of Core Uncovered 750 Accumulator Injection Begins 966 PCT Occurs 974 Top of Core Recovered 1000

Westinghouse Non-Proprietary Class .3 CAE-19-26/CCE-19-25 Attachment 3 Page 25 of84 Table 12. Braidwood Station and Byron Station, Units 1: Sequence of Events for Region II Analysis PCT Transient, Offsite Power Available (OPA)

Event Time after Break (sec)

Start of Transient 0.0 Fuel Rod Burst Occurs 2.7 PCT Occurs 4.1 Safety Injection Signal 5.4 Accumulator Injection Begins 12.0 End of Blowdown 23.0 Bottom of Core Recovery 33.0 Safety Injection Begins 42.4 Accumulator Empty 56.0 LHSI Switchover to Miniflow Closed 57.4 All Rods Quenched 150 Table 13. Braidwood Station and Byron Station, Units 1: Sequence of Events for Region II Analysis PCT Transient, Loss of Offsite Power (LOOP)

Event Time after Break (sec)

Start of Transient 0.0 Fuel Rod Burst Occurs 4.9 Safety Injection Signal 5.7 PCT Occurs 9.1 Accumulator Injection Begins 10.0 End of Blowdown 15.0 Bottom of Core Recovery 32.0 Accumulator Empty 55.0 Safety Injection Begins 55.7 LHSI Switchover to Miniflow Closed 70.7 All Rods Quenched 200

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 26 of 84 Table 14. Braidwood Station and Byron Station, Units 2: Sequence of Events for Region II Analysis PCT Transient, Offsite Power Available (OPA)

Event Time after Break (sec)

Start of Transient 0.0 Fuel Rod Burst Occurs 3.9 Safety Injection Signal 5.8 Accumulator Injection Begins 12.0 End ofBlowdown 16.0 Bottom of Core Recovery 32.0 Safety Injection Begins 42.8 LHSI Switchover to Miniflow Closed 57.8 Accumulator Empty 60.0 PCT Occurs 107 All Rods Quenched 260 Table 15. Braidwood Station and Byron Station, Units 2: Sequence of Events for Region II Analysis PCT Transient, Loss of Offsite Power (LOOP)

• Event Time after Break (sec)

Start of Transient 0.0 Fuel Rod Burst Occurs 4.3 Safety Injection Signal 5.8 Accumulator Injection Begins 9.0 End ofBlowdown 14.0 Bottom of Core Recovery 32.0 PCT Occurs 35.0 Safety Injection Begins 55.8 Accumulator Empty 57.0 LHSI Switchover to Miniflow Closed 70.8 All Rods Quenched 200

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 27 of84 Table 16. Braidwood Station and Byron Station, Units 1: Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases Region Case DECAY_HT (units of o-) DECAY_HT (absolute unitsi1>

PCT +0.3720cr 1.85%

Region I ML0C3> +1.2693cr 6.49%

cwo N/A<2> '

N/A<2l PCT +0.1008cr 0.48%

Region II ML0<3> +0.6052cr 3.09%

(OPA) cwo +0.1973cr 0.96%

PCT +0.0959cr 0.47%

Region II 3 ML0< > +0.6052cr 3.09%

(LOOP) cwo +0.3736cr 1.79%

Notes:

1. Approximate uncertainty in total decay heat power at 1 second after shutdown as defined by the ANSVANS-5.1-1979 decay heat standard for 235 U, 239Pu, and 238 U assuming infinite operation.

2. No decay heat uncertainty value is provided for the SBLOCA (Region I) CWO case since the analysis result for all runs is 0.0%.

3. The decay heat uncertainty values for the MLO cases are applicable to both the ZIRLO and Optimized ZIRLO cladding results.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 28 of84 Table 17. Braidwood Station and Byron Station, Units 2: Sampled Value of Decay Heat Uncertainty Multiplier, DECAY_HT, for Region I and Region II Analysis Cases Region Case DECAY- HT (units of a) '

DECAY_HT (absolute unitsi 1>

PCT +0.9811cr 4.96%

Region I ML0(3) + 1.2172cr 6.21%

cwo N/A<2> N/A<2>

PCT +0.7415cr 3.56%

Region II ML0<3> +0.3640cr 1.86%

(OPA) cwo + 1.586Icr 7.60%

PCT +0.1002cr 0.49%

Region II ML0<3> +0.3640cr 1.86%

(LOOP) cwo +0.6586cr 3.17%

Notes:

1. Approximate uncertainty in total decay heat power at 1 second after shutdown as defined by the ANSI/ANS-5.1-1979 decay heat standard for 235 0, 239Pu, and 238 0 assuming infinite operation.

2. No decay heat uncertainty value is provided for the SBLOCA (Region I) CWO case since the analysis result for all runs is 0.0%.

3. The decay heat uncertainty values for the MLO cases are applicable to both the ZIRLO and Optimized ZIRLO cladding results.

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-1 9-25 Attachment 3 Page 29 of 84 Void rroction in the Cell Ad jocen t to the Brea k 0.8 ...... . ................. ... ........ .

0-6 * *

• p * * * * * * * * * *

• C:

0 e

u..

2 0

0.4 * * * * * * * * * * * * * -

• 0.2 'I 'I** 0 0 0 0 IO t IO* 'I ** a *

• 0 0 I* 0 IO I* 4 **I IO I> 0 + I 4 O O I a I* I Io It Io t I If I Q-1----,L'""""'""'-..LI.----r---'--L...--'-----,L-T--'--'--'----'--..----'---L-...,____._~

0 500 1000 1500 2000 Time After Break (sec)

Figure 1: Braidwood Station and Byron Station, Units 1: Break Flow Void Fraction for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE-19-25 Attachment 3 Page 30 of 84 Total Safety Injection Flow T otal B reak Flow 2000

- (I)

- E 1500

...0 Cl)

-+J 0

a:::

3:: I 0

Ci: I l . . . . . . . . . . . . . : ...... .. ...... : . . . . . . . . . ; .... : . . . . . . . . . . . . . .

cn (I) 1000 I

0

t I

~

\

- 'I I "I 500 ..... ... : ..... : . .... . . ....... . ... .. ......... .. .... . ....... .

..... y _ -  :

~

500 1000 1500 2000 Time After Break (sec)

Figure 2: Braidwood Station and Byron Station, Units 1: Total Safety Injection Flow (not Including Accumulator Flow) and Total Break Flow for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- 19-25 Attachment 3 Page 3 1 of 84 RCS P ressure 2000 c, 1500

• ci5 0....

500 . ************ *************** ************* -~- **********

o -+-__.__.___......___._--,,--....____.___.._......_-r-__.__.___......___.____,.--.....____.___,.____.__~

0 500 1000 1500 2000 Time After Break (sec)

Figure 3: Braidwood Station and Byron Station, Units 1: RCS Pressure for the Region I Analysis PCT Case

l Westinghouse on-Proprietary Class 3 CAE-19-26/CCE- I 9-25 Attachment 3 Page 32 of84 Ho t Assembly Two Ph a se Mixture Level 10 * * * * * * * * * * * * * * * * * * * * * * *

• 8

-:=

~

~ 6 Q)

~

X

E 4

2 .............. .. ... .. ........ . .............. . .. ' .......... .

0 0 500 1000 1500 2000 Time After Break {sec)

Figure 4: Braidwood Station and Byron Station, Units 1: Hot Assembly Two-Phase Mixture Level (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case

Westi nghouse Non-Proprietary Class 3 CAE- 19-26/CCE-19-25 Attachment 3 Page 33 of 84 Dummy Rod 1 Du mmy Rod 2 Hot Ro d Hot As .s em b ly Average Rod 1 Aver a ge Rod 2 Low Power Rod 1200 400 0 500 1000 1500 2000 Time After Break (sec)

Figure 5: Braidwood Station and Byron Station, Units 1: Peak Cladding Temperature for all Rods for the Region I Analysis PCT Case

Westinghouse on- Proprietary Class 3 CAE- 19-26/CCE-19-25 Attachment 3 Page 34 of84 Crossover leg 1 Crossover leg 2 Crossover Leg J Crossover leg -4 250

- 200 . . .

I' *'

VJ E

.Q Cl) 150 ***** ***** ****i i,~ * * * *

• 1* * * *

• t*******i**************

-+J 0

a:::

1 1

1111'1*\,

r *

• 'I 3:

0 Ci: 100 Cl)

Cl) 0

I

.............. 11* ............ : ....

~:~\~\ViJ,\}~.-*.**.*.

.ii:  :,

.  : ,, ~

50 . .

-1

l

I

I Q4----'~"-----.___ _ _ _ _ _ ___,.

-so-+-_.~-----......-----......~-------~-------~..............

0 500 1000

~.------~-------1 1500 2000 Time After Break (sec)

Figure 6: Braidwood Station and Byron Station, Units 1: Vapor Mass Flow Rate through the Crossover Legs for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- 19-25 Attachment 3 Page 35 of84 Hot Assembly Channe l Non-Gu i de Tube Average Channel Gu i de Tube Aver a ge Channel Low Power Chonnel 10 o+J

~

~

2 CT 8

.::i

""O (1)

~

Cl..

..S2 0

u 6

4------------------------------------1 0 500 1000 1500 2000 Time After Break (sec)

Figure 7: Braidwood Station and Byron Station, Units 1: Core Collapsed Liquid Levels (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- I 9-25 Attachment 3 Page 36 of 84 Loop 2 Accumula t o r Injection Loop J Ac c umu I o t o r I n ! e c t i on Loop 4 Accumulator lnJeclion 300------------------------------,,

250

- Cl) 200 E

..0

..::=.-

Q)

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0

~ 150

~

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en en 0

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100 50 . . ... . . . ...... * * ..... ........ . .....

0......~___.--'---'---,.-'--"""--_..__._ __,__--"'.........i.....Ll..111..J~--a.&......a....iL....1..ILJ."'---1 0 500 1000 1500 2000 Time After Break (sec)

Figure 8: Braidwood Station and Byron Station, Units 1: Accumulator Injection Flow for the Region I Analysis PCT Case

Westinghouse on-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 37 of 84 Vessel Fluid Moss 200000 18()()00

(/)

(/)

0

E 140000 ******

120000 * * * * * * * * * * * *

• 100000 8()()00 0 500 1000 1500 2000 Time After Break (sec)

Figure 9: Braidwood Station and Byron Station, Units 1: Vessel Fluid Mass for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page 38 of 84 so Secondary Si de 1

---. SG S e cond a ry Si d e 2 SG Second ar y Si d e 3

---

SG Secondary Side 4 1300 I ',

.. ,.. I '

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a... 1000 900 aoo-+-.......-------....---__._....__.._....._ ........__.___.__...___,.._______..._....

0 500 1000 1500 2000 Time After Break (sec)

Figure 10: Braidwood Station and Byron Station, Units 1: Steam Generator Secondary Side Pressure for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE-19-25 Attachment 3 Page 39 of84 Hot Rod Hot Assembly Rod Aver o ge Rod 25 2  :',

\

\

\

\

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L Q) .

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Figure 11: Braidwood Station and Byron Station, Units 1: Normalized Core Power Shapes for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- I9-25 Attachment 3 Page 40 of 84 Relative Core Po wer 1.2 1- ....... ' ...... : ....... .. ..... : ...... ' .. ' ' ... : ......... ' ... .

- 0 C

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l ~ - - :l:-----;l: ------;l; ------;l;---_:._--:l: - - ,l: - - ,I, ---,-~

0 500 1000 1500 2000 Time After Break (sec)

Figure 12: Braidwood Station and Byron Station, Units 1: Relative Core Power for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 4 1 of84 Temperature {f')

Vapor Temperature al Core Outlet Void Fract i on

- - - - - Void f'ractioo o t Core Outlet 1000--------------.*-------,----------------........j----------.......- 1

~

11

~I I

~

J

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E 1\'

I

.I

• I

)1 '"C g

~ I " Q,4 I

600 *) * ........... ..........

0.2 500 400 0 0 500 1000 1500 2000 TI me After Break (sec)

Figure 13: Braidwood Station and Byron Station, Units 1: Vapor Temperature and Void Fraction at Core Outlet (Hot Assembly Channel) for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- I 9-25 Attachment 3 Page 42 of84 Dummy Rod 1 Dummy Rod 2 Hot Rod Hot Assembly Average Rod 1 Average Rod 2 Low Power Rod 1600 1400 *" f 1 I

.~,, *

\

t' 'f~.Xi,,;,.~ ~w-'t.: '

EC

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1200 *t *.. V~... . .... ,~:.,i\ . ....... ...... ................. .

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r,........ . ... .. ................. .

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v.. ':/ ~'fl .._ - -~ ~~11.! 1 ~----~----.dlt 200-+-..__........_.___..._.................____.._........_,............_.........._,._._.........._.._,_..._.___..a....,t 0 50 100 150 2~ 250 300 Time After Break (sec J Figure 14: Braidwood Station and Byron Station, Units 1: Peak Cladding Temperature for all Rods for the Region II Analysis PCT Case, LOOP

Westinghouse on-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attachment 3 Page 43 of 84 PCT Location for Limiting Dummy Rod 10 ***********************************

8 . ... ..... . . . . . . . ... . .

C:

0 6 ......... . .. .

0

<L>

GJ 4

2 .... . .... . .... . .... . .. . ...... . . ...... . . .. ..... . . . . .

o -_.__.__.__._-r-......._......._....._...__..--...__.__.__........,,.....................................__,____.___..__.___..---..--L__.___._--L--f 0 50 100 150 2QO 250 300 Time After Break (sec J Figure 15: Braidwood Station and Byron Station, Units 1: Peak Cladding Temperature Elevation (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attachment 3 Page 44 of84 Vessel-Side Break Flo w 80000 60000

-- en

- E

..0 (l,)

~ 40000 * * * * * * * * * * * * * * .. * * * .. * * * * * * * * * * * * * * .. * * * * * * * * * * * * * * * * * * * * *

~

0 Ci::

en Cl) 0

E 20000

• 0 0 50 100 150 2~ 250 300 Time After Break (sec}

Figure 16a: Braidwood Station and Byron Station, Units 1: Vessel-Side Break Mass Flow Rate for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 45 of 84 Pump-Side Break Flow 30000

~

- E

....0 Q.)

0 a::: 20000

~

0 c:;:

Cl)

Cl) 0

e 10000 o +-..L....Ji:,...i.--L.,~d,,,_L....l,_..._..,.....,__.......,,__._-1-..i.e.....L.....J-..!ol.il~--i....a..t:¥...&..11..C::....&...l.l-"'4 0 50 100 150 2, 250 300 TI me After Break (secJ Figure 16b: Braidwood Station and Byron Station, Units 1: Pump-Side Break Mass Flow Rate for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attachment 3 Page 46 of84 L o wer P lenum Co l lapsed Liqui d Level 20 15 *******

Q)

~

"O

i 10 *****

er

~

5 ....

o--......_.....__.....__............--...___...___....__....__-r--..__..___.__........,,.............................___.__,___.___.__.__.........................___.___._......

0 50 100 150 2QO 250 300 Time After Break (sec J Figure 17: Braidwood Station and Byron Station, Units 1: Lower Plenum Collapsed Liquid Level (Relative to Inside Bottom of Vessel) for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Propr ietary Class 3 CA E- I 9-26/CCE-1 9-25 Attachment 3 Page 47 of 84 Top Cel I of the Non-GT Avg Channel 10 ****

- (/)

- E

...0 Q)

.._J 0

c:::: 5

~

0 G:

(/)

(./)

0

E 0 ' ...... . .... ' .... .

+-__._....___._...._-r-...._...._...._...._-r-. . . . _ ~ ~ ~ - r - - . . _ _..__..__..__-r--..__....._....._....._r--.__.__.__'--I 0 5 10 1s ~o 25 30 Time After Break (sec J Figure 18: Braidwood Station and Byron Station, Units 1: Vapor Mass Flow Rate per Assembly at the Top Cell Face of the Core Average Channel not Under Guide Tubes for the Region Il Analysis PCT Case, LOOP

Westi nghouse Non-Proprietary Class 3 CAE-1 9-26/CCE- I9-25 Attachment 3 Page 48 of 84 RCS P ressure 80 ....

- C,

• ci) 0...

60 ....

~

en en Q) a.. 40 .... . .. .

20 ****** ** ***************************************************

o --_._........_........_....___-r-......_...._..__..__.....-..........___.___.~__.___.__...._..__ _.__._..................~ ................._._....1.......,1 0 50 100 150 2QO 250 300 Time After Break (sec; 2082788341 Figure 19: Braidwood Station and Byron Station, Units 1: RCS Pressure for the Region Il Analysis PCT Case, LOOP

Westi nghouse Non-Proprietary Class 3 CAE-19-26/CCE- I 9-25 Attachment 3 Page 49 of 84 Loop 2 Accumula t o r Injection Loop J Accumulator tniection Loop 4 Accumulator l nJection 1500 * * * * * *

• 500 ..... '

01-~~ ... .......... ' ... . ...... ' ... .. ~:-:-:~ .........--:~~ ~ ~---t

- 500+-.....&...----l.~....._"'T'-"_._-""~l---,---'-_..--,1~'T"'"""........_..--'~r-...._--'---'---1 0 20 40 80 100 Time After Break (sec)

Figure 20: Braidwood Station and Byron Station, Units 1: Accumulator Injection Flow per Loop for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 50 of &4 Containment Pressure 30 0

'ci)

CL e 2s

J (I)

(I)

Q) lo...

a..

20 ***********************************************************

15-1--~~~...........~~~~-,-~~~.__,.~~~~-.-~~~........,..---~~~--1 0 50 100 150 2QO 250 300 Time After Break (sec J Figure 21: Braidwood Station and Byron Station, Units 1: Containment Pressure for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- I 9-25 Attachment 3 Page 51 of 84 Vessel Fluid Moss 200000 150000 E

..Q (I)

(I) 0

IE 100000 50000

• 0 0 50 100 150 200 250 300 Time After Break (sec) 2082788J.41 Figure 22: Braidwood Station and Byron Station, Units 1: Vessel Fluid Mass for the Region II Analysis PCT Case, LOOP

We tinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- I 9-25 Attach ment 3 Page 52 of 84 Hot Assembly Channel Non-Guide Tube Average Chonnel Guide Tube Average Cnonnel Low Power Channel 10 ***********************************************************

4 .. .... .

2 0~  :.:Ll,,,iiLo....aL.Ll.--,.....L.....L.......L......I............L....L---L....L-.--...L.....L...L......l'"--r---L....L.....L.......L...-,--.L.....JL....L......L~

0 50 100 150 2QO 250 300 Time After Break (sec J Figure 23: Braidwood Station and Byron Station, Units 1: Collapsed Liquid Level for Each Core Channel (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case, LOOP

Westinghouse on-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attachment 3 Page 53 of 84 Average Oowncomer Col lapsed Liquid Level 30 ...................... . ............... .. ............... . .. .

25 Q)

~ 20

-c, c:T

.::J

-0 v

(I) 15 ******

a..

.2 0

u 10 * * *

• 5 .

o -+-...__.,__....__L.....,,--L--'-...................-r-_..__......_.,__L.....,__.__.___.___._......._._...__......._.,__r--'L.......L.........__._-1 0 50 100 150 2QO 250 300 Time After Break (secJ Figure 24: Braidwood Station and Byron Station, Units 1: Average Downcomer Collapsed Liquid Level (Relative to the Bottom of the Upper Tie Plate) for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 54 of84 Loop 2 Sofety Injection Loop J So f ety ln~ection Loop 4 Sofely lnJection 100

- (I')

E

..0

..::=,..

Q) 0 0:: 50 3:

0

• LL Cl)

Cl)

C

ii:

o-------' ............. ..... . .. . .. .... . . . ....... . ...... . . .

-~ ..................._........,_.__.__._...._...-.._..............._,._.__._.....__._.,.........................._,.-.-------1 0 50 100 150 2QO 250 300 Time After Break (sec; Figure 25: Braidwood Station and Byron Station, Units 1: Total Safety Injection Flow Rate per Loop (not including Accumulator Flow) for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page 55 of 84 Hot Rod Hot Assembly Rod Aver a ge Rod I'  : ......

2 .. ' ... "./

./

I.

I: '\ -

~

I ' . \

1......

Q.)

I= 1.5 0

a..

"'O Q.)

N 0 ,, .

E I.... ,I I

I 0 .,.

z I I

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I I

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05 *,I .

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o.........................__--'---l~,...._r-__-'--.........-y--'---'-_.......___._,...._...._...._..._......._s..-1 0 2 4 6 8 10 12 Elevation (ft)

Figure 26: Braidwood Station and Byron Station, Units 1: Normalized Core Power Shapes for the Region II Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 56 of84 Relative Core Po wer 1.2""T""------------------------------,

- CJ C

.E 0,8 0

z:

0 C:

0

~

e

(,.)

I.&..

(\)

3 o.6 0

a...

~ 0,4 0

u 0.2 a********

• o O O O I I I I I I < I I I 1 t I O t t. 1 f 1 1 o o Io o t, I O I I I I a 1 0 *

• O *

• 1 a 1 *

• 0 0 50 100 150 2QO 250 300 Time After Break (sec J 20827W41 Figure 27: Braidwood Station and Byron Station, Units 1: Relative Core Power for the Region Il Analysis PCT Case, LOOP

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- I 9-25 Attachment 3 Page 57 of 84 Void Fraction in the Cell Adjocenl to the Break 0.8 .............. . . .

Q. 6 ............ ... . . . ~

C:

0

~

e LL..

.:-2 0

0.4 * - * * * * * * * * * * - - -

• 0.2 o -+-......1.-..0.::..-.L.JL.......1--,----L.-L--....L..--l.--r--.....L..__._...,____.__.--....,__........_ ..,____,._---4 0 500 1000 1500 2000 Time After Break (sec)

Figure 28: Braidwood Station and Byron Station, Units 2: Break Flow Void Fraction for the Region I Analysis PCT Case

Westinghouse Non- Proprietary Class 3 CAE- 19-26/CCE- I9-25 Attachment 3 Page 58 of 84 Total Safety I nject i on Flow T otal Break F l ow 2000 I **.. ,., *. , *.** : ..********* ,,, : .,, .**** . ..*** : . **** , ********

I I

(/)

(/)

1000 l * * * * * * * * * * * * * : * ** * * * * ** * * * * * : * * * * ** * * * * * * * * : * * * * * * * * * * * * *

• 0

ii: I 1

I

\

\

500 . \~ ~~-r~1"'\ ... .. ... :. ....... ... .:... ....... ... .

• I * *

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--~ _.,,- "'.,-...,. - - - - - - _... ..... -

~

o -i1;;;;;;:::1:::;;;;;;J;;;;;;;;;;1;;:;::::::;::::::::L-1..-1..--1._;..-1.---1...!....1--1L...-;._1--.L......1._..J..--1 0 500 1000 1500 2000 Time After Break (sec)

Figure 29: Braidwood Station and Byron Station, Units 2: Total Safety Injection Flow (not Including Accumulator Flow) and Total Break Flow fo r the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 59 of84 RCS Pressure 2000

'o' 1500

• en a..

500 0

0 500 1000 1500 2000 Time After Break (sec)

Figure 30: Braidwood Station and Byron Station, Units 2: RCS Pressure for the Region I Analysis PCT Case

Westinghouse on-Proprietary Class 3 CAE- l 9-26/CCE- 19-25 Attachment 3 Page 60 of 84 Hot Assembly Two Phase Mixture Level 10 ***********************************************************

8

~

~ 6 Q.)

~

X

~

4 2

o -+-__.._....____.__L...---y----1.-....____._---J,___-r-__.__......__._---J'--r---'--......__._---J'--f 0 500 1000 1500 2000 Time After Break (sec)

Figure 31: Braidwood Station and Byron Station, Units 2: Hot Assembly Two-Phase Mixture Level (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attachment 3 Page 6 1 of84 Dummy Rod 1 Dummy Rod 2 Hot Rod Hot Assembly Average Rod 1 Average Rod 2 Low Power Rod 1200-------------------------,

La..

~

1000

'1

-+-"

C, Cl>

~

E Q,)

1-- .. ... }

t

'/

800 CTI C:

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400 +---'---1-"'--...L.--,..--1-"'---'-......L.--,....-""'---'-......L.......l...._....._...1..--1.-...._...I..--I 0 500 1000 1500 2000 Time After Break (sec)

Figure 32: Braidwood Station and Byron Station, Units 2: Peak Cladding Temperature for all Rods for the Region I Analysis PCT Case

Westinghouse on-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attach ment 3 Page 62 of84 Crossover Leg 1 Crossov e r Leg 2 Crossover leg J Crossover leg -4 250

. .. . . . .. ......  : ,I

(/)

E 200

-.J::l Q)

C, 150 a:::

3:

0 Ci: 100

(/)

(() I C,

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50 I 0

-50+--'-~.__.__.~------'-.............~-----i.~..._........_,,........._......~._----1 0 500 1000 1500 2000 Time After Break (sec)

Figure 33: Braidwood Station and Byron Station, Units 2: Vapor Mass Flow Rate through the Crossover Legs for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-19-25 Attachment 3 Page 63 of84 Hot Assembly Channel Non-G u l de Tube Aver a ge Channel Guide Tube Average Chonnel Low Power Channel 12 10

~

~

-0

• 5 8 CT

=i "t:J Q,)

(I)

Cl..

..2 0

u 6

4-+-__...._..._....___.__ _.............._...._.......__,,_..__.....___.._...._.......__._..__..............- - 1 0 500 1000 1500 2000 Time After Break {sec)

Figure 34: Braidwood Station and Byron Station, Units 2: Core Collapsed Liquid Levels (Relative to Bottom of Active Fuel) for the Region I Analysis PCT Case

Westinghouse on-Proprietary Class 3 CAE- 19-26/CCE- I9-25 Attach ment 3 Page 64 of 84 Loop 2 Accumula t or I nject i on Loop J Accu mutot or l n1ect i on Loop 4 AccumuJo t or l nJeclion 250 - * * * * * * * * * * * * * *: * * * * * * * * * * - * * *: * *

...- 200- * * * * * * * * * * * * * * .* * * * * * * * * * * * * * * *

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150- * * * * * * * * * * * - * * : * * * * * * * * * * * * *

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ii=

100- * * * * * * * * * * * * * *: * * * * * * * * * * * *

• 50- ... . . ... ... - .. : .. ........ .. .

0-+-__.1_.....,1....___.1....___.,....__..,,_...._ ....._ ....._ ...._..."T,-..-__..*...._...,_ .....

1 __,.._...., _ ...., _....1_.._.---t 0 500 1000 1500 2000 Time After Breok (sec)

Figure 35: Braidwood Station and Byron Station, Units 2: Accumulator Injection Flow for the Region I Analysis PCT Case

We ti nghouse on-Proprietary Class 3 CAE-1 9-26/CCE-l 9-25 Attachment 3 Page 65 of 84 Vessel Fluid Moss 220000 200000 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *

.-.. 160000 E

-..c VJ Cl) 0

~ 140000 * * * * * * *

• 120000 100000 800()0 -+---'----"'--....__.___,,--_.___.____..__......__.......-___._...._~___.-,--~----...._~-

o 500 1000 1500 2000 Time After Break (sec) 3512561&5 Figure 36: Braidwood Station and Byron Station, Units 2: Vessel Fluid Mass for the Region I Analysis PCT Case

We ti nghouse Non-Proprietary Class 3 CAE- 19-26/CCE-1 9-25 Attach ment 3 Page 66 of 84 so Second a ry Side 1

-- --- SG

---

- SG Second a ry Second a ry Si de 2 Si de 3

---* SG Secondary Si de 4 1300 1200

. \..1:'- . .

-0 1100 ............... ... . .... ...'"'

f

..*~-x' ............. ..............

• en c..

Q) en en

<1.>

CL 1000 900 aoo-----------------------.. . .---------

0 500 1000 1SOO 2000 Time After Break (sec)

Figure 37: Braidwood Station and Byron Station, Units 2: Steam Generator Secondary Side Pressure for the Region I Analysis PCT Case

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 67 of84 Hot Rod Hot Assembly Rod Average Rod 2

- -~ ._

L Q.)

31: 1.5 0

a..

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I o-+--'---'---1..---........&-..J----L---.---"'---'--...._""T""'.......--'-__._---..........-..11.-..1.-..--..........._-'---f 0 2 4 6 8 10 12 Elevation (ft) 3512!1614,S Figure 38: Braidwood Station and Byron Station, Units 2: Normalized Core Power Shapes for the Region I Analysis PCT Case

Westinghouse on-Proprietary Class 3 CAE- 19-26/CCE- I9-25 Attachment 3 Page 68 of 84 Rel a t i ve Core Po wer 1.2----------------------------- ----.

1- ............................. . ........ ' ............. ' . ' ' .. .

- 0 C

- 0 C:

0

(.)

~

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1_ ...__

1 _ ..__1_..__

1 -i 0 500 1000 1500 2000 Time After Break (sec)

Figure 39: Braidwood Station and Byron Station, Units 2: Relative Core Power for the Region I Analysis PCT Case

Westinghouse on-Proprietary Class 3 CAE- 19-26/CCE- 19-25 Attachment 3 Page 69 of84 Temperature (F) voror Tempe r ature ot Core Outlet Void rroc ion

- - - -

• Void fraction ot Core Outlet 1000....--~~~~~~~~-.-~~~~~~~~~~~~~--r-

,1 11 900 ......... ' . ' * * *.

• II * * * *

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Figure 40: Braidwood Station and Byron Station, Units 2: Vapor Temperature and Void Fraction at Core Outlet (Hot Assembly Channel) for the Region I Analysis PCT Case

Westinghouse on-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 70 of84 Dummy Rod 1 Dummy Rod 2 Hot Ro d Hot Assembly Average Rod 1 Average Rod 2 Low Power Rod

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0 50 100 150 2'{0 250 300 Time After Break (sec) 8!H717151 Figure 41: Braidwood Station and Byron Station, Units 2: Peak Cladding Temperature for all Rods for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE-1 9-25 Attachment 3 Page 71 of84 PCT Locat i on for Limiting Dummy Rod 10 *********** ***************************************

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0 0 50 100 150 2QO 250 300 Time After Break (sec J

~7171~1 Figure 42: Braidwood Station and Byron Station, Units 2: Peak Cladding Temperature Elevation (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case, OPA

Westinghouse on-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page 72 of84 Vessel-Side Break Flow 60000 (f)

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0 50 100 150 2, 250 300 Time After Break (sec) 8"717151 Figure 43a: Braidwood Station and Byron Station, Units 2: Vessel-Side Break Mass Flow Rate for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 73 of84 Pump-Side Break Flow 30000

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0 50 100 150 2, 250 300 Time After Break (sec J 8947171!1 Figure 43b: Braidwood Station and Byron Station, Units 2: Pump-Side Break Mass Flow Rate for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 74 of84 Lower P lenum Col lapsed Liquid Level 15 (l.)

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0 50 100 150 2QO 250 300 Time After Break (sec J 894717151 Figure 44: Braidwood Station and Byron Station, Units 2: Lower Plenum Collapsed Liquid Level (Relative to Inside Bottom of Vessel) for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- 19-25 Attachment 3 Page 75 of84 Top Cell of the Non-OT Avg Channel 10

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-5 0 5 10 1s ~o 25 30 Time Atter Breck {sec) 8M717151 Figure 45: Braidwood Station and Byron Station, Units 2: Vapor Mass Flow Rate per Assembly at the Top Cell Face of the Core Average Channel not U oder Guide Tubes for the Region II Analysis PCT Case, OPA

Westi nghouse Non-Proprietary Class 3 CAE-1 9-26/CCE- l 9-25 Attach ment 3 Page 76 of84 RCS Pressure 100 80 ....

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~7171$1 Figure 46: Braidwood Station and Byron Station, Units 2: RCS Pressure for the Region II Analysis PCT Case, OPA

Westi nghouse Non-Proprietary Class 3 CAE-1 9-26/CCE- I 9-25 Attachment 3 Page 77 of 84 loop 2 Accumula t or Inject i on Loop 3 Accumu l a t or ln,ect!on Loop 4 Accumu l a t o r lnJecl1on 1500 *******.

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~717151 Figure 47: Braidwood Station and Byron Station, Units 2: Accumulator Injection Flow per Loop for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- 19-25 Attachment 3 Page 78 of84 Cont a inment Pressure 30 * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * * *' * * * *

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• 15-1-~~~........,..........~ ~~""T'"""~~~........,.......... ~~~_,...~~~~ ..........~~~-1 0 50 100 150 2QO 250 300 Time After Break (sec J Figure 48: Braidwood Station and Byron Station, Units 2: Containment Pressure for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attachment 3 Page 79 of 84 Vessel Flui d Moss 250000 200000 150000 en en 0

E 100000 *** ***

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Westinghouse Non-Proprietary Class 3 CAE-1 9-26/CCE- I 9-25 Attachment 3 Page 80 of 84 Hot Assembly Channel Non-Gu i de Tube Average Channel Guide Tube Average C~onnel Low Power Chon 'n el 10 ***** ******** ***************** ************** ********* ******

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0 50 100 150 2QO 250 300 Time After Break (secJ a9411m1 Figure 50: Braidwood Station and Byron Station, Units 2: Collapsed Liquid Level for Each Core Channel (Relative to Bottom of Active Fuel) for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attachment 3 Page 81 of84 Aver a ge Downeomer Col lapsed Liquid level 30 - - * * * * * * - * - * * * * * - - * * * * * - - - - * * * - * * * - * * * * * * - - - * * * * - - - - * * * * * - -

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~717151 Figure 51: Braidwood Station and Byron Station, Units 2: Average Downcomer Collapsed Liquid Level (Relative to the Bottom of the Upper Tie Plate) for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE- l 9-25 Attachment 3 Page 82 of84 Loop 2 Sofely lnJection Loop J S ofe t y lnJ e ction Loop 4 Sofety lnJ e ction 100

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0 50 100 150 2QO 250 300 Time After Break (sec J 894717151 Figure 52: Braidwood Station and Byron Station, Units 2: Total Safety Injection Flow Rate per Loop (not including Accumulator Flow) for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE-19-26/CCE-l 9-25 Attach ment 3 Page 83 of 84 Hot Rod Hot Assembly Rod Aver a ge Rod 2 . ... ,,, . . ,,; ..,...

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0 2 4 6 8 10 12 Elevation (ft) 8947171$1 Figure 53: Braidwood Station and Byron Station, Units 2: Normalized Core Power Shapes for the Region II Analysis PCT Case, OPA

Westinghouse Non-Proprietary Class 3 CAE- 19-26/CCE- l 9-25 Attachment 3 Page 84 of84 Relative Core Power 1.2.....-------------------------------,

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0 50 100 150 2QO 250 300 Time After Break (sec J etf7171~1 Figure 54: Braidwood Station and Byron Station, Units 2: Relative Core Power for the Region II Analysis PCT Case, OPA