AEP-NRC-2012-71, Revised Small Break Loss-of-Coolant Accident Analysis

Revised Small Break Loss-of-Coolant Accident Analysis

ML12256A685
Person / Time
Site:	Cook
Issue date:	08/31/2012
From:	Gebbie J Indiana Michigan Power Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
AEP-NRC-2012-71
Download: ML12256A685 (69)

Text

z INDIANA MICHIGAN Indiana Michigan Power POWER' One Cook Place Bridgman, MI 49106 A unit of American Electric Power IndianaMichiganPower, com August 31, 2012 AEP-NRC-2012-71 10 CFR 50.46 Docket No.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Donald C. Cook Nuclear Plant Unit 1 REVISED SMALL BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS

References:

Letter from R. A. Hruby, Indiana Michigan Power Company (I&M), to Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Unit 2, Schedule for Submittal of Revised Unit 1 and Unit 2 Small Break Loss of Coolant Accident Analyses Addressing Residual Heat Removal System Spray Issue," AEP-NRC-2010-30, dated March 29, 2010 (ADAMS Accession Number ML100060923).

This letter provides a revised small break loss of coolant accident (SBLOCA) analysis which addresses a Residual Heat Removal (RHR) system spray diversion issue for Donald C. Cook Nuclear Plant (CNP) Unit 1.

In the referenced letter, I&M described an assumption in the Unit 1 SBLOCA analysis that had been determined to be inconsistent with CNP Emergency Operating Procedure (EOP) provisions to realign the RHR pump discharge to provide containment spray under certain conditions, thus diverting RHR system flow from the Reactor Coolant System (RCS) during the cold leg recirculation phase of the accident. I&M committed to revise the Unit 1 SBLOCA analysis to address the RHR flow diversion issue and transmit the revised analysis to the NRC no later than August 31, 2012.

The RHR flow diversion issue has been addressed by changing the applicable EOPs such that the train cross-tie valves in Safety Injection (SI) System discharge are maintained open during the cold leg recirculation phase. Since each SI pump discharge is piped to two RCS loops if a cross-tie valve is closed, maintaining the cross tie valves open results in increased SI system flow to the RCS if only one SI pump is operating due to a postulated single failure. The increased SI system flow to the RCS compensates for the RHR flow being diverted from the RCS to containment spray.

The Unit 1 SBLOCA analysis has been revised accordingly.

4oz..

U. S. Nuclear Regulatory Commission AEP-NRC-2012-71 Page 2 The enclosure to this letter provides a report of the revised analysis. As described in the enclosed report, the results of the revised SBLOCA analysis conform to the emergency core cooling system acceptance criteria of 10 CFR 50.46. The calculated peak cladding temperature of 1725 0 F is below the 10 CFR 50.46 limit of 22000 F.

There are no new commitments in this submittal. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.

Sincerely, Joel P. Gebbie Site Vice President J RW/kmh Enclosure Donald C. Cook Nuclear Plant Unit 1, Revised Small Break Loss-Of-Coolant Accident Analysis Report c: C. A. Casto, NRC Region III J. T. King, MPSC S. M. Krawec, AEP Ft. Wayne, w/o enclosure MDEQ - RMD/RPS NRC Resident Inspector P. S. Tam, NRC Washington, DC

Enclosure to AEP-NRC-2012-71 Donald C. Cook Nuclear Plant Unit 1 REVISED SMALL BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT

Westinghouse Non-Proprietary Class 3 D. C. COOK UNIT 1 SMALL BREAK LOCA ANALYSIS REPORT 1 INTRODUCTION The purpose of this report is to document the small break loss-of-coolant accident (SBLOCA) analysis performed for Donald C. Cook Nuclear Plant Unit I (D. C. Cook Unit 1) at a core power level of 3304 MWt (plus 0.34% uncertainty) with the emergency core cooling system (ECCS) cross-tie configurations as follows: 1) During the injection phase of the transient, the high head safety injection (HHSI) cross-tie valves are conservatively modeled to be closed and the residual heat removal (RHR) cross-tie valves are assumed to be open and 2) During the recirculation phase, the HHSI cross-tie valves are assumed to be open and the RHR cross-tie valves are assumed to be closed to isolate the trains. Additionally, it is conservatively modeled that RHR spray is active from the start of cold leg recirculation. The SBLOCA analysis modeling of the HHSI cross-tie valve closed during the ECCS injection phase supports operation with the HHSI cross-tie valve open or closed during the ECCS injection phase. The purpose of analyzing the SBLOCA is to demonstrate conformance with the 10 CFR 50.46 (Reference 1) requirements for the specified conditions. Important input assumptions, as well as analytical models and analysis methodology for the SBLOCA are contained in subsequent sections. Analysis results are provided in the form of tables and figures, as well as a more detailed description of the limiting transient. The analysis has shown that no design or regulatory limit related to the SBLOCA transient would be exceeded due to the specified cross-tie configuration and plant operation with the associated plant parameters.

2 INPUT PARAMETERS AND ASSUMPTIONS The important plant conditions and features for D. C. Cook Unit 1 that are supported by this analysis are listed in Table 1. Additional considerations for several parameters identified in Table 1 are discussed below.

Figure 1 depicts the hot rod axial power shape modeled in the Small Break LOCA analysis. This shape was chosen because it represents a distribution with power concentrated in the upper regions of the core (the axial offset is +20%). Such a distribution is limiting for Small Break LOCA since it minimizes coolant swell while maximizing vapor superheating and fuel rod heat generation at the uncovered elevations. The chosen power shape has been conservatively scaled to a standard 2-line segment K(Z) envelope based on the peaking factors shown in Table 1.

Figures 2 through 3a provide the ECCS flows modeled in the Small Break LOCA analysis. Tables 2 through 3a provide the flows used to generate Figures 2 through 3a, respectively. Figures 2 and 2a show the ECCS pumped injection flow versus pressure curves utilized during the injection phase. Figures 3 and 3a show the ECCS pumped injection flow versus pressure curves utilized during the cold leg recirculation phase (with HHSI available and RHR isolated, assuming RHR spray active). Note that the 1.5-, 2.0-, 2.5-,

2.75-, 3.0-, 3.25-, 3.5- 3.75-, and 4.0-inch break cases modeled more conservative ECCS pumped injection flows during cold leg recirculation phase than the available flow documented herein. Figures 2 and 3 show the flows from one charging (CHG) pump, one HHSI pump and one RHR pump, where the broken (or faulted) loop flow spills to reactor coolant system (RCS) pressure. Figures 2a and 3a show flows from one CHG pump, one HHSI pump and one RHR pump for the scenario where the break is postulated along the accumulator line. In this scenario, the faulted loop CHG flow spills to RCS pressure I

Westinghouse Non-Proprietary Class 3 and the faulted loop HHSI/RHR flow spills to 0 psig containment pressure. Note that hereafter, pumped injection subsystems of the ECCS (CHG, HHSI and RHR) are referred to collectively as safety injection (SI).

The analysis utilizes an adjusted nominal vessel average temperature (Tavg) of 577.4 0 F (with + 4VF uncertainty specified by NOTRUMP-EM) to support the D. C. Cook Unit 1 specific Tavg value of 575.4°F with +5.1°F uncertainty. The analysis supports operation for a nominal full-power Tavg range of 553.7°F to 575.4°F with +5.10 F/-4.1°F uncertainty. Additionally, the analysis utilizes a nominal pressurizer pressure of 2250 psia (plus +67 psi uncertainty) and supports operation at nominal pressurizer pressures of 2 100 psia and 2250 psia with +67 psi uncertainty.

3 DESCRIPTION OF ANALYSIS 3.1 ANALYTICAL MODEL The requirements for an acceptable Emergency Core Cooling System (ECCS) evaluation model are presented in Appendix K of 10 CFR 50. For LOCAs due to Small Breaks, less than 1 square foot in area, the Westinghouse NOTRUMP Small Break LOCA ECCS Evaluation Model (References 2, 3, and 4) is used. The Westinghouse NOTRUMP Small Break LOCA ECCS Evaluation Model (NOTRUMP-EM) was developed to determine the RCS response to design basis Small Break LOCAs, and to address NRC concerns expressed in NUREG-0611 (Reference 5).

The NOTRUMP-EM consists of the NOTRUMP and LOCTA-IV computer codes. The NOTRUMP code is employed to calculate the transient depressurization of the Reactor Coolant System (RCS), as well as to describe the mass and energy release of the fluid flow through the break. Among the features -of the NOTRUMP code are: calculation of thermal non-equilibrium in all fluid volumes, flow regime-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, regime-dependent drift flux calculations in multiple-stacked fluid nodes and regime-dependent heat transfer correlations. These features provide NOTRUMP with the capability to accurately calculate the mass and energy distribution throughout the RCS during the course of a SBLOCA.

The RCS model is nodalized into volumes interconnected by flow paths. The broken loop and each of the three intact loops are modeled explicitly, primarily to model the asymmetric safety injection flows that result from closure of one or both valves in the high head safety injection cross-tie. Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum.

The multi-node capability of the program enables explicit, detailed spatial representation of various system components which, among other capabilities, enables a calculation of the behavior of the loop seal during a SBLOCA. The reactor core is represented as heated control volumes with associated phase separation models to permit transient mixture height calculations.

Fuel cladding thermal analyses are performed with SBLOCTA, a SBLOCA version of the LOCTA-IV code (Reference 3), using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture heights as boundary conditions. The SBLOCTA code models the hot rod and the average hot assembly rod, assuming a conservative power distribution that is skewed to the top of the core. Figure 4 illustrates the code interface for the Small Break Model.

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Westinghouse Non-Proprietary Class 3 3.2 ANALYSIS The SBLOCA analysis for D. C. Cook Unit 1 considered a spectrum of eleven different break cases, including 1.5-, 2-, 2.5-, 2.75-, 3-, 3.25-, 3.5-, 3.75-, 4-, 6- and 8.75-inch breaks. The 3.25-inch break was found to be limiting for peak cladding temperature (PCT) and local oxidation. The 1.5- and 8.75-inch cases showed no core uncovery and therefore PCT information was not calculated.

The most limiting single active failure used for a Small Break LOCA is that of an emergency power train failure which results in the loss of one complete train of ECCS components. In addition, a Loss-of-Offsite Power (LOOP) is postulated to occur coincident with reactor trip. This means that with the assumed loss of emergency power there is a loss of one CHG pump, one HHSI pump and one RHR pump. The Small Break LOCA analysis performed for D. C. Cook Unit 1 models the ECCS injection phase flow as being delivered to both the intact and broken loops at the RCS backpressure for breaks smaller than the accumulator line inner diameter (1.5-inch through 6-inch breaks). For breaks equal to or greater than the accumulator line inner diameter (8.75-inch break), the broken loop flow spills to containment pressure. Note that for the 8.75-inch break, the broken loop CHG flow is assumed to inject to the cold leg at the RCS backpressure since it is not affected by the accumulator line break (CHG injects via a separate connection to the cold leg). The ECCS pumped injection flow rates for these scenarios are illustrated in Figures 2 and 2a for Injection Phase, and 3 and 3a for Cold Leg Recirculation Phase. The LOOP and the failure of an emergency diesel generator to start as the limiting single failure for SBLOCA is part of the NRC approved methodology. The single failure assumption is extremely limiting due to the fact that one train of safety injection (SI), one motor driven auxiliary feedwater (AFW) pump, and power to the reactor coolant pumps (RCPs) are all modeled to be lost. Any other active single failure would not result in a more limiting scenario since increased SI flow would improve the overall transient results.

Prior to break initiation, the plant is in a full power (100.34%) equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. Other initial plant conditions used in the analysis are given in Table 1. Subsequent to the break opening, a period of reactor coolant system blowdown ensues in which the heat from fission product decay, the hot reactor internals, and the reactor vessel continues to be transferred to the RCS fluid. The heat transfer between the RCS and the secondary system may be in either direction and is a function of the relative temperatures of the primary and secondary conditions. In the case of continuous heat addition to the secondary side during a period of quasi-equilibrium, an increase in the secondary system pressure (due to the assumed turbine trip following the reactor trip discussed below) results in steam relief via the steam generator safety valves.

When a Small Break LOCA occurs, depressurization of the RCS causes fluid to flow into the loops from the pressurizer resulting in a pressure and level decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low-pressure reactor trip setpoint, conservatively modeled as 1860 psia, is reached. LOOP is postulated to occur coincident with reactor trip. A safety injection signal is generated when the pressurizer low-pressure safety injection setpoint, conservatively modeled as 1715 psia, is reached: Safety injection flow is delayed 54 seconds after the occurrence of the low-pressure condition. This delay conservatively accounts for signal processing, diesel generator start up and emergency power bus loading consistent with the loss-of-offsite power coincident with reactor trip, as well as the pump acceleration and valve delays.

The following countermeasures limit the consequences of the accident in two ways:

3

Westinghouse Non-Proprietary Class 3 I. Reactor trip and borated water injection supplement void formation in causing a rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay. No credit is taken in the SBLOCA analysis for the boron content of the injection water. In addition, credit is taken in the SBLOCA analysis for the insertion of Rod Cluster Control Assemblies (RCCAs) subsequent to the reactor trip signal such that the core is rendered sub-critical. A rod drop time of 2.4 seconds was used while also considering an additional 2 seconds for the signal processing delay time. Therefore, a total delay time of 4.4 seconds from the time of reactor trip signal to full rod insertion was used in the SBLOCA analysis.

2. Injection of borated water provides sufficient flooding of the core to prevent excessive cladding temperatures.

During the earlier part of the Small Break transient (prior to the postulated loss-of-offsite power coincident with reactor trip), the loss of flow through the break is not sufficient to overcome the positive core flow maintained by the reactor coolant pumps. During this period, upward flow through the core is maintained. However, following the reactor coolant pump trip (due to a LOOP) and subsequent pump coastdown, a period of core uncovery occurs. Ultimately, the Small Break transient is terminated when the top of the core is recovered or the core mixing level is increasing, and ECCS flow provided to the RCS exceeds the break flow rate.

The core heat transfer mechanisms associated with the Small Break transient include the break itself, the injected ECCS water, and the heat transferred from the RCS to the steam generator secondary side. Main feedwater (MFW) is conservatively isolated in 8 seconds following the generation of the pressurizer low-pressure SI signal. Additional makeup water is also provided to the secondary using the auxiliary feedwater (AFW) system. An AFW actuation signal is derived from the pressurizer low-pressure reactor trip signal and results in the delivery of AFW flow 80 seconds after reactor trip. The heat transferred to the secondary side of the steam generator aids in the reduction of the RCS pressure.

Should the RCS depressurize to approximately 600 psia (accumulator minimum pressure), the cold leg accumulators begin to inject borated water into the reactor coolant loops as reflected in Table 6.

4 ACCEPTANCE CRITERIA The acceptance criteria for the LOCA are described in 10 CFR 50.46 (Reference 1) as follows:

1. The calculated maximum fuel element cladding temperature shall not exceed 2200'F.

2. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

4. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

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Westinghouse Non-Proprietary Class 3

5. After any calculated successfud initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Criteria 1 through 3 are explicitly covered by the Small Break LOCA analysis.

For criterion 4, the appropriate core geometry was modeled in the analysis. The results based on this geometry satisfy the peak clad temperature (PCT) criterion of 10 CFR 50.46 and consequently, demonstrate that the core remains amenable to cooling.

For criterion' 5, Long-Term Core Cooling (LTCC) considerations are not directly applicable to the Small Break LOCA transient analysis addressed herein, with the exception of predicting switchover from ECCS injection phase to ECCS recirculation phase and ensuring the SBLOCA transient remains terminated.

The acceptance criteria were established to provide a significant margin in ECCS performance following a LOCA.

5 RESULTS In order to determine the conditions that produced the most limiting SBLOCA case (as determined by the highest calculated peak cladding temperature), eleven break cases were examined for D. C. Cook Unit 1.

These cases were investigated to capture the most severe postulated Small Break LOCA event. The following discussion provides insight into the analyzed conditions.

The results of the generic study documented in Reference 6 demonstrate that the cold leg break location is limiting with respect to postulated cold leg, hot leg and pump suction leg break locations. The PCT results for D. C. Cook Unit 1 are shown in Tables 4 and 5. Inherent in the Small Break analysis are several input parameters (see Section 2 and Table 1), while Table 6 provides the key transient event times.

5.1 LIMITING BREAK CASE The SBLOCA analysis for D. C. Cook Unit 1 showed that the 3.25-inch break is the limiting case. A time-in-life study to determine the limiting PCT for this case considering clad burst concluded that the maximum PCT occurs at beginning-of-life (BOL). A summary of the transient response for the limiting PCT case is shown in Figures 5 through 15. These figures present the response of the following parameters:

• RCS Pressure

• Core Mixture Level

• Core Exit Vapor Temperature

• Broken and Intact Loops Secondary Pressures

• Break Vapor Flow Rate

• Break Liquid Flow Rate

• Broken and Intact Loops Accumulator Flow Rates

• Broken and Intact Loops Pumped Safety Injection Flow Rates

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Westinghouse Non-Proprietary Class 3

• Clad Temperature at PCT Elevation

• Hot Spot Fluid Temperature at PCT Elevation

• Rod Film Heat Transfer Coefficient at PCT Elevation Upon initiation of the limiting 3.25-inch break for D. C. Cook Unit 1, there is an initial rapid depressurization of the RCS followed by an intermediate equilibrium at approximately 1150 psia (see Figure 5). The limiting 3.25-inch break depressurizes to the accumulator injection setpoint of 600 psia at approximately 1264 seconds (see Figure 11). During the initial period of the Small Break transient, the effect of the break flow rate is not sufficient to overcome the flow rate maintained by the reactor coolant pumps as they coast down. As such, normal upward flow is maintained through the core and core heat is adequately removed. Following reactor trip, the removal of the heat generated as a result of the decay of fission products is accomplished via a two-phase mixture level covering the core. The core mixture level and peak clad temperature transient plots for the limiting break calculations are illustrated in Figures 6 and 13, respectively. These figures show that the peak clad temperature occurs near the time when the core is most deeply uncovered and the top of the core is being cooled by steam. This time is characterized by the highest vapor superheating above the mixture level (refer to Figure 7). For D. C.

Cook Unit 1, the limiting PCT time-in-life was determined to be BOL.

A comparison of the flow provided by the safety injection system to the intact and broken loops can be found in Figure 12. The cold leg break vapor and liquid mass flow rates are provided in Figures 9 and 10, respectively. Figures 14 and 15 provide additional information on the fluid temperature at the hot spot and hot rod surface heat transfer coefficient at the hot spot, respectively. Figure 8 depicts the secondary side pressure for both the intact and broken loops for the limiting PCT break case.

Maximum Local Oxidation For the D. C. Cook Unit 1 SBLOCA analysis, the maximum local oxidation case was the 3.25-inch break case. Based on the time-in-life study, the maximum local transient oxidation is 3.61% at 11,500 MWD/MTU. The limiting transient oxidation occurs at the hot rod burst elevation and includes both outside oxidation and post-rupture inside oxidation in the burst region. Pre-existing (pre-transient) oxidation was also considered and the sum of the pre-transient and transient oxidation remains below 17% at all times in life, for all fuel resident in the core.

Core Wide Average Oxidation Tables 4 and 5 indicate that for the D. C. Cook Unit 1 SBLOCA analysis, the core wide average oxidation for all cases is less than 1%. Therefore the calculated total amount of hydrogen generation is less than the 1% limit defined by 10 CFR 50.46 (Reference 1).

5.2 NON-LIMITING BREAK CASES Summaries of the transient responses for the non-limiting break cases (1.5-, 2-, 2.5-, 2.75-, 3-, 3.5-, 3.75-,

4-, 6- and 8.75-inch) analyzed for D. C. Cook Unit I are shown in Figures 16 through 43. The beginning-of-life (BOL) results for these break spectrum cases are given in Table 4. The 1.5-and 8.75-inch cases did not show core uncovery, therefore PCT information was not calculated. The plots for each of the additional non-limiting break cases include:

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Westinghouse Non-Proprietary Class 3 1L RCS Pressure

2. Core Mixture Level

3. Clad Temperature at PCT Elevation (Note no PCT plots provided for 1.5- and 8.75-inch cases)

The PCTs for each of the additional breaks considered are shown in Table 4 and are less than the limiting 3.25-inch break case.

5.3 ADDITIONAL ANALYSIS DETAILS 5.3.1 Switchover from ECCS Injection Phase to ECCS Recirculation Phase When the refueling water storage tank (RWST) volume of 280,000 gallons is delivered via safety injection and containment spray, NOTRUMP predicts switchover from ECCS injection phase to ECCS recirculation phase. At that time RHR flow is re-aligned to the sump and an interruption in RHR flow for up to 5 minutes may occur. For break cases that have a calculated RCS pressure at or below the RHR cut-in pressure, the 5 minute interruption in RHR flow is considered. The applicable transients were shown to satisfy the analysis termination conditions, as discussed in more detail below.

5.3.2 Transient Termination The 10 CFR 50.46 criteria (Reference 1) continue to be satisfied beyond the end of the calculated transient due to the presence of the following conditions:

1. The RCS pressure is gradually decreasing or reached equilibrium.

2. The net mass inventory is increasing or reached equilibrium.

3. The core mixture level is recovered, or recovering due to increasing mass inventory.

4. As the RCS inventory continues to gradually increase, the core mixture level will continue to increase and the fuel cladding temperatures will continue to decline indicating that the temperature excursion is terminated.

6 CONCLUSIONS The Small Break LOCA analysis for D. C. Cook Unit 1 considered a break spectrum of 1.5-, 2-, 2.5-,

2.75-, 3-, 3.25-, 3.5-, 3.75-, 4-, 6- and 8.75-inch diameters. The limiting peak cladding temperature of 1725°F was calculated at BOL for the 3.25-inch case and a maximum local transient oxidation of 3.61%

was calculated at the limiting time-in-life of 11,500 MWD/MTU for the 3.25-inch case. The analysis is applicable to a core power up to and including 3304 MWt (plus 0.34% uncertainty) and supports operation with both the HHSI and RHR cross-tie valve(s) open during the injection phase and with HHSI cross-tie valves open and RHR cross-tie valves closed during the recirculation phase.

The analysis presented herein shows that the accumulator and SI subsystems of the ECCS, together with the heat removal capability of the steam generators, provide sufficient core heat removal capability to maintain the calculated PCT for Small Break LOCA below the required limit of 10 CFR 50.46 (Reference 1). Furthermore, the analysis shows that the local cladding oxidation and core wide average oxidation, including consideration of pre-existing and post-LOCA oxidation, and cladding outside and post-rupture inside oxidation, are less than the 10 CFR 50.46 (Reference 1) limits at all times in life for all fuel 7

Westinghouse Non-Proprietary Class 3 resident in the core. Note that the core wide average oxidation results illustrate that the total hydrogen generation is less than 1%.

Table 7 provides a results summary for the D. C. Cook Unit I SBLOCA analysis including PCT, maximum local transient oxidation and total hydrogen generation.

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Westinghouse Non-Proprietary Class 3 7 REFERENCES

1. "Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors," 10 CFR 50.46, August 2007 and "ECCS Evaluation Models," Appendix K of 10 CFR 50, June 2000.

2. Meyer, P. E., "NOTRUMP - A Nodal Transient Small Break and General Network Code,"

WCAP- 10079-P-A, (proprietary) and WCAP- 10080-NP-A (non-proprietary), August 1985.

3. Lee, N. et al., "Westinghouse Small Break ECCS Evaluation Model. Using the NOTRUMP Code," WCAP- 10054-P-A (proprietary) and WCAP- 10081-NP-A (non-proprietary),

August 1985.

4. Thompson, C. M. et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," WCAP-10054-P-A, Addendum 2, Rev. 1 (proprietary), July 1997.

5. "Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse - Designed Operating Plant," NUREG-06 11, January 1980.

6. Rupprecht, S. D. et al., "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," WCAP- 11145-P-A (proprietary), October 1986.

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Westinghouse Non-Proprietary Class 3 Table 1 Input Parameters Used in the Small Break LOCA Analysis Input Parameter Value Core Rated Thermal Power-100% 3304 MWt Calorimetric Uncertainty, % 0.34 Fuel Type 15 x 15 Upgrade Fuel Total Core Peaking Factor, FQ 2.32 HotChannel Enthalpy Rise Factor, FA 1.55 Hot Assembly Average Power Factor, PHA 1.38 Maximum Axial Offset, % 20 Initial RCS Loop Flow, gpm/loop 83,200 0

Initial Vessel Tavg, F 577.4(')

Initial Pressurizer Pressure (plus uncertainties), psia 2317(2)

Reactor Coolant Pump Type 93A Pressurizer Low-Pressure Reactor Trip Setpoint, psia 1860 Reactor Trip Signal Delay Time, seconds 2.0 Rod Drop Time, seconds 2.4 Auxiliary Feedwater Temperature (Maximum), 'F 120 AFW Flow (Minimum) to all 4 Steam Generators, gpm 427.25 AFW Flow Delay Time (Maximum), seconds 80 AFW Actuation Signal Reactor Trip/Low Pressurizer Pressure Maximum AFW Piping Purge Volume, ft3 78 Steam Generator Tube Plugging (Maximum), % 10 Maximum MFW Isolation, seconds 8 MFW Isolation Signal Safety Injection Actuation Steam Generator Secondary Water Mass, lbm/SG 101,169 Containment Spray Flowrate for 2 Pumps, gpm 7,400 RWST Deliverable Volume (Minimum), gallons 280,000 Notes:

(1) Analysis supports operation over the range of nominal full-power Tavg values of 553.7°F - 575.4°F.

(2) Analysis supports operation at nominal initial pressurizer pressure (without uncertainties) of 2100 psia and 2250 psia.

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Westinghouse Non-Proprietary Class 3 Table 1 (continued)

Input Parameters Used in the Small Break LOCA Analysis

,Input Parameter Value SI Temp at Cold Leg Recirculation Time (Maximum), IF 190 ECCS Configuration 1 CHG pump, 1 HHSI pump, I RHR pump -

faulted loop injects to RCS pressure (1.5-inch through 6-inch breaks) 1 CHG pump, 1 HI-SI pump, 1 RHR pump -no RHRIHHSI in the faulted loop because the break is postulated along the accumulator line, faulted loop CHG flow injects to RCS pressure (8.75-inch)

ECCS Water Temperature (Maximum), IF 105 Pressurizer Low-Pressure Safety Injection Setpoint, psia 1715 SI Flow Delay Time, seconds 54 ECCS Flow vs. Pressure (Jnjection and Cold Leg Recirculation See Tables 2, 2a, 3 and 3a Phases)

Initial Accumulator Water/Gas Temperature, IF 130 Initial Nominal Accumulator Water Volume, ft 946 Minimum Accumulator Pressure, psia 600 11

Westinghouse Non-Proprietary Class 3 Table 2 Safety Injection Flows Used in the Small Break LOCA Analysis - Injection Phase (1 CHG pump, 1 HHSI pump, 1 RHR pump - faulted loop injects to RCS pressure - 1.5-inch through 6-inch breaks)

RCS Pressure Broken Loop Intact Loops (lbm/sec)

(psia) (lbm/sec)

Loop 1 Loop 2 Loop 3 Loop 4 14.7 188.63 158.38 175.00 159.93 114.7 127.20 102.43 117.32 103.38 214.7(1) 38.89 12.50 34.37 12.50 314.7 37.49 12.21 33.10 12.21 414.7 35.93 11.90 31.73 11.90 514.7 34.30 11.57 30.26 11.57 614.7 32.56 11.23 28.64 11.23 714.7 30.73 10.88 26.98 10.88 814.7 28.80 10.54 25.22 10.54 914.7 26.74 10.18 23.35 10.18 1014.7 24.34 9.82 21.14 9.82 1114.7 21.49 9.46 18.54 9.46 1214.7 18.00 9.10 15.32 9.10 1314.7 13.03 8.70 10.73 8.70 1414.7(2) 10.36 8.29 8.29 8.29 1514.7 9.82 7.86 7.86 7.86 1614.7 9.27 7.42 7.42 7.42 1714.7 8.72 6.98 6.98 6.98 1814.7 8.08 6.47 6.47 6.47 1914.7 7.41 5.93 5.93 5.93 2014.7 6.72 5.38 5.38 5.38 2114.7 5.94 4.76 4.76 4.76 2214.7 5.02 4.01 4.01 4.01 2314.7 0 0 0 0 Notes:

(1) RHR cut-in pressure (2) HHSI cut-in pressure 12

Westinghouse Non-Proprietary Class 3 Table 2a Safety Injection Flows Used in the Small Break LOCA Analysis - Injection Phase (1 CHG pump, I HHSI pump, I RHR pump -no RHR/HHSI in the faulted loop because the break is postulated along the accumulator line, faulted loop CHG flow injects to RCS pressure - 8.75-inch break)

RCS Pressure Broken Loop (lbm/sec) Intact Loops (Ibm/sec)

(psia)

Loop I - CHG Loop 1 - Loop 2 Loop 3 Loop 4 RHR/HHSI 14.7 16.34 157.03 139.36 156.01 140.26 34.7 16.27 215.73 126.84 99.71 127.64 54.7 16.20 274.56 115.46 32.18 116.17 74.7() 16.13 303.81 97.04 12.90 97.63 94.7 16.06 323.64 72.95 12.84 73.37 114.7 15.98 345.72 43.70 12.78 43.93 134.7(2) 15.91 365.86 12.73 12.73 12.73 154.7 15.84 365.86 12.67 12.67 12.67 214.7 15.63 365.86 12.50 12.50 12.50 314.7 15.25 365.86 12.21 12.21 12.21 414.7 14.83 365.86 11.90 11.90 11.90 514.7 14.45 365.86 11.57 11.57 11.57 614.7 14.03 365.86 11.23 11.23 11.23 714.7 13.60 365.86 10.88 10.88 10.88 814.7 13.17 365.86 10.54 10.54 10.54 914.7 12.73 365.86 10.18 10.18 10.18 1014.7 12.29 365.86 9.82 9.82 9.82 1114.7 11.83 365.86 9.46 9.46 9.46 1214.7 11.38 365.86 9.10 9.10 9.10 1314.7 10.88 365.86 8.70 8.70 8.70 1414.7 10.36 365.86 8.29 8.29 8.29 1514.7 9.82 365.86 7.86 7.86 7.86 1614.7 9.27 365.86 7.42 7.42 7.42 1714.7 8.72 365.86 6.98 6.98 6.98 1814.7 8.08 365.86 6.47 6.47 6.47 1914.7 7.41 365.86 5.93 5.93 5.93 2014.7 6.72 365.86 5.38 5.38 5.38 13

Westinghouse Non-Proprietary Class 3 Table 2a Safety Injection Flows Used in the Small Break LOCA Analysis - Injection Phase (1 CHG pump, 1 HHSI pump, I RHR pump -no RHR/HHSI in the faulted loop because the break is postulated along the accumulator line, faulted loop CHG flow injects to RCS pressure - 8.75-inch break)

RCS Pressure Broken Loop (Ibm/sec) Intact Loops (lbm/sec)

(psia)

Loop 1 - CHG Loop 1 - Loop 2 Loop 3 Loop 4 RHR/HHSI 2114.7 5.94 365.86 4.76 4.76 4.76 2214.7 5.02 365.86 4.01 4.01 4.01 2314.7 0 365.86 0 0 0 Notes:

(1) HHSI cut-in pressure (2) RHR cut-in pressure 14

Westinghouse Non-Proprietary Class 3 Table 3 Safety Injection Flows Used in the Small Break LOCA Analysis - Recirculation Phase (1 CHG pump, 1 HHSI pump, 1 RHR pump - faulted loop injects to RCS pressure - RHR Spray active -

1.5-inch through 6-inch breaks)

RCS Pressure Broken Loop (Ibm/sec) Intact Loops (Ibm/sec)

(psia) Loop 1 Loop 2 Loop 3 Loop 4 14.7 36.1 32.1 31.3 32.3 34.7 35.9 31.9 31.1 32.1 54.7 35.7 31.7 30.9 31.9 74.7 35.5 31.5 30.7 31.7 94.7 35.3 31.3 30.5 31.5 114.7 35.0 31.1 30.3 31.3 134.7 34.8 30.9 30.1 31.1 154.7 34.6 30.7 29.9 30.9 174.7 34.3 30.5 29.7 30.6 194.7 34.1 30.3 29.5 30.4 214.7 33.9 30.1 29.3 30.2 234.7 33.6 29.9 29.1 30.0 254.7 33.4 29.6 28.9 29.8 274.7 33.2 29.4 28.7 29.6 294.7 32.9 29.2 28.4 29.3 314.7 32.7 29.0 28.2 29.1 414.7 31.5 27.9 27.2 28.0 514.7 30.2 26.7 26.0 26.8 614:7 28.8 25.5 24.9 25.6 714.7 27.4 24.2 23.6 24.3 814.7 26.0 22.9 22.3 23.0 914.7 24.5 21.5 21.0 21.6 1014.7 22.6 19.9 19.4 20.0 1114.7 20.7 18.1 17.7 18.1 1214.7 18.4 15.9 15.6 16.0 1314.7 15.3 13.0 12.8 13.1 1414.7 11.4 9.4 9.3 9.4 1514.7 9.2 7.4 7.4 7.4 15

Westinghouse Non-Proprietary Class 3 3

Table Safety Injection Flows Used in the Small Break LOCA Analysis - Recirculation Phase (1 CHG pump, 1 HHSI pump, 1 RHR pump - faulted loop injects to RCS pressure - RHR Spray active -

1.5-inch through 6-inch breaks)

RCS Pressure Broken Loop (lbm/sec) Intact Loops (lbm/sec)

(psia) Loop 1 Loop 2 Loop 3 Loop 4 1614.7 8.7 7.0 7.0 7.0 1714.7 8.1 6.6 6.6 6.6 16

Westinghouse Non-Proprietary Class 3 Table 3a Safety Injection Flows Used in the Small Break LOCA Analysis - Recirculation Phase (1 CHG pump, I HltSI pump, 1 RHR pump -faulted loop CHG flow injects to RCS pressure and faulted loop HHISI/RHR flow shills to containment (0I nsia* - RHR Snrav active - R17.-ineh break*

RCS Pressure Broken Loop (ibm/sec) Intact Loops (Ibm/sec)

(psia)

Loop I - CIG Loop I - Loop 2 Loop 3 Loop 4 RHR/HHSI 14.7 15.4 20.7 32.1 31.3 32.3 34.7 15.4 39.8 31.9 12.4 32.0 54.7 15.3 39.9 31.6 12.4 31.8 74.7 15.3 40.0 31.4 12.3 31.5 94.7 15.2 40.1 31.1 12.2 31.2 114.7 15.1 40.3 30.8 12.2 31.0 134.7 15.0 40.4 30.6 12.1 30.7 154.7 15.0 40.5 30.3 12.1 30.4 174.7 14.9 40.6 30.0 12.0 30.1 194.7 14.8 40.8 29.7 12.0 29.9 214.7 14.8 40.9 29.4 11.9 29.6 234.7 14.7 41.1 29.2 11.8 29.3 254.7 14.6 41.2 28.9 11.8 29.0 274.7 14.5 41.3 28.6 11.7 28.7 294.7 14.5 41.4 28.3 11.7 28.4 314.7 14.4 41.6 28.0 11.6 28.1 414.7 14.1 42.2 26.5 11.3 26.6 514.7 13.7 42.9 24.9 11.0 25.0 614.7 13.3 55.9 22.6 10.7 22.6 714.7 12.8 57.8 19.8 10.4 19.8 814.7 12.4 59.8 16.5 10.0 16.6 914.7 12.0 62.2 12.7 9.7 12.7 1014.7 11.6 64.3 9.3 9.3 9.3 1114.7 11.1 64.3 9.0 9.0 9.0 1214.7 10.7 64.3 8.6 8.6 8.6 1314.7 10.2 64.3 8.3 8.3 8.3 1414.7 9.8 64.3 7.9 7.9 7.9 1514.7 9.2 64.4 7.4 7.4 7.4 17

Westinghouse Non-Proprietary Class 3 Table 3a Safety Injection Flows Used in the Small Break LOCA Analysis - Recirculation Phase (1 CHG pump, I HHSI pump, 1 RHR pump -faulted loop CHG flow injects to RCS pressure and faulted loop HHSI/RHR flow spills to containment (0 psia) - RHR Spray active - 8.75-inch break)

RCS Pressure Broken Loop (Ibm/sec) Intact Loops (Ibm/sec)

(psia)

Loop 1 - CHG Loop 1 - Loop 2 Loop 3 Loop 4 RHR/HHSI 1614.7 8.7 64.4 7.0 7.0 7.0 1714.7 8.1 64.4 6.6 6.6 6.6 18

Westinghouse Non-Proprietary Class 3 Table 4 SBLOCTA BOL Results Break Size (in) 2 2.5 2.75 3 3.25 3.5 3.75 4 6 PCT('F) 968.4 1433.4 1452.5 1584.0 1725.0 1705.3 1517.9 1411.2 670.6 PCT Time (s) 2284.4 2684.0 2140.6 2000.5 1483.4 1249.3 1129.8 986.2 404.1 PCT Elevation (ft) 11.0 11.50 11.50 11.75 11.75 11.75 11.50 11.25 11.0 Max. Local ZrO 2 (%) 0.03 0.70 0.54 1.26 2.08 1.72 0.5.6 0.26 0.0 Max. Local ZrO 2 Elev. (f1) 11.0 11.50 11.50 11.75 11.75 11.50 11.50 11.25 11.0 Core-Wide Avg. ZrO 2 (%) 0.0 0.09 0.07 0.17 0.30 0.26 0.08 0.04 0.0 Table 5 SBLOCTA Limiting Results from the 3.25-inch Time-in-Life Study Time-in-Life (MWD/MTU) BOL 11,500 PCT ('F) 1725.0 1720.5 PCT Time (s) 1483.4 1480.3 PCT Elevation (fi) 11.75 11.75 Hot Rod Burst Time (s) N/A 1478.4 Hot Rod Burst Elevation (f1) N/A 11.75 Max. Local Transient ZrO 2 (%) 2.08 3.61 Max. Local Transient ZrO 2 Elev. (ft) 11.75 11.75 Core-Wide Avg. ZrO 2 (%) 0.30 0.22 19

Westinghouse Non-Proprietary Class 3 Table 6 Time Sequence of Events 1.5- 2.5- 2.75- 3.25- 3.5- 3.75- 8.75-Event Time inch 2-inch inch inch 3-inch inch inch inch 4-inch 6-inch inch Break Initiation (s) 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Reactor Trip Signal (s) 89.8 45.9 27.9 ' 22.7 19.0 16.2 14.0 12.3 10.9 6.0 4.5 S-Signal (s). 112.8 60.9 38.2 31.6 26.8 23.2 20.3 17.9 15.7 8.4 6.7 SI Flow Delivered (1), (s) 166.8 114.9 92.2 85.6 80.8 77.2 74.3 71.9 69.7 62.4 60.7 Loop Seal Clearing (2)(s) 2492 1341 857 628 516 445 386 390 302 146 30 Core Uncovery (3) (s) N/A 1897 1027 1017 963 780 664 630 602 342 N/A Accumulator Injection (s) N/A N/A 3065 2129 1707 1264 1031 940 823 345 168 RWST Volume Delivered 14) (s) 2166 2158 2145 2138 2130 2121 2114 2110 2106 2043 1591 PCT Time (BOL) (s) N/A 2284 2684 2141 2001 1483 1249 1130 986 670.6 N/A Core Recovery (3) (s) N/A 6663 4032 4081 3977 3840 3973 4110 3404 423 N/A Notes:

(1) SI is assumed to begin 54.0 seconds (SI delay time) after the S-Signal.

(2) Loop seal clearing is assumed to occur when the steam flow through the broken loop, loop seal is sustained above I Ibm/s and the mixture level in the pump suction leg is at or below the top of the loop seal.

(3) The latest point of sustained core uncovery/recovery is reported.

(4) The analysis assumes minimum usable RWST volume (280,000 gal) delivered via ECCS injection and containment spray before the low level RWST water level signal for switchover to cold leg recirculation is reached.

20

Westinghouse Non-Proprietary Class 3 Table 7 SBLOCA Results Summary Peak Cladding Temperature (IF) 1725 Maximum Local Transient Oxidation (%) 3.61 Total Hydrogen Generation (%) < 1%

21

Westinghouse Non-Proprietary Class 3 L..

a) 0 0~

L..

0 0)

-B C-)

0

-J 0 2 4 6 8 10 12 Elevation (ft)

Figure 1 Small Break Hot Rod Power Shape 22

Westinghouse Non-Proprietary Class 3 Li Broken Loop)

L2 Intact Loop L3 Intact Loop L4 Intact Loop) 200-150- "

C/)

E c:100-I

• 03 0U) 0 500 1000 Pressure (psi1(5]) 2000 2500 Figure 2 Small Break LOCA Safety Injection Flows - Injection Phase I CHG pump, 1 HHSI pump, 1 RHR pump - faulted loop injects to RCS pressure - 1.5-inch through 6-inch breaks) 23

Westinghouse Non-Proprietary Class 3 Loop Broken Loop - CHG Flow)

Loop Broken Loop - HHSI/RHR Flow)

Loop 2 Intact Loop 3 Intact Loop Loop 4 Intact Loop)

CO E

..a,

-o Cn Ma 0

0 500 1000 15m0 2000 2500 Pressure (psiQ)~

Figure 2a Small Break LOCA Safety Injection Flows - Injection Phase (1 CHG pump, 1 HHSI pump, 1 RHR pump -no RHR/HHSI in the faulted loop because the break is postulated along the accumulator line, faulted loop CHG flow injects to RCS pressure -

8.75-inch break) 24

Westinghouse Non-Proprietary Class 3 Li Broken Loop)

. . L2 Intact Loop)

Loop)

-L3 Intact Lo o p L4 Intact o-o cr)

C/3 0 500 1000 1500 2000 Pressure (psia)

Figure 3 Small Break LOCA Safety Injection Flows - Cold Leg Recirculation Phase 1 CHG pump, I HHSI pump, 1 RHR pump - faulted loop injects to RCS pressure- 1.5-inch through 6-inch breaks) 25

Westinghouse Non-Proprietary Class 3 Loop Broken Loop CHG Flow)

Loop Broken Loop HHSI/RHR Flow)

Loop 2 Intact Loop)

Loop 3 Intact Loop Loop 4 Intact Loop)

CO, E

a, M

0 500 1000 1500 2000 Pressure (psia)

Figure 3a Small Break LOCA Safety Injection Flows - Cold Leg Recirculation Phase (1 CHG pump, 1 HHSI pump, 1 RHR pump -no RHR/HHSI in the faulted loop because the break is postulated along the accumulator line, faulted loop CHG flow injects to RCS pressure -

8.75-inch break) 26

Westinghouse Non-Proprietary Class 3 CORE PRESSURE, CORE FLOW, MIXTURE LEVEL, AND FUEL ROD POWER N HISTORY O<TIME<CORE COVERED L T 0 R C U T M A P

Figure 4 Code Interface Description for Small Break Model 27

Westinghouse Non-Proprietary Class 3

.2 1500 . .. . .. . .. . .. . .. .. . .. . .. .. . .. . . .. .. . .. . .

CL co I I I I I I I I I I I I I I I I I I I 0 1000 2000lime (s) 3000 4000 5000 Figure 5 3.25-inch Break RCS Pressure 28

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft a,

-J 0)

L..

0 1000 2000 3000 4000 5000 Time (s)

Figure 6 3.25-inch Break Core Mixture Level 29

Westinghouse Non-Proprietary Class 3

%- 1000-M_

E I8 I . I. I. .. . .. . .. .

600 400-0 1000 2000 3000 4000 500M Time (s)

Figure 7 3.25-inch Break Core Exit Vapor Temperature 30

Westinghouse Non-Proprietary Class 3 L1 Broken Loop)

L2 Intact Loop L3 Intact Loop L4 Intact Loop) 1150-1100 ................................................

1050 ..............................................

CL 1000 ................................................

c')950 02 900- ................................................

850-I I I I l l l l I I I I I I I I I I I I I owu 0 1000 2000 0 5000 3000 Time (s)

Figure 8 3.25-inch Break Broken and Intact Loops Secondary Pressures 31

Westinghouse Non-Proprietary Class 3 cn E

0 cf) 0n 0 1000 2000 lime (s) 3000 40(0 5000 Figure 9 3.25-inch Break Break Vapor Flow Rate 32

Westinghouse Non-Proprietary Class 3 E

0 15--

00 . . . . . . . . . . . . . .. ..... . . . . . . . . . . . . .

0 1000 2000 MW M 5M Time (s)

Figure 10 3.25-inch Break Break Liquid Flow Rate 33

Westinghouse Non-Proprietary Class 3 L1 Broken Intact Loop L2 Intoct Loop L3 Intact Loop L4 oU 601 '

L 0n 201*

k. .... AAA' l l n-a U LI6aI *-*---t - -.------. -. -

0 -~ -.- -.

0 1000 2000 3000 4000 5000 Time (S)

Figure 11 3.25-inch Break Broken and Intact Loops Accumulator Flow Rates 34

Westinghouse Non-Proprietary Class 3 L1 Broken Loop L2 Intact Loop L3 Intact Loop L4 Intact Loop)

E

-O 0)

M 0 1000 2000 3000 4000 5000 Time (s)

Figure 12 3.25-inch Break Broken and Intact Loops Pumped Safety Injection Flow Rates 35

Westinghouse Non-Proprietary Class 3

1000 . . . . . . . . . . . ... .. . .. . . . .. . . . . . . . . . . . . . . . . . . ..

00 60 .

001000 2000 M M 5 Time (s)

Figure 13 3.25-inch Break Clad Temperature at PCT Elevation (11.75 ft) 36

Westinghouse Non-Proprietary Class 3 1200-

1. .. . . . . . . . .

• 1000..

800-600. . ... .... ... ... . .

4'0 I I I I I I I I I I I I I I I 4I00I 0 1000 2000 3000 400( 5000 Time (s)

Figure 14 3.25-inch Break Hot Spot Fluid Temperature at PCT Elevation (11.75 ft) 37

Westinghouse Non-Proprietary Class 3 5

10 0 1000 2000 3000 4000 5000 Time (s)

Figure 15 3.25-inch Break Rod Film Heat Transfer Coefficient at PCT Elevation (11.75 ft) 38

Westinghouse Non-Proprietary Class 3

~1800" CL 0)

L...

2 1600-1400" 1000 '--**

0 1000 2000 3000 4000 5000 6000 Time (s)

Figure 16 1.5-inch Break RCS Pressure 39

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft

-X-

=E 0 1000 2000 Time (s) 0 5000 6000 Figure 17 1.5-inch Break Core Mixture Level 40

Westinghouse Non-Proprietary Class 3 I0O

.2

? 1600 . . . . . . . . . . . . . . . . . .

,-, low" 1400_

1200-1000-0 1000 2000 3000M 490 5000 6000 7000 lime (S)

Figure 18 2-inch Break RCS Pressure 41

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft 0 1000 2000 3 W00 4O0) 5000 6000 Time (s) 7000 Figure 19 2-inch Break Core Mixture Level 42

Westinghouse Non-Proprietary Class 3 I RJu 900 1........

(

U-001 . ........

1.~

aL..

'I, E ..............

0) 7001.......

6001. .......

fiII I I I I I I I I o i l I I k I I I I I a I I

, I~n I I I I I I - . . . .

1000 2000 3000 M00 5000 M00 7000 Time (s)

Figure 20 2-inch Break Clad Temperature at PCT Elevation (11.0 ft) 43

Westinghouse Non-Proprietary Class 3 CI)

(3) 1500 - . ........ ........ . . . . . . . . . . . . . . . . . . . .

C/3 (n)

L..

1000-500 00 lime (s)

Figure 21 2.5-inch Break RCS Pressure 44

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft

-4j X) 0 1000 2000 30M0 (s) 40 5w 600 Time Figure 22 2.5-inch Break Core Mixture Level 45

Westinghouse Non-Proprietary Class 3 01000-800 600 -. . . . . . . . . . .. . . . . . . . . .

p^ I I I I I I I I I I I I I I I I I I I 400 1000 2000 3000lime (s)4000 5000 6000 Figure 23 2.5-inch Break Clad Temperature at PCT Elevation (11.5 ft) 46

Westinghouse Non-Proprietary Class 3 CO 1500 ...........................................

5)0 I.--

1000..............................................

0 1000 2000 3000 4000 5000 6000 Time (s)

Figure 24 2.75-inch Break RCS Pressure 47

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft

-Jý X.

0 1000 2000 3000 4000 5000 6000 Time (s)

Figure 25 2.75-inch Break Core Mixture Level 48

Westinghouse Non-Proprietary Class 3 M, IU-V 400 1000 2000 Mw Time (s) 4)M Figure 26 2.75-inch Break Clad Temperature at PCT Elevation (11.5 ft) 49

Westinghouse Non-Proprietary Class 3

.2 1500" 0n C.

00 -

5a ..... . .. . . .

I I I I I 0 1000 2000 Time (s)3000 4000 5000 Figure 27 3-inch Break RCS Pressure 50

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft

=3

-4J X,

=E 0 1000 2000 3000 4000 5000 Time (s)

Figure 28 3-inch Break Core Mixture Level 51

Westinghouse Non-Proprietary Class 3 0.)

E!1000 ..............................................

L'.

EQ_

800-....... ..............

600........................................

400 0 2000 4000 ,5000 1000 Time (s) 3000 Figure 29 3-inch Break Clad Temperature at PCT Elevation (11.75 ft) 52

Westinghouse Non-Proprietary Class 3

.* 1500 . .. ........................

00 500- 0: 1000200 I I I I I  ! I I I I I 0 1000 2000 -)3000 4000 5000 Time (

Figure 30 3.5-inch Break RCS Pressure 53

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft 4-a,

-J a,

L..

0 1000 2000 3000 4000 5100 Time (s)

Figure 31 3.5-inch Break Core Mixture Level 54

Westinghouse Non-Proprietary Class 3 1200.

E 1000 - . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . ..

400 0 1000 2000 3000 4000 5000 Time (s)

Figure 32 3.5-inch Break Clad Temperature at PCT Elevation (11.75 ft) 55

Westinghouse Non-Proprietary Class 3 2 1500 cn o=3 cn o'3 (D

o..1000 .. . . . . . . . . . . . . . . . . . . .

1000 . ... . . . . . .. . . . . . . .. . . . . . . . . .. . . . . . . . . . .. . . . . . . ..

500-

^ I I I I I I I I I I I I I I I I I I I I UI II 0 1000 2000Time (s)3000 4000 5000 Figure 33 3.75-inch Break RCS Pressure 56

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft 1-.

L..

0 1000 2000 3000 4000 5M0w Time (s)

Figure 34 3.75-inch Break Core Mixture Level 57

Westinghouse Non-Proprietary Class 3 800.

600 - .. . .. .. . .. . .. . .

I ! I I , I I I I , I I I I , I I I I I I 400-0 1000 2000 3000 4000 5000 Time (s)

Figure 35 3.75-inch Break Clad Temperature at PCT Elevation (11.5 ft) 58

Westinghouse Non-Proprietary Class 3

.2 1500.....................................

L.o CO Cl)

()1000 .................................................

5000. .................... ............... ..............

0 i a t I , I I II I I I I I I I I I 0 1000 200030040 Time (s)

Figure 36 4-inch Break RCS Pressure 59

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft X)

MJ 0 1000 2000 3000 400M Time (s)

Figure 37 4-inch Break Core Mixture Level 60

Westinghouse Non-Proprietary Class 3 t looo-

. . . . ). .. . . . . . . . . . . . . . . . . .

10 00. .

FT 80 -. .. . . .. . . .

60 -. . .. . . . . . .

400-Time(s Figure 38 4-inch Break Clad Temperature at PCT Elevation (11.25 ft) 61

Westinghouse Non-Proprietary Class 3

.2 1500 ......

co C/)

co' Cf) 1000-I i i i I i p p p p 3 I I I I I I I I p I I  !

0 1000 2000 3000 4000 5000 Time (s)

Figure 39 6-inch Break RCS Pressure 62

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft

=3 XJ 0 1000 2000 lime (s)3000 4000 5000 Figure 40 6-inch Break Core Milxture Level 63

Westinghouse Non-Proprietary Class 3 55 550" 450 . .... . . . . . . .

4 0 ... ...

350 0 1000 2000Tie()300 lime (s) 4000 5000 Figure 41 6-inch Break Clad Temperature at PCT Elevation (11.0 ft) 64

Westinghouse Non-Proprietary Class 3 21500 co Cl) c ..1000-. . . . . .. . . . .. . . . .. . . .

500-0 1000 2000Time 4000 5000 (s) 3000 Figure 42 8.75-inch Break RCS Pressure 65

Westinghouse Non-Proprietary Class 3 Core Mixture Level Top of Core = 22.0762 ft 36 34-32 - . .... . . . . . . . . . . . . . . . ...

328 0)

L..

0 1000 2000 Time (s) 3000 4000 5000 Figure 43 8.75-inch Break Core Mixture Level 66