ML18100A291
| ML18100A291 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/31/1993 |
| From: | Akers J, Schrader K, Sklarsky D WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML18100A290 | List: |
| References | |
| WCAP-13657, NUDOCS 9304090305 | |
| Download: ML18100A291 (67) | |
Text
9304090305 930402 .
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WCAP-13657 WESTINGHOUSE PROPRIETARY CLASS 3 Salem Units 1&2 Small Break LOCA NOTRUMP Analysis Engineering Report in Support of the Fuel Upgrade I Margin Recovery Program.
March 1993
- K. J. Schrader D. W. Sklarsky J. J. Akers APPROVE~' -~
Nuclear Safety Analysis and Strategic Development WESTINGHOUSE PROPRIETARY DATA This document contains information proprietary to Westinghouse Electric Corporation; it is submitted in confidence and is to be used solely for the purpose for which it is furnished and returned upon request. This document and such information is not to be reproduced, transmitted, disclosed, or used otherwise in whole or in part without prior written authorization of Westinghouse Electric Corporation, Nuclear and Advanced Technology Division .
C!!l Westinghouse Electric Corporation 1993, All Rights Reserved Westinghouse Electric Corporation Nuclear and Advanced Technology Division Engineering Technology Department P. 0. Box 355 Pittsburgh, PA 15230-0355
ABSTRACT Presented in this report are results of a Salem Units 1&2 Small Break Loss of Coolant Accident (LOCA) analysis performed using the Westinghouse 1985 small break LOCA Emergency Core Cooling System (ECCS) Evaluation Model incorporating NOTRUMP analysis technology. The analysis was performed at enhanced plant conditions mutually agreed upon by Westinghouse and the customer, Public Service Electric and Gas (PSE&G) Corporation.
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ENGINEERING ANALYSIS REPORT:
SMALL BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR THE FUEL UPGRADE I MARGIN RECOVERY PROGRAM FOR SALEM UNITS 1 AND 2 TABLE OF CONTENTS Table of Contents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii List of Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv List of Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v 1.0 Executive Summary . . . . . . . . . . . . . . . .- . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 Introduction . . . . . . . . . . . . . . . . . . . . *. . . . . . . . . . . . . . . . . . . . . . . . 2 3 .0 LOCA Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4.0 Small Break LOCA Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4.1 Purpose of Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4.2 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4.3 Analytical Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 4.4 Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 4.5 Methodology . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.5.1 Break Spectrum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 4.5.2 Reactor Coolant System Temperature . . . . . . . . . . . . . . . . . . . . . 7 4.6 Small Break LOCA Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.6.1 2-Inch Equivalent Diameter Cold Leg Break . . . . . . . . . . . . . . . . . 8 4.6.2 1.5-Inch Equivalent Diameter Cold Leg Break . . . . . . . . . . . . . . . . 9 4.6.3 3-Inch Equivalent Diameter Cold Leg Break . . . . . . . . . . . . . . . . . 9 4.6.4 4-Inch Equivalent Diameter Cold Leg Break . . . . . . . . . . . . . . . . . 9 4.7 Limiting Temperature Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 5.0 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 6.0 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . 12 7.0 Appendix A: Tables and Figures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 iii
LIST OF TABLES Small Break LOCA Analysis Time Sequence of Events 2 List of Important Small Break LOCA Analysis Parameters for Salem 3 Small Break LOCA Fuel Cladding Results iv
LIST OF FIGURES Code Interface Description for the Small Break Model 2 Small Break Hot Rod Power Shape 3 Small Break Safety Injection Flow Rate 4 Reactor Coolant System Depressurization Transient, 2-Inch Cold Leg Break, High TAva 5 Core Mixture Level, 2-Inch Cold Leg Break, High TAva 6 Clad Average Temperature - Hot Rod, 2-Inch Cold Leg Break, High TAva 7 Core Exit Steam Flow, 2-Inch Cold Leg Break, High TAva 8 Heat Transfer Coefficient, 2-Inch Cold Leg Break, High TAva 9 Fluid Temperature - Hot Spot, 2-Inch Cold Leg Break, High TAva 10 Cold Leg Break Mass Flow, 2-Inch Cold Leg Break, High TAva 11 ECCS Pumped Safety Injection, 2-Inch Cold Leg Break, High TAva 12 Reactor Coolant System Depressurization Transient, 4-Inch Cold Leg Break, High TAva 13 Core Mixture Level, 4-Inch Cold Leg Break, High TAva 14 Clad Average Temperature - Hot Rod, 4-Inch Cold Leg Break, High TAva 15 Core Exit Steam Flow, 4-Inch Cold Leg Break, High TAva 16 Heat Transfer Coefficient, 4-Inch Cold Leg Break, High TAva 17 Fluid Temperature - Hot Spot, 4-Inch Cold Leg Break, High T Ava 18 Cold Leg Break Mass Flow, 4-Inch Cold Leg Break, High TAva 19 ECCS Pumped Safety Injection, 4-Inch Cold Leg Break, High TAva 20 Reactor Coolant System Depressurization Transient, 3-Inch Cold Leg Break, High TAva 21 Core Mixture Level, 3-Inch Cold Leg Break, High TAva 22 Clad Average Temperature - Hot Rod, 3-Inch Cold Leg Break, High TAva 23 Core Exit Steam Flow, 3-foch Cold Leg Break, High TAva v
24 Heat Transfer Coefficient, 3-Inch Cold Leg Break, High TAva 25 Fluid Temperature - Hot Spot, 3-Inch Cold Leg Break, High TAva 26 Cold Leg Break Mass Flow, 3-Inch Cold Leg Break, High TAva 27 ECCS Pumped Safety Injection, 3-Inch Cold Leg Break, High TAva 28 Reactor Coolant System Depressurization Transient, 1.5-Inch Cold Leg Break, High TAva 29 Core Mixture Level, 1.5-Inch Cold Leg Break, High TAva 30 Clad Average Temperature - Hot Rod, 1.5-Inch Cold Leg Break, High TAva
- 31. Core Exit Steam Flow, 1.5-Inch_ Cold Leg Break, High TAvo 32 Heat Transfer Coefficient, 1.5-Inch Cold Leg Break, High TAva 33 Fluid Temperature - Hot Spot, 1.5-Inch Cold Leg Break, High TAva 34 Cold Leg Break Mass Flow, 1.5-Inch Cold Leg Break, High TAva 35 ECCS Pumped Safety Injection, 1.5-Inch Cold Leg Break, High TAva 36 Reactor Coolant System Depressurization Transient, 2-Inch Cold Leg Break, Low TAva 37 Core Mixture Level, 2-Inch Cold Leg Break, Low TAvo
. 38 Clad Average Temperature - Hot Rod, 2-Inch Cold Leg Break, Low TAva 39 Core Exit Steam Flow, 2-Inch Cold Leg Break, Low TAva 40 Heat Transfer Coefficient, 2-lnch Cold Leg Break, Low TAvo 41 Fluid Temperature - Hot Spot, 2-lnch Cold Leg Break, Low TAva 42 Cold Leg Break Mass Flow, 2-Inch Cold Leg Break, Low TAva 43 ECCS Pumped Safety Injection, 2-lnch Cold Leg Break, Low TAva 44 Core Power Transient vi
ENGINEERING ANALYSIS REPORT:
SMALL BREAK LOSS-OF-COOLANT ACCIDENT ANALYSIS FOR THE FUEL UPGRADE I MARGIN RECOVERY PROGRAM FOR SALEM UNITS 1 AND 2 1.0 EXECUTIVE
SUMMARY
This engineering analysis report contains information regarding the small break loss-of-coolant accident (LOCA) analyses and evaluations performed in support of th~ Fuel Upgrade I Margin Recovery Program for the Salem Units 1 and 2. The following important plant conditions and features were incorporated:
- 1) A licensed core power of 3411 MWt, FQT=2.4, F,18 =1.65,
- 2) A range of nominal vessel average operating temperatures from 566°F to 580°F, accounting for a +/-5°F allowance for uncertainties, *
- 3) A uniform steam generator tube plugging level to 25%,
- 4) Reactor coolant system (RCS) primary side pressure of 2300 psia which includes a + 50 psia allowance for uncertainties,
- 5) Thermal Design Flow of 82500 gpm, and
- 6) PERFORMANCE+, VANTAGE-SH, and STANDARD(275 psig backfill) fuel.
The analyses were performed with the Westinghouse 1985 small break LOCA ECCS Evaluation Model incorporating NOTRUMP analysis technology. This analysis will replace the previous analysis which used WFLASH technology. A spectrum of 1.5-inch, 2-inch, 3-inch, and 4-inch equivalent diameter cold leg breaks was performed. A sensitivity of the limiting transient to the reactor coolant system average temperature was also performed.
- A break spectrum supporting the high nominal vessel average temperature, TAvo=580°F, was performed.
Peak clad temperatures of 685°F, 1580°F, 1508°F, and 1343°F were calculated for the 1.5-inch, 2-inch, 3-inch, and 4-inch cold leg breaks, respectively, thus identifying the 2-inch equivalent diameter break as limiting. A sensitivity to low nominal vessel average temperature, TAvo=566°F, was performed. The calculated peak clad temperature was 1558°F, identifying the 2-inch equivalent diameter cold leg break, high nominal vessel average temperature, as the limiting case. Table 2 lists the important plant operating characteristics assumed in the analysis. Tables 1 and 3 summarize the results. Figures 4 through 43 provide the following parameters for the 1.5-, 2-, 3-, and 4-inch breaks:
- The reactor coolant system depressurization transient
- The core mixture level*
- The hot rod clad average temperature
- The core outlet vapor flow
- The hot rod heat transfer coefficient
- The hot spot fluid temperature
- The total break flow
- The pumped safety injection flow
- Levels are relative to the inside bottom of the lower plenum.
Based upon the results provided in this engineering analysis report, compliance with the Acceptance Criteria prescribed by 10 CPR 50.46 is demonstrated for the Salem Units 1 and 2 Fuel Upgrade I Margin Recovery program as described in Section 4.0.
2.0 INTRODUCTION
The following documentation includes the acceptance criteria for the LOCA, the analytical models used to evaluate the LOCA and a description of the analysis methodology. Results are provided in the form of tables and figures. Descriptions of the transients are also provided. It was determined that no design or regulatory limit related to LOCA would be exceeded.
3.0 LOCA EVENTS A loss-of-coolant accident (LOCA) is the result of a pipe rupture of the reactor coolant system (RCS) pressure boundary. For the analyses reported here, a small break is defined as a rupture of the RCS piping with a cross-sectional area less than 1.0 ft2, in which the normally operating charging system flow is not sufficient to sustain pressurizer level and pressure. This event is considered an American Nuclear Society (ANS)
Condition III event which are faults which may occur very infrequently during the life of a plant.
. The Acceptance Criteria for the LOCA are described in 10 CFR 50.46 (Reference 1) as follows:
A. The calculated peak fuel element clad temperature does not exceed 2200°F.
B. The amount of fuel element cladding that reacts chemically with water or steam to generate hydrogen, does not exceed 1 percent of the total amount of fuel rod cladding ..
C. The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.
D. The core remains amenable to cooling during and after the break.
E. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.
The criteria were established to provide a significant margin in emergency core cooling system (ECCS)
- performance following a LOCA.
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4.0 SMALL BREAK LOCA ANALYSIS 4.1 Purpose of Analysis The purpose of analyzing the small break LOCA is to demonstrate that conformance with the 10 CPR 50.46 requirements listed previously is maintained for the conditions associated with the Fuel Upgrade I Margin Recovery Program.
4.2 Assumptions The small break analysis was based upon the input assumptions and conditions listed in Table 2.
4.3 ** Analytical Model For small breaks (less than 1.0 ff) the NOTRUMP computer code (References 3 and 4) is employed to calculate the transient depressurization of the reactor coolant system (RCS) as well as to describe the mass and energy of the fluid flow through the break. The NOTRUMP computer code is a state-of-the-art one-dimensional general network code incorporating a number of advanced features. Among these advanced features 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. The NOTRUMP small break LOCA emergency core cooling system (ECCS) Evaluation
. Model 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 RCS model is nodalized into volumes interconnected by flowpaths. The broken loop is modeled*
explicitly, while the intact loops are lumped into a second loop. Transient behavior of the system is determined from the governing conservation equations of mass, energy, and momentum. The multinode capability of the program enables explicit, detailed spatial representation of various system components which, among other capabilities, enables a proper calculation of the behavior of the loop seal during a LOCA.
The reactor core is represented as heated control volumes with associated phase separation models to permit transient mixture height calculations.
Clad thermal analyses are performed with a version of the LOCTA-IV code (Reference 6) using the NOTRUMP calculated core pressure, fuel rod power history, uncovered core steam flow and mixture heights as boundary conditions (Figure 1).
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4.4 Analysis A LOCA is the result of a pipe rupture of the RCS pressure boundary. For the analysis reported here, a small break is defined as a rupture of the RCS piping with a cross-sectional area less than 1.0 ft2. This event is an American Nuclear Society (ANS) Condition III event, in that it is an infrequent fault that may occur during the lifetime of the Salem Nuclear Generating Station.
Salem operates three dedicated emergency diesel generators for each unit. For the limiting power availability condition, Loss-Of-Offsite-Power (LOOP), the most limiting single active failure for small break LOCA has been determined to be the failure of a diesel generator which results in the loss of one high head Charging/SI pump (HHSI), and one intermediate head pump (IHSI). Therefore, credit can be taken for one HHSI, one IHSI, and two motor driven auxiliary feedwater pumps. The Charging/SI subsystem ties directly into the RCS cold legs via 1.5-inch diameter connections. For the small break LOCA PCT analysis, any assumed RCS break size greater than or equal to the corresponding Charging/SI branch line diameter (1.5-inch) should consider safety injection flows calculated assuming one of the four Charging/SI lines spilling to containment backpressure. For smaller RCS break sizes ( <- 1.5-inch), safety injection flows calculated assuming one of the four Charging/SI lines spilling to RCS backpressure may be used. The IHSI and RHR subsystems tie into the accumulator line at a location between two check valves in the accumulator line. A break between the accumulator check valves, in the IHSI or in the RHR systems would not result in a LOCA since the accumulator line check valve closest to the cold leg boundary would prevent loss of inventory from the RCS.
However, a break in the accumulator line between the check valve closest to the RCS pressure boundary and the cold leg nozzle could result in a loss of RCS inventory and IHSI and RHR injection flow. For breaks smaller than the area equivalent to the 10-inch accumulator line diameter, the safety injection flows calculated
. assuming one of the four IHSI and RHR lines spilling to RCS backpressure may be used. Therefore any breaks analyzed for 1.5-inch through 10-inch equivalent diameter break sizes will use safety injection flows calculated assuming that one of the four Charging/SI lines will spill to containment backpressure, and one of the four IHSI and RHR lines will spill to RCS backpressure. To minimize delivery to the core, the branch line with the least resistance is chosen as the spilling line.
Should a small break LOCA occur, 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 1715 psia (including uncertainties) is reached. LOOP is assumed 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 (including uncertainties) is reached. These countermeasures limit the consequences of the accident in two ways:
A. 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 LOCA analysis for the boron content of the injection water.
- However, an average RCS/sump mixed boron concentration is calculated to ensure that the post-LOCA core remains subcritical. In addition, credit is taken in the small break LOCA analysis for the insertion of Rod Cluster Control Assemblies (RCCAs) subsequent to the reactor trip signal, while assuming the most reactive RCCA is stuck in the full out position.
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B. Injection of borated water ensures sufficient flooding of the core to prevent excessive clad temperatures.
Before the break occurs, the plant is in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary sys.tern. *After the small break LOCA is initiated, system depressurization to the pressurizer low-pressure setpoint leads to reactor trip, control rod insertion, and core shutdown. During the early part of the transient, upward flow through the core is maintained by the reactor coolant pumps as they coast down following LOOP. Heat from fission product decay, hot internals, and the vessel continues to be transferred to the RCS. The heat transfer between the RCS and the secondary system may be in either direction depending on the relative temperatures. Early in the transient, heat addition to the secondary results in increased steam generator secondary pressure, which leads to steam relief via the main steam safety valves.
Makeup to the secondary is automatically provided by the auxiliary feedwater pumps. The safety injection signal, initiated from the pressurizer low-pressure SI setpoint, isolates normal feedwater flow by closing the main feedwater control and bypass valves. Heat transfer to the secondary aids break flow in the reduction of RCS pressure. The reduction in RCS inventory*, due to break flow, and core boil off, due to decay heat, may lead to a partial core uncovery. Pumped ECCS injection replaces system mass lost to break flow, recovering the core (if core uncovery occurs), and providing adequate inventory to accommodate continued boiloff due to decay heat with no subsequent core uncovery. Passive injection from the accumulators occurs when the cold leg pressures reach the accumulator injection pressure.
Safety injection systems consist of gas pressurized accumulator tanks and pumped injection systems. The small break LOCA analysis assumed nominal accumulator water volume (850 ft3) with a cover gas pressure of
. 592 psia. Minimum ECCS availability is assumed for the analysis, and pumped ECCS is conservatively assumed to be at the maximum RWST temperature. Assumed pumped safety injection characteristics as a function of RCS pressure used as boundary conditions in the analysis are shown in Figure 3. The safety injection flow rates presented are based on pump performance curves degraded from the design head (7 % for High Head Safety Injection (HHSI), 10% for Intermediate Head Safety Infection (IHSI) ) and an assumed charging system branch line imbalance of 10.5 gpm for HHSI, and 12 gpm for IHSL The effect of flow from the RHR pumps is not considered in the small break LOCA analyses since the shutoff head is lower than the RCS pressure during the time portion of the transient considered here. Safety injection is delayed 32 seconds after the occurrence of the low pressure condition. This accounts for signal initiation, diesel generator startup and emergency power bus loading consistent with the assumed loss of offsite power coincident with reactor trip as well as the delay involved in aligning the .valves and bringing the pumps up to speed.
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4.5 Methodology
- In order to determine the conditions that produced the most limiting SBLOCA case, a total of five cases were examined. These cases included the investigation of variables including break size and RCS temperature to ensure that the most severe postulated SBLOCA was evaluated. The following discussions provide insight into the analyzed conditions ..
4.5 .1 Break Spectrum Typically, a spectrum of cold leg breaks is analyzed. The cold leg has been determined to be the limiting break location. Breaks in the cold leg location have the greatest propensity for uncovering of the reactor core since they result in higher reactor coolant system pressure and consequently lower safety injection flow during the transient than equivalent breaks in other locations in the system. For cold leg breaks, effective depressurization of the system cannot occur until steam produced from the core decay heat is able to travel through the RCS loops and out the break. Steam flow out the cold leg break is minimized until the period when the liquid in the piping from the steam generator outlet to the reactor coolant pump inlet (loop seal) is cleared.
The limiting break, and the spectrum analyzed to identify the limiting break, are functions of the plant design and the ECCS design. Smaller breaks will depressurize more slowly which minimizes the amount of safety
- injection flow to the RCS. However, the inventory depletion is also lower for smaller breaks. The integral effects of the break flow and safety injection flow combine to result in a break size which will be limiting .
. Due to the plant design and ECCS flow characteristics, the following spectrum of cold leg breaks were considered:
4-inch equivalent diameter cold leg break:
This case represents the largest break in the discrete spectrum for which calculations were performed. Because of accumulator/intermediate head safety Injection (IHSI) interactions, for all break sizes less than the accumulator line inside diameter (10 inches), IHSI is assumed to spill to RCS backpressure. Charging/SI flow is conservatively assumed to spill to containment backpressure (14.7 psia), for all breaks considered. While this break size will result in substantial.depletion of the RCS primary mass inventory, the rapid RCS depressurization will result in a high safety injection flow rate earlier in the transient. For this analysis, these competing effects result in the 4-inch break size being a non-limiting break size.
3-inch equivalent diameter cold leg break:
This case represents a break in the discrete spectrum which will result in a decreased rate of inventory depletion, but also results in decreased safety injection flow which is a result of the decreased RCS depressurization rate. These effects result in the 3-inch break size being a non-limiting break size.
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2-inch equivalent diameter cold leg break:
This case represents a break in the discrete spectrum which results in both a more reduced rate of inventory depletion and a more reduced safety injection flow since the depressurization rate is slower. Due to these synergistic effects, the 2-inch equivalent break size was determined to be the limiting break size.
1.5-inch equivalent diameter cold leg break:
This case represents the smallest break in the discrete spectrum which will result in a highly reduced rate of inventory depletion, but also results in a stalled safety injection flow delivery during the transient due to the RCS pressure remaining at a high and relatively constant value for much of the transient. Although the depressurization takes longer, the reduced loss of.inventory results in a less limiting LOCA.
4.5 .2 Reactor Coolant System Temperature
- Reduced operating temperatures typically result in peak cladding temperature benefits for the small break LOCA. Due to competing effects and the complex nature of the SBLOCA transients, there have been some instances where more limiting results have been observed for the case of reduced operating temperatures. For this reason, sensitivity analyses were performed for the two sets of operating temperatures from Table 2.
4.6 Small Break LOCA Analysis Results The evaluations to determine the limiting break size were completed before the limiting vessel average temperature sensitivity was performed, and are explained in greater detail in sections 4. 7 .1 and 4. 7 .2.
The evaluations were completed based on a vessel average temperature of 580°F. The limiting 2-inch break conditions resulted in a PCT of 1580°F. The 4-inch, 3-inch and 1.5-inch breaks were analyzed at the higher RCS (vessel) temperature and resulted in PCTs of 1343°F, 1508°F and 685°F, respectively, which showed that the 2-inch break was limiting. The vessel average temperature sensitivity was performed to support a vessel average temperature of 566 °F, and resulted in a PCT of 1558°F. Therefore, the high vessel average temperature of 580°F,represents the more limiting operating condition. Thus the small break LOCA results for this case bound operation in the range of vessel average temperatures between 566°F and 580°F (accounting for a +/-5°F uncertainty).
Complete results of the analyses are provided in Tables 2 and 3. A detailed discussion of the 2-inch limiting break is provided below, followed by brief discussions of the 1.5-, 3-, and 4-inch break transients at the limiting temperature condition. A discussion of the transients for the limiting temperature condition is contained in section 4. 7.
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4.6.1 2-Inch Equivalent Diameter Cold Leg Break:
Upon initiation of the 2-inch break (Figures 4 to 11), there was a slow depressurization of the RCS (Figure 4). At 188.8 seconds into the transient, the pressure decreased to the pressurizer low-pressure setpoint conservatively modeled at 1715 psia, which generated the reactor trip signal.
The LOOP was assumed to occur coincident with reactor trip plus trip delays. The pressurizer low-pressure safety injection setpoint is also modeled at 1715 psia. For this analysis the following modeling assumptions were made: reactor trip coincident with the LOOP assumption initiated the reactor coolant pump trip, turbine trip and isolation, and auxiliary feedwater pump actuation. The safety injection signal initiated main feedwater pump trip and isolated with the appropriate delays.
The auxiliary feedwater system began delivering flow to the steam generator secondary sides about 60 seconds after reactor trip. The section of the main feedwater piping, downstream of the main-aux feedwater interface, has been flushed of main feedwater by aux feedwater flow at about 420 seconds, so credit was taken for the lower auxiliary feedwater enthalpy at that time. The steam generator secondary side was effectively isolated (since no steam dump operation is assumed) and pressurized to the steam generator safety valve setpoint of 1117 psia, and steam was released through the safety valves.
Safety injection flow was assumed to begin 32 seconds after initiation of the SI-signal, conservatively allowing for signal processing delay time, diesel generator start-up time, flowpath alignment, sequencing of equipment, and time for the pumps to come up to speed. Safety injection flow delivery occurred at approximately 220 seconds after initiation of the break. Throughout this initial depressurization period, there was only liquid flow out of the break.
By approximately 200 seconds, the RCS depressurization was slowed as the RCS tended to equilibrate at a pressure condition just above the steam generator secondary side pressure. This quasi-equilibrium period lasted until approximately 1200 seconds. During this period, the loop seals remained plugged with liquid, the break flow was all liquid and the steam generators were active heat sinks. As a result of the liquid plugs in the loop seals, steam produced within the RCS was unable to vent out the break. Consequently, the RCS reached a quasi-equilibrium pressure condition in which the net rate of volumetric steam production within the RCS was balanced by the net rate at which RCS volume became available for occupancy by vapor.
The quasi-equilibrium condition was upset at 1200 seconds when the loop seal in the broken loop began the clearing and steam venting process. This occurred as a result of the liquid in the downleg of the loop seal being depressed low enough to allow steam to begin to vent through the loop seal.
The loop seal replugged after a few seconds and remained plugged until continuous venting began at 1310 seconds. Steam which was previously trapped in the RCS was able to vent to the break. Break flow, which was previously all liquid, now began the transition to a flow that was predominantly vapor. The vapor flow out of the break was an effective heat removal mechanism and consequently became the primary mechanism for decay heat energy removal. The volumetric flow of vapor out the break exceeded the volume of vapor produced in the core. As a result, the rate of RCS depressurization increased.
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Following loop seal clearing, the depressurization of the RCS was accompanied by a decrease in the inner vessel mixture level as the core boil-off rate exceeded the rate of safety injection (Figure 5) plus condensation in the system. Although the inner vessel mixture continued to decrease, signs of imminent turnaround were becoming evident. Around 1900 seconds, the liquid flow out the break was replaced by vapor flow only, the vapor flow rate turned around at about 2500 seconds, and the pumped safety injection continued to increase due to the RCS depressurization. Around 1500 seconds, the pumped safety injection flow rate began exceeding the core boil-off rate and break flow rate so that core mixture level recovery began. Core recovery occurred at 2887 seconds. The system did not depressurize sufficiently, prior to core recovery, to allow accumulator injection. The transient was terminated at 8000 seconds.
4.6.2 1.5-Inch Equivalent Diameter Cold Leg Break The 1.5-inch break (Figures 28-35) was characterized by an extremely slow depressurization (Figure 28). With such a small break size, the safety injection flow was able to exceed the break flow by the time that the loop seal cleared. This phenomena permitted the core to remain covered, (Figure 29).
As the RCS further depressurized, additional safety injection flow was provided, and the break flow decreased further. As with the 2-inch break, the RCS did not depressurize to the point to allow the accumulators to inject. With the SI flow exceeding the break flow and the core covered, the transient was terminated at 8857 seconds.
4.6.3 3-Inch Equivalent Diameter Cold Leg Break
. The 3-inch break (Figures 20-27) was characterized by a fairly slow RCS depressurization and core recovery (Figures 20 and 21). Upon reactor trip, the RCS depressurized to secondary side conditions. At approximately 540 seconds the loop seal cleared, replugged momentarily, and then cleared again. At this time, the break flow was predominately vapor, which allowed for a high break flow rate. The safety injection flow was unable to match the break flow, causing the core to uncover at 920 seconds. The core continued to uncover until, at 1550 seconds, the RCS had depressurized enough to allow the safety injection flow to exceed the break flow. The RCS depressurized further, and at 1830 seconds the accumulators injected, resulting in rapid core level recovery. The core was completely covered at 2650 seconds. With the core covered, and the break flow being exceeded by the safety injection flow, the transient was terminated at 3000 seconds.
4.6.4 4-Inch Equivalent Diameter Cold Leg Break The 4-inch break (Figures 12-19) was characterized by a relatively quick RCS depressurization and core recovery (Figures 12 and 13). The loop seal venting phenomena occurred at 270 seconds which allowed for a high break flow rate at a time where the high RCS pressure did not allow for much SI flow. The core mixture level quickly dropped until after 900 seconds. At that time, liquid flow out
- the break had stopped, the vapor flow out of the break stabilized around 80 lb/sec which was exceeded by the SI-flow, and accumulator injection began at 894 seconds. The result was a quick rise 9
in core mixture level, but was coupled with void collapse in the lower plenum and lower core nodes.
This resulted in a quasi-equilibrium at a core mixture level around 1 ft. away from complete recovery. The RCS pressure also remained quasi-static during this time until about 1200 seconds when the pressure began to decrease again due in part to the decreased break flow and increased SI flow. This slight depressurization was enough to activate the accumulators again and soon after the core was recovered (1313 seconds). Since the safety injection flow rate exceeded the core boil-off rate and break flow rate, there was assurance that the core would remain covered and the transient was terminated at 1500 seconds.
- 4. 7 Limiting Temperature Conditions The following section describes the analysis which was performed in order to determine the limiting temperature condition. This sensitivity analysis was based on the limiting 2-inch break case from the break spectrum analyses previously described. The condition to be examined was:
Vessel average temperature of 566°F or 580°F.
The nominal TAVG operating window of 566°F to 580°F required analysis due to the concerns described in section 4.5.)'.
~
Per Figures 4-11 (High TAvo) and Figures 36-43 (Low TAva), the transients appear to behave very
. similarly with probably the greatest difference being the core mixture levels (Figures 5 and 37). The low TAvo core mixture level does not drop as soon or as far as the high TAvo core mixture level does.
This phenomena can be attributed to the complex interactions which result in loop seal clearing. For the low TAvo transient, continuous steam flow is established following the initial loop seal clearing.
For the high TAvo transient, replugging of the loop seal leads to establishing continuous steam flow later than for the low TAvo case. As a consequence, the liquid break flow is greater for the high temperature case which leads to a reduced primary inventory during the core uncovery period. It is the reduced primary inventory that leads to the deeper core uncovery for the high temperature case.
As a result, the high TAvo transient becomes the limiting condition for Salem.
10
5.0 CONCLUSION
S
- A four break spectrum was performed which included the 1.5, 2, 3, and 4-inch breaks. An additional analysis was performed to ensure that the limiting condition for the vessel temperature, in the allowable operating range, was established. The limiting small break was the 2-inch equivalent diameter cold leg break which.resulted in a calculated peak cladding temperature of 1580°F. The limiting condition was determined to be the high reactor coolant system average temperature (TAva=580°F).
Compliance with the following Federal regulatory limits established in 10 CFR 50.46 has been demonstrated for the Salem Nuclear Power Station:
- 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
- 4) generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
Calculated changes in core geometry shall be such that the core remains amenable to cooling. 1
- 5) After any calculated successful 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. 2 1
The appropriate core geometry was modeled in the analysis. The results based on this geometry satisfy the PCT criterion of 10 CFR 50.46 and consequently, demonstrate the core remains amenable to cooling.
2 Long-term core cooling considerations are evaluated on a cycle specific basis and are not reported here.
11
6.0 REFERENCES
- !) "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 August 31, 1992, (57 39353).
- 2) U.S. Nuclear Regulatory Commission 1975, "Reactor Safety Study - An Assessment of Accident Risks in U.S. commercial Nuclear Power Plants," WASH-1400, NUREG-75/014.
- 3) Meyer, P .E., "NOTRUMP - A Nodal Transient Small Break and General Network Code,"
WCAP-10079-P-A, (Proprietary), August 1985.
- 4) Lee, N. et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A, (Proprietary), August 1985.
- 5) "Generic Evaluation of Feed water Transients and Small Break Loss-of-Coolant Accidents in Westinghouse - Designed Operating Plant," NUREG-0611, January 1980.
- 6) Bordelon, F. M. et al., "LOCTA-IV Program: Loss-of-Coolant Transient Analysis,"
WCAP-8305, June 1974, WCAP-8301 (Proprietary), June 1974.
12
Appendix A Tables and Figures 13
TABLE 1 SMALL BREAK LOCA ANALYSIS TIME SEQUENCE OF EVENTS Break Spectrum, (High TAva) 4-inch 3-inch 2-inch 1.5-inch Break Occurs (sec) 0.0 0.0 0.0 0.0 Reactor Trip Signal (sec) 22.88 41.72 188.84 104.02 Safety Injection Signal (sec) 22.88 41.72 188.84 112.76 Safety Injection Begins (sec) 54.1 73.5 220.5 144.5 Loop Seal Venting (sec) 320 588 1308 2604 Top Of Core Uncovered (sec) 540 922 1918 NIA1 Accumulator Injection Begins (sec) 810 1830 NIA NIA Peak Clad Temperature Occurs (sec) 883 1532 3275 NIA1 Top Of Core Covered (sec) 1300 2650 6630 NIA1 Results for the 2-inch -break size High TAVG Low TAva
- Break Occurs (sec) 0.0 0.0 Reactor Trip Signal (sec) 188.84 55.21 Safety Injection Signal (sec) 188.84 55.21 Safety Injection Begins (sec) 220.5 91.0 Loop Seal Venting (sec) 1308 1448 Top Of Core Uncovered (sec) 1918 2152 Accumulator Injection Begins (sec) NIA NIA Peak Clad Temperature Occurs (sec) 3275 3333 Top Of Core Covered (sec) 6630 6560 1
Momentary core uncovery occurred during prelude to loop seal clearing. Extended core uncovery was not experienced. The momentary temperature excursion maintains clad temperatures well below 700°F.
14
TABLE2 INPUT PARAMETERS USED IN THE SMALL BREAK LOCA ANALYSIS Parameter High TAVG Low TAva Reactor core rated thermal power 1 , (MWt) 3411 3411 Peak linear power1 *2 , (kw/ft) 12.812 12.812 Total peaking factor (FQ ~ at peak2 2.40 2.40 Power shape See Figure 2 FaH 1.65 1.65 Fuei3 17 x 17 17 x 17 Accumulator water volume, nominal (ft3/accumulator) 850 850 Accumulator tank volume, nominal (ft3/accumulator) 1350 1350 Accumulator gas pressure, minimum (psia) 592 592 Pumped safety injection flow See Figure 3 Steam generator tube plugging level (%)4 25 25 Thermal Design Flow/loop, (gpm) 82,500 82,500 Vessel average temperature, (°F) 580 566 Reactor coolant pressure, (psia) 2300 2300 Min. aux. feedwater flowrate/loop, (lb/sec)5 44.13 44.13 1
Two percent is added to this power to account for calorimetric error. Reactor coolant pump heat is not modeled in the SBLOCA analyses.
2 This represents a power shape corresponding to a peaking factor envelope (K(Z~ based on FQT =2.40.
3 The Performance+ fuel features analyzed included ZIRLO' cladding. Zirc-4 cladding was analyzed and ZIRLO' cladding was determined to be limiting. Results are not included in this report because the differences in PCT's between fuel cladding materials was insignificant. The analysis bounds operation with PERFORMANCE+, VANTAGE-SH, and STANDARD (275 psig backfill) fuels.
4 Uniform.
5 Flowrates per steam generator.
15
TABLE 3 SMALL BREAK LOCA ANALYSIS FUEL CLADDING RESULTS Break Spectrum, (High TA yo) 4-inch 3-inch 2-inch 1.5-inch Peak Clad Temperature (°F) 1343 1508 1580 N!A1 Peak Clad Temperature Location (ft) 11.25 11.50 11.50 NIA Peak Clad Temperature Time (sec) 883 1532 3275 NIA Local Zr/H 20 Reaction, Max (%) 0.1323 0.7343 1.5456 0.0333 Local Zr/HP Reaction Location (ft) 11.25 11.50 11.75 11.00 Total Zr/H 20 Reaction(%) < 1.0 < 1.0 < 1.0 < 1.0 Hot Rod Burst Time (sec) No Burst No Burst No Burst No Burst Hot Rod Burst Location (ft) NIA NIA NIA NIA
- Results for the 2-inch break size High TAYO Low TAYo Peak Clad Temperature (°F) 1580 1558 Peak Clad Temperature Location (ft) 11.50 11.75 Peak Clad Temperature Time (sec) 3275 3333 Local Zr/H20 Reaction, Max(%) 1.5456 1.3064 Local Zr/H 20 Reaction Location (ft) 11.75 11.75 Total Zr/HP Reaction(%) < 1.0 < 1.0 Hot Rod Burst Time (sec) No Burst No Burst Hot Rod Burst Location (ft) NIA NIA 1
Momentary core uncovery occurred during prelude to loop seal clearing. Extended core uncovery was not experienced. The momentary temperature excursion maintains clad temperatures well below 700°F.
16
N L 0 0 T CORE PRESSURE, CORE c
R FLOW, MIXTURE LEVEL, T u AND FUEL ROD POW.ER msTORY A
M p
--- - 0 < TIME < CORE RE-COVERED 1*
CODE INTERFACE DESCRIPTION FOR PUBLIC SERVICE ELECTRIC AND GAS COMPANY SMALL BREAK MODEL SALEM NUCLEAR GENERATING STATION Figure 1 17
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Small Break Hot Rod Power Shape PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Figure 2 18
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SMALL BREAK SAFETY INJECTION PUBLIC SERVICE ELECTRIC AND GAS COMPANY FLOW RATE SALEM NUCLEAR GENERATING STATION Figure 3 19
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40 35
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Core Mixture Level PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, High T Ava>
SALEM NUCLEAR GENERATING STATION Figure 5 21
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. 22
2500 u
(I)
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Core Exit Steam Flow PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, High TA vol SALEM NUCLEAR GENERATING STATION Figure 7 23
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Heat Transfer Coefficient - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 8 24
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SALEM NUCLEAR GENERATING STATION Figure 9 25
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SALEM NUCLEAR GENERATING STATION Figure 11 27
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SALEM NUCLEAR GENERATING STATION Figure 12 28
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- PUBLIC SERVICE ELECTRIC AND GAS COMPANY (4-lnch Break, High T Avo}
SALEM NUCLEAR GENERATING STATION Figure 15 31
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Heat Transfer Coefficient - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY (4-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 16 32
1200 I
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SALEM .NUCLEAR GENERATING STATION Figure 17 33
~.- .
3000 2500 u
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- TI ME (sec)
Cold Leg Break Mass Flow PUBLIC SERVICE ELECTRIC AND GAS COMPANY (4-lnch Break, High T Avo>
SALEM NUCLEAR GENERATING STATION Figure 18 34
100 -** -- ..
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ECCS Pumped Safety Injection PUBLIC SERVICE ELECTRIC AND GAS COMPANY (4-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 19 35
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Reactor Coolant System Depressurization Transient PUBLIC SERVICE ELECTRIC AND GAS COMPANY (3-lnch Break, High T Ava>
SALEM NUCLEAR GENERATING STATION Figure 20 36
40 35"I
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- 1000** 1500 2000 2500 3000 TIME (sec)
Core Mixture Level PUBLIC SERVICE ELECTRIC AND GAS COMPANY (3-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 21 37
1600 I/ '
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Clad Average Temperature - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY (3-lnch Break, High T Ava>
SALEM NUCLEAR GENERATING STATION Figure 22 38
1000 u
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Core Exit Steam Flow PUBLIC SERVICE ELECTRIC AND GAS COMPANY (3-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 23 39
105 r-..
l.J....
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~
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100 0 500 ..
1000 1500 2000 2500 3000
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TIME (sec)
Heat Transfer Coefficient. - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY (3-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 24 40
1400 ,..!,\
I J
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400 2000 500 1000 1500 2000 2500 3000 TIME (sec)
Fluid Temperature - Hot Spot*
PUBLIC SERVICE ELECTRIC AND GAS COMPANY (3-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 25 41
2000
........... 1500 u -* .
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00 500 1000 1500. 2000 2500 3000 TIME (sec)
Cold Leg Break Mass Flow PUBLIC SERVICE ELECTRIC AND GAS- COMPANY (3-lnch Break, High T Aval
.SALEM NUCLEAR GENERATING STATION Figure 26 42
80
.. . .. -... ~ - - .
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__J 00 500 1000 1500 2000 **2500 3000 TIME (sec)
ECCS Pumped Safety Injection PUBLIC SERVICE ELECTRIC AND GAS COMPANY (3-lnch Break, High T AVCJ)
SALEM NUCLEAR GENERATING STATION Figure 27 43
2500 2250.
2000 a
- -en a.
..__, 1750 w
0:::
- J (f) 1500 (f) w ,.,
0:::
0.. .I 1250 I
I I
1000 7500 2000 4000 6000 8000 10000
- TI ME (sec)
Reactor Coolant System Depressurization Transient PUBLIC SERVICE ELECTRIC AND GAS COMPANY (1.5-lnch Break, High TAvol SALEM NUCLEAR GENERATING STATION Figure 28 44
40 t-i 35
+-' --
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100 2000 4000 6000 8000 10000 TIME (sec)
Core Mtxture Level PUBLIC SERVICE ELECTRIC AND GAS COMPANY (1.5-lnch Break, High TAva>
SALEM NUCLEAR GENERATING STATION Figure 29 45
700
,-..._ 650 '
Li...
(Jl Q)
IJ w
Ct: 600 f-
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w Q_
~ I w \
f-550 5000 2000 4000 6000 8000 10000 TIME (sec)
Clad Average Temperature - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY ( 1.5-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 30 46
400 u
w U1 ..
............ 200
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-600 0 2000 4000 6000 8000 10000 TIME (sec)
Core Exit Steam Flow PUBLIC SERVICE ELECTRIC AND GAS COMPANY (1.5-lnch Break, High T Aval SALEM NUCLEAR GENERATING STATION Figure 31 47
106 Li...
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I 105 0:::
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10 1 0 2000 4000 6000 8000 10000 TIME (sec)
Heat Transfer Coefficient - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY (1.5-lnch Break, High- TAvol SALEM NUCLEAR GENERATING STATION Figure 32 48
600 580 l.J....
Qi Q)
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~
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~ 540 w
Q_
- 2 w
I-520 5000 2000 4000 6000 8000 10000 TIME (sec)
Fluid Temperature - Hot Spot PUBLIC SERVICE ELECTRIC AND GAS COMPANY (1.5-lnch Break, High TAval SALEM NUCLEAR GENERATING STATION Figure 33 49
400
,. . .__ 300 u
(J)
Ul E
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=
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2000 6000 8000 10000 TIME (sec)
Cold Leg Break Mass Flow PUBLIC SERVICE ELECTRIC AND GAS COMPANY (1.5-lnch Break, High TAval SALEM NUCLEAR GENERATING STATION Figure 34 50
40 I
j-()
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(J) 30 "
E Ll
~
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(/')
L-J w
Q_
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00 2000 4000 6000 8000 10000 TIME (sec)
- PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION ECCS Pumped Safety Injection (1.5-lnch Break, High T Aval Figure 35 51
2500 2000 0 -
CJ]
a...
1500 II"\
w \
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(f) w *-
~
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500 00 2000 4000 6000 8000 Tl~E (sec)
Reactor Coolant System Depressurization Transient PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low T Ava>
SALEM NUCLEAR GENERATING STATION Figure 36 52
40 35 ...
-+--' - -
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Core Mixture Level PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low T Aval SALEM NUCLEAR GENERATING STATION Figure 37 53
1600 I \
I \
I 1400 \
\
\.
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w- 600 ~
~
' I 400 2000 4000 6000 8000 2000
- TI ME (sec)
Clad Average Temperature - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low T Aval SALEM NUCLEAR GENERATING STATION Figure 38 54
2500
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~ 2000 E
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- J 0
w 500 0:::
0 u
,., I or 1 .
0 2000 4000 6000 .. 8000 TIME (sec)
-~ .-
...*... *v* '** ..
Core Exit Steam Flow PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low T Ava>
SALEM NUCLEAR GENERATING STATION Figure 39 55
106 l..J....
I N
f-l..J....
I 105 a::::
I
- J f-m
........... 104 l..J....
l..J....
w 0
(.)
103
- a::::
w l..J....
(J) z
<(
a:::: 102 f-f-
<( / ......
I w ""' ,,,.-
10 1 0 2000 4000 6000 8000 TIME (sec)
Heat Transfer Coefficient - Hot Rod PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low TA vol SALEM NUCLEAR GENERATING STATION Figure 40 56
1600
{
1400
.......... I \
LL 1200 I \
(J'I - r
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IJ
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<( 800 ~
a::
w \
0.. L l I
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w 600 .r "\....,
I- .r '-t\_
400 2000 2000 4000 6000 8000 TIME (sec)
Fluid Temperature - Hot Spot PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low T Aval SALEM NUCLEAR GENERATING STATION Figure 41 57
800
,........ 600 u
Q)
Ul - .
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_J LL I~
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<(
w ff 0:::
(lJ 200
,,Lal 1.. 1.
- 111111 11111 :i 00 2000 4000 6000 8000 TIME (sec)
Cold Leg Break Mass Flow PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low T Ava>
SALEM NUCLEAR GENERATING STATION Figure 42 58
60
......... 50 '/ "'i-.... _..,.,.
u Q) ./
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LI... 30
(/)
0 w 20 0...
~
- J 0...
....J
- 10
- 00 2000 4000 6000 8000 TIME (sec)
ECCS Pumped Safety Injection PUBLIC SERVICE ELECTRIC AND GAS COMPANY (2-lnch Break, Low T Aval SALEM NUCLEAR GENERATING STATION Figure 43 59
8 z
3 Q
- c en
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N I
....0 WATTS/WATT AT POWER Core Power Transient PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION Figure 44 60