LOCA Mass and Energy Release Methodology · The mass and energy release analysis uses the...

LOCA Mass and Energy Release Methodology PR1400-Z-A-NR-14007-NP, Rev.0

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LOCA Mass and Energy Release Methodology

Revision 0

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November 2014

Copyright ⓒ 2014

Korea Electric Power Corporation &

Korea Hydro & Nuclear Power Co., Ltd

All Rights Reserved

LOCA Mass and Energy Release Methodology APR1400-Z-A-NR-14007-NP, Rev.0

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REVISION HISTORY

Revision Date Page Description

0 November 2014 All First Issue

This document was prepared for the design certification

application to the U.S. Nuclear Regulatory Commission and

contains technological information that constitutes intellectual

property.

Copying, using, or distributing the information in this

document in whole or in part is permitted only by the U.S.

Nuclear Regulatory Commission and its contractors for the

purpose of reviewing design certification application

materials. Other uses are strictly prohibited without the

written permission of Korea Electric Power Corporation and

Korea Hydro & Nuclear Power Co., Ltd.


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ABSTRACT

This technical report describes the methodology of the loss-of-coolant accident (LOCA) mass and energy(M/E) release analysis which is intended to be used for the U.S. Nuclear Regulatory Commission (NRC) Design Certification (DC) Application of the Advanced Power Reactor 1400 (APR1400). The purpose of this report is to explain the methodology used and demonstrate the applicability to the APR1400 LOCA mass and energy release analysis.

An analysis of LOCA M/E release for APR1400 was performed using the methodology described in this report. The analysis includes evaluation of the system of advanced design features - four EDGs for four trains of emergency core cooling system (ECCS), the fluidic device installed inside the safety injection tank (SIT-FD) and the in-containment refueling water storage tank (IRWST).

The mass and energy release analysis uses the acceptance criteria described in Standard Review Plan (SRP) Section 6.2.1.3 which supports adherence to General Design Criterion 50 and 10 CFR Part 50, Appendix K. The transient containment pressure is limited under the containment design pressure of 60 psig.

Analyses were performed for the spectrum of break locations and SI flows :

- Double-ended suction leg slot break with maximum SI pump flow

- Double-ended suction leg slot break with minimum SI pump flow

- Double-ended discharge leg slot break with maximum SI pump flow

- Double-ended discharge leg slot break with minimum SI pump flow

- Double-ended hot leg slot break

The limiting case of the mass and energy release calculation was the case of double-ended discharge leg slot break with maximum SI pump flow. The analysis shows that the results including the effects of the advanced design features, meets the acceptance criteria of Standard Review Plan.

The methodology as presented will be used when determining maximum containment pressure and temperature following a loss-of-coolant accident for the APR1400.


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TABLE OF CONTENTS

1 INTRODUCTION .................................................................................................. 1

2 DESIGN DESCRIPTION OF THE APR1400 ........................................................... 2

2.1 RCS General Description .......................................................................................................... 2

2.1.1 Reactor Vessel ......................................................................................................................... 2

2.1.2 Steam Generators ..................................................................................................................... 3

2.1.3 Reactor Coolant Pumps ............................................................................................................ 3

2.1.4 Pressurizer ............................................................................................................................... 3

2.1.5 Reactor Coolant Piping ............................................................................................................. 4

2.2 Safety Injection System ............................................................................................................. 4

2.2.1 Safety Injection Pumps ............................................................................................................. 5

2.2.2 Safety Injection Tanks ............................................................................................................... 5

2.2.3 Fluidic Devices .......................................................................................................................... 5

2.3 Containment ............................................................................................................................. 5

2.4 In-Containment Water Storage System ..................................................................................... 6

2.5 Containment Spray System ....................................................................................................... 6

3 METHODOLOGY DESCRIPTION FOR LOCA MASS AND ENERGY RELEASE ..... 19

3.1 Acceptance Criteria ..................................................................................................................19

3.2 Accident Description for LOCA Mass and Energy Release .....................................................19

3.3 Mass and Energy Release Data ................................................................................................20

3.4 Energy Sources ........................................................................................................................21

3.5 Description of Blowdown Model.................................................................................................22

3.6 Description of Core Reflood Model ............................................................................................23

3.7 Description of Post-Reflood Model ............................................................................................24

3.8 Description of Decay Heat Phase Model ....................................................................................24

3.9 Single Active Failure Analysis ...................................................................................................24

3.10 Metal-Water Reaction ..............................................................................................................25

3.11 Evaluation of Effects from the Advanced Design Features ..........................................................25

3.11.1 Effect of Fluidic Device Controlled by K-Factor .........................................................................25

3.11.2 Effect of IRWST Water Temperature on Mass and Energy Release..........................................26

3.12 Description of Containment Pressure and Temperature Analysis ................................................26

4 LOCA M/E ANALYSIS FOR THE APR1400 ......................................................... 32

4.1 M/E Data for the Analyzed Cases .............................................................................................32

4.2 Containment Pressure and Temperature Results .......................................................................32


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5 SUMMARY ......................................................................................................... 60

6 REFERENCES .................................................................................................... 61

APPENDIX A METHODOLOGY DESCRIPTION FOR CONTAINMENT RESPONSE ANALYSIS ............................................................................................. A1

APPENDIX B CALCULATIONS USING APR1400 CONTAINMENT MODEL .................. B1

APPENDIX C CASE STUDIES FOR THE MODELING CHARACTERISTICS................... C1

APPENDIX D CONTAINMENT EXTERNAL PRESSURE ANALYSIS BY THE INADVERTANT OPERATION OF CONTAINMENT SPRAY ..................... D1

APPENDIX E CONCLUSIONS .......................................................................................E1

APPENDIX F REGULATORY REQUIREMENTS FOR CONTAINMENT RESPONSE ANALYSIS .............................................................................................. F1

APPENDIX G REFERENCES ........................................................................................ G1


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LIST OF TABLES

Table 3-1 Containment P/T with 1 Percent Metal-Water Reaction in LOCA ....................................... 27

Table 4-1 Spectrum of Postulated LOCAs ........................................................................................ 33

Table 4-2 Double-Ended Discharge Leg Slot Break - Maximum SIS Flow ......................................... 34

Table 4-3 Initial Conditions for Containment Peak Pressure Analysis................................................ 55

Table 4-4 ESF Systems Parameters for Containment Peak Pressure Analysis ................................. 55

Table 4-5 Primary Side Resistance Factors used in FLOOD3 Code.................................................. 57


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LIST OF FIGURES

Figure 2-1 Reactor Coolant System Block Diagram ........................................................................... 8

Figure 2-2 Reactor Vessel and Internals ............................................................................................ 9

Figure 2-3 Steam Generator.............................................................................................................10

Figure 2-4 Reactor Coolant Pump ....................................................................................................11

Figure 2-5 Pressurizer ......................................................................................................................12

Figure 2-6 Safety Injection System (Short-Term Mode) ....................................................................13

Figure 2-7 Safety Injection System (Long-Term Mode) .....................................................................14

Figure 2-8 Safety Injection Tank and Fluidic Device .........................................................................15

Figure 2-9 In-Containment Water Storage System............................................................................16

Figure 2-10 IRWST and HVT Plan View .............................................................................................17

Figure 2-11 Containment Spray System .............................................................................................18

Figure 3-1 CEFLASH-4A Node Diagram for LOCA Discharge Leg Break ..........................................28

Figure 3-2 FLOOD3 Hydraulic Network Model for LOCA Discharge Leg Break .................................29

Figure 3-3 Containment P/T for DEDL Max.SI Case with 1 Percent Metal-Water Reaction................30

Figure 3-4 Flow Diagram for Fluidic Device Implementation..............................................................31

Figure 4-1 Normalized Decay Heat Curve (0 second ~ EOPR) .........................................................58

Figure 4-2 Double-Ended Discharge Leg Slot Break (DEDLSB) – Max. SI Pump Flow vs. Time .......59


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ACRONYMS AND ABBREVIATIONS

APR1400 Advanced Power Reactor 1400

ASME American Society of Mechanical Engineering

BC boundary condition

CEA control element assembly

CFR Code of Federal Regulations

CFS cavity flooding system CRF carryout rate fraction

CS containment spray

CSHX containment spray heat exchanger

CSS containment spray system

CVTR Carolinas-Virginia Tube Reactor

DBA design basis accident

DC design certification

DCD Design Control Document

DEDLSB double-ended discharge leg slot break

DEHLSB double-ended hot leg slot break

DESLSB double-ended suction leg slot break

DLM diffusion layer model

DVI direct vessel injection

ECCS emergency core cooling system

ECSB emergency containment spray backup

EDG emergency diesel generators

EOB end of blowdown

EOPR end of post-reflood

EQ equipment qualification

ESF engineered safety feature

FD fluidic device

GDC General Design Criterion

HVAC heating, ventilation, and air conditioning

HVT holdup volume tank

ICI in-core instrumentation

IRWST in-containment refueling water storage tank

IWSS in-containment water storage system

KEPCO Korea Electric Power Corporation


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KEPCO E&C KEPCO Engineering & Construction Company

KEPCO NF KEPCO Nuclear Fuel Company

KHNP Korea Hydro and Nuclear Power Company

LBLOCA large break loss-of-coolant accident

LCO limiting condition for operation

LOCA loss-of-coolant accident

M/E mass and energy

MSIV main steam isolation valve

MSLB main steam line break

NAI Numerical Applications

NPSH net positive suction head

NPSHa net positive suction head available

NRC U.S. Nuclear Regulatory Commission

POSRV pilot-operated safety and relief valve

P/T pressure and temperature

PWR pressurized water reactor

PZR pressurizer

RCGVS reactor coolant gas vent system

RCS reactor coolant system

RCP reactor coolant pump

RV reactor vessel

RVI reactor vessel internals

SBLOCA small break loss-of-coolant accident

SC shutdown cooling

SDVS safety depressurization and vent system

SG steam generator

SI safety injection

SIAS safety injection actuation signal

SIP safety injection pump

SIS safety injection system

SIT safety injection tank

SIT-FD safety injection tank with the fluidic device

SMD Sauter Mean Diameter

SS stainless steel

TCD thermal conductivity degradation

TLOFW total loss of feedwater


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TS Trade Secret

UA heat transfer coefficient


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1 INTRODUCTION

This technical report describes the methodology of loss-of-coolant accident (LOCA) mass and energy(M/E) release analysis which is intended to be used for the Nuclear Regulatory Commission (NRC) design certification (DC) application of the Advanced Power Reactor 1400 (APR1400). The purpose of this report is to explain the methodology used and demonstrate its applicability to the APR1400 LOCA mass and energy release analysis.

The APR1400 is an advanced pressurized water reactor (PWR) with improved design features to enhance reliability while retaining many basic features of existing PWRs. Therefore, it is required for the APR1400 process that the results, including the effects of the advanced design features on the mass and energy release analysis and containment pressure/temperature analysis meets the acceptance criteria of Standard Review Plan (SRP) (Reference 1) for those analyses.

The methods as presented are intended for use when determining maximum containment pressure and temperature following a LOCA.

Chapter 2 presents an overview of the APR1400 plant design features related to the LOCA analyses to assist in understanding applicability of the approved methodologies for current PWRs to the APR1400.

Chapter 3 addresses the codes and methodology used for the mass and energy release analysis. In this section, the mass and energy release models for the blowdown, reflood and the post-reflood phases are described. The modifications in the calculation process for reflecting the advanced design features of the APR1400 are described.

Chapter 4 presents a sample mass and energy release calculation result performed by using the analytical models described above and its containment response.

Chapter 5 summarizes the version of the mass and energy release evaluation model with some modifications incorporating the advanced design features, which is based on the approved methodology and describes the conclusions.


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2 DESIGN DESCRIPTION OF THE APR1400

2.1 RCS General Description

The RCS block diagram is shown in Figure 2-1. The major components of the system are a reactor vessel (RV), two parallel heat transfer loops, each containing one SG and two RCPs, and a pressurizer connected to the RV outlet pipe in RCS loop 2. The two SGs and the four RCPs are symmetrically located on opposite sides of the reactor vessel. All components are located inside the containment building.

During normal operation, the reactor coolant is circulated through the reactor vessel and SGs by the RCPs. The reactor coolant is heated as it passes through the reactor vessel by energy produced by the fissioning fuel in the core and is cooled in the SGs as it transfers heat to the secondary system. Feedwater entering the shell side of the SGs absorbs the heat from the primary system forming steam.

The RCS pressure is maintained and controlled through the use of a pressurizer where steam and water are maintained in thermal equilibrium. During full-load operation, the pressurizer volume is almost evenly divided between saturated water and saturated steam. Steam is formed by energizing immersion heaters in the pressurizer, or is condensed by a subcooled pressurizer spray as necessary to maintain operating pressure and limit pressure variations due to plant transients.

2.1.1 Reactor Vessel

The reactor vessel and internals are shown in Figure 2-2. The reactor vessel is essentially a vertical right cylinder with two hemispherical heads. The lower head is welded to the reactor vessel shell and contains [ ]

TSin-core instrumentation penetrations. The upper closure head can be removed to provide access to

the reactor vessel internals. This head is penetrated by [ ]TS

control element drive mechanism nozzles.

The reactor vessel contains the fuel bundles, control rods and other internals necessary for support and flow direction. During normal operation, reactor coolant enters the vessel through four inlet nozzles and flows downward through the RV annulus, then turns and flows upward through the active core. The reactor coolant flows up parallel to the axis of the fuel bundles, removing the heat generated within the fuel as it passes through the core. The reactor coolant then turns and leaves the reactor vessel through two outlet nozzles to the RCS hot legs.

A small core bypass flow progresses from the four inlet cold leg nozzles upward through the downcomer region where a small flow leaks across the alignment keyways between the core support barrel and the reactor vessel. In addition there is a small leakage through the outlet nozzle clearance with the core support barrel. The bypass flow enters and mixes with the fluid in the regions above the inner barrel assembly, passes down through the inner barrel assembly, and mixes with the flow exiting the core. Additional flow paths, which do not provide flow used directly in core heat transfer, are as follows:

1. Through the ICI guide tubes

2. Through the CEA guide tubes

3. Through the core support barrel annulus

Four direct vessel injection (DVI) nozzles are located above the centerline of the cold and hot legs nozzles, but below the upper vessel flange. These nozzles provide safety injection system (SIS) flow to the reactor vessel during abnormal operation and accident mitigation. The SIS fluid is injected directly into the reactor vessel downcomer above the elevation of the cold legs.


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2.1.2 Steam Generators

The steam generator shown in Figure 2-3 is a vertical U-tube heat exchanger that operates with primary coolant on the tube side and secondary coolant on the shell side. Hot reactor coolant from the reactor vessel enters the SG through the inlet nozzle in the primary head. From here it flows through the U-tubes, where it transfers heat to the secondary coolant, to the outlet side of the primary head where the flow splits and leaves through the two outlet nozzles.

Feedwater flow enters into the section of integral economizer through two nozzles in the distribution box and is pre-heated before discharge into the evaporator section. Flow in the evaporator section passes upward through annuli formed by the tubes and baffle plate into the axial flow region. This region is basically a counterflow heat exchanger, with feedwater directed upward outside the tubes and primary flow directed downward inside the tubes. Feedwater then exits the economizer with a condition slightly subcooled and enters the boiling region of the SG. The flow after the boiling region is in the condition of steam-water mixture. The steam-water mixture becomes high-quality steam by passing through the moisture separators and steam dryer. The steam exits the SG through the steam restrictor installed in steam nozzle. The restrictor is sized so that there is a [ ]

TSpercent reduction in flow area enough to

reduce containment peak pressure and temperature and to limit the return to power following a steam line break.

2.1.3 Reactor Coolant Pumps

The four identical RCPs are vertical single-stage, bottom suction, horizontal discharge, motor-driven centrifugal pumps designed to overcome the system flow resistances and circulate the reactor coolant at the flow rate required for design power operation. They are also used to heat up the reactor coolant during plant startup. The reactor coolant pump is shown in Figure 2-4.

Each RCP motor is provided with a flywheel to increase the rotating inertia of the RCP assembly. This inertia increases the RCP coastdown time and reduces the rate of decay of reactor coolant flow if electrical power to the RCP motors is lost, thus providing reasonable assurance that fuel design limits are not exceeded during the incident. An anti-reverse rotation device is provided in each RCP motor to prevent pump windmilling in the reverse direction and to limit backflow through a stopped RCP.

2.1.4 Pressurizer

The pressurizer is shown in Figure 2-5. The pressurizer is designed to maintain RCS operating pressure and compensate for changes in reactor coolant volume during temperature/load changes. The pressurizer is connected to the hot leg in RCS loop 2 via a surge line and to the cold leg in loop 2A and 2B via spray lines. The pressurizer controls the RCS pressure by maintaining the temperature of the pressurizer liquid at the saturation temperature corresponding to the desired system pressure.

Pressurizer pressure is controlled by heaters and spray. The pressurizer heaters are sheath-type immersion heaters that protrude vertically into the pressurizer through sleeves welded in the lower head. The heaters consist of proportional heaters and backup heaters. The proportional heaters, a small portion of the pressurizer heaters, are operated continuously to offset heat losses to ambient and to heat the continuous spray water from the cold leg to be in saturation. The proportional heater output varies between 0 and 100 percent and is automatically controlled by the pressurizer pressure control system (PPCS). The backup heaters, the remaining pressurizer heaters are automatically controlled by the PPCS. The backup heater output is not variable; on-off control is used. The backup heaters are normally deenergized but are turned on by a low pressurizer pressure signal or high level error signal.

The pressurizer heaters and their controls are non-Class 1E using normal power. A sufficient number of backup heaters are capable of being manually aligned to the emergency power supply in order to maintain natural circulation at hot standby conditions when normal power is not available.


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The average reactor coolant temperature is programmed to vary as a function of load. A reduction in load causes a reduction in the average reactor coolant temperature to its programmed value for the lower power level. The resulting contraction of the reactor coolant lowers the pressurizer water level, causing RCS pressure to fall. The pressurizer water level error from the programmed level setpoint which is a function of TAVG is controlled by the operation of the charging control valve and letdown orifice isolation valves.

An auxiliary spray line is provided from the charging pump to permit pressurizer spray during plant cooldown after the RCPs must be shut down due to low system pressure.

A vent line connection to the RCGVS is provided from the counter flange of each POSRV inlet nozzle. The vent line is used to vent air from the pressurizer prior to plant startup. In addition, the vent line allows non-condensable gases to be vented to the RCGVS during post-accident operations when these gases may collect in the pressurizer steam space. The vent connection also provides the capability of using the RCGVS for RCS pressure control during natural circulation cooldowns and after some accidents in which the use of pressurizer main and/or auxiliary spray is not possible.

The safety depressurization and vent system (SDVS) provides the capability to initiate feed-and-bleed operation during the beyond-design-basis event of total loss of feedwater (TLOFW). The SIS provides a "feed" flow to the RCS while the "bleed" flow exits the RCS through the POSRVs to the IRWST. The combined feed-and-bleed operation would result in an adequate core cooling and safe shutdown for the beyond-design-basis event. The SDVS provides the capability to depressurize the RCS in response to a severe accident scenario. The POSRVs are designed and qualified in accordance with requirements associated with the RCS overpressure protection and rapid depressurization functions.

2.1.5 Reactor Coolant Piping

Each loop of the RCS contains five pipe assemblies; one [ ]TS

internal diameter pipe assembly between the reactor vessel outlet nozzle and SG inlet nozzle, two [ ]

TSinternal diameter pipe

assemblies from the SG's two outlet nozzles to the RCP suction nozzles, and two [ ]TS

internal diameter pipe assemblies from the RCP discharge nozzles to the reactor vessel inlet nozzles. These pipe assemblies are referred to as the hot leg, the suction legs, and the pump discharge legs, respectively. The suction leg and pump discharge leg are also referred to as the cold leg. The other major section of reactor coolant piping, the surge line, is a [ ]

TS pipe assembly between the

pressurizer and the hot leg in loop 2.

2.2 Safety Injection System

The SIS is designed to inject borated water into the RCS to flood and cool the core following a LOCA. The SIS provides the feed function and restores the RCS liquid inventory when the RCS pressure decreases rapidly by opening Pilot Operated Safety Relief Valves (POSRVs) during beyond-design-basis event of a total loss of feedwater (TLOFW) to the steam generators.

The SIS consists of four mechanically separated trains, four SITs, and associated valves, piping and instrumentation. Each train contains one SI pump, one SIT, and associated suction and discharge paths. The pumps take suction from the IRWST. Motor-operated valves and a pump in each train receive power from either the normal power source or the emergency diesel generators (EDGs).

Normal power connections for Class 1E are through four independent electrical trains with each train providing power to each bus. In the event of a LOCA, in conjunction with a single failure in the electrical supply, the flow from at least three SI pumps is available for core protection. One independent electrical train, as described above, supplies power to each SI pump and associated valves.


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2.2.1 Safety Injection Pumps

The four SI pumps are horizontal, multi-stage, and centrifugal pumps driven by induction motors. The SI pumps deliver water from the IRWST to the reactor vessel downcomer via DVI nozzles. Each SI pump discharge line is connected to the DVI nozzle or the DVI nozzle/hot leg. The SI lines 1 and 2 inject borated water to the RCS through the DVI nozzles and the SI lines 3 and 4 to the DVI nozzles or hot leg injection lines for the long-term mode. This is schematically illustrated in Figures 2-6 and 2-7.

2.2.2 Safety Injection Tanks

The four safety injection tanks provide a means of rapidly reflooding the core following a large break LOCA, and keeping it covered until flow from the SI pumps becomes available (in the event that offsite power is lost, there will be [ ]

TS for SI pump delay time from reaching the safety injection

setpoint to the time when SI pumps deliver flow to the RCS at full speed). For the purpose of safety analysis, water injected up until the end of RCS blowdown is disregarded and refill/reflood is initiated when the reactor vessel is empty.

Whenever pressurizer pressure is above [ ]TS

, the SITs are isolated from the RCS by only two check valves in series. If RCS pressure should fall below SIT pressure, the tanks will begin to discharge borated water into the RCS. Thus, the tanks are an extremely reliable passive core flooding system.

2.2.3 Fluidic Devices

A passive fluidic device, installed in the SITs, can provide two operation stages of a SI water injection into the RCS and allow more effective use of borated water in case of LOCA. Once LOCA occurs, the system will deliver a high flow rate of cooling water for a certain period of time, and thereafter, the flow rate is reduced.

The fluidic device (FD) consists of vortex chamber, main (supply) port, control port, exit port, and standpipe. The vortex chamber shaped a flat slice of a cylinder and is installed horizontally inside the SIT at the bottom, its axis overlapping with the SIT centerline. The four main ports and four control ports are connected [ ]

TS in a symmetric manner to the vortex chamber with a certain angle. The

standpipe is connected to the main port and its axis is parallel to the SIT centerline.

The time of flow switching is determined by the height of the standpipe. The water flows through both of the main and control ports when the water level is above the top of the standpipe. In this case, flow passage is almost straight and flow resistance is minimal. Conversely, when the water level is below the top of the standpipe, the water flows only through the control ports. The SIT-FD is shown in Figure 2-8.

2.3 Containment

The containment building encloses the entire pressurized water reactor, steam generators, reactor coolant loops, pressurizer, and portions of the auxiliary and engineered safety features (ESF) systems. The containment structure consists of a reinforced concrete base slab, a prestressed concrete cylindrical shell and hemispherical dome, and a reinforced concrete internal structure. The inside face of the slab, shell, and dome of the containment interior boundary is lined with a leak tight carbon steel liner.

The containment structure provides reasonable assurance that leakage of radioactive material to the environment does not exceed the acceptable dose limit as defined in 10 CFR 50.34 even if a LOCA were to occur. Also, the containment must withstand the pressure and temperatures of the design basis accident (DBA) without exceeding the design leakage rate of [ ]TS volume

for the first 24 hours. The limit

thereafter is based on a leak rate associated with half of the design leak rate.


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The containment design pressure is based on the worst-case LOCA that bounds all of the secondary system piping rupture events for the peak containment pressure. A design margin of a minimum of 10 percent is taken into account for determining the containment design pressure. In addition, the containment heat removal system is designed with sufficient heat removal capability to reduce and maintain the containment pressure at less than 50 percent of peak calculated pressure within 24 hours after the postulated accident.

2.4 In-Containment Water Storage System

The in-containment water storage system (IWSS) performs water collection, delivery, storage and heat sink functions inside containment during normal operation and accident conditions. It consists of the in-containment refueling water storage tank (IRWST), the holdup volume tank (HVT), and the cavity flooding system (CFS). The IRWST and HVT are integral parts of the reactor containment building internal structures. Each is a reinforced concrete structure, and the inside surface is lined with stainless steel. The IWSS is shown in Figure 2-9. A plan view for IRWST and HVT is shown in Figure 2-10.

The IRWST provides a water source to fill the refueling pool during a refueling outage. Also, the IRWST provides a continuous suction source for the SI, containment spray (CS) and shutdown cooling (SC) pumps, thus eliminating operator actions to realign the suction in an emergency such as a LOCA. This improves reliability and reduces the complexity of the design compared to conventional systems that require suction switchover to the sump after the water in the supply tank is depleted. The HVT provides a low collection point in containment to collect water released from pipe breaks and containment sprays.

The IRWST is designed to have a sufficient inventory of boric acid water for refueling and long-term core cooling during a LOCA. The capacity of the IRWST water is in the range of [ ]

TS and [ ]

TS during the normal operating condition. The minimum IRWST water volume that

considers the water trapped in containment during an accident is [ ]TS

and this minimum water volume is chosen as the initial value for containment response analyses to LOCA and MSLB accidents.

The reactor coolant discharged through a break and the water from the containment spray system during the LOCA is collected in the HVT. The accumulated water in the HVT overflows back into the IRWST, to maintain a minimum water level. The temperature in the IRWST during normal operation is in the range of [ ]

TS. The design temperature of the IRWST water is [ ]

TS.

2.5 Containment Spray System

The CS system is an engineered safety feature (ESF) designed to reduce the containment atmosphere pressure and temperature below containment design limits with a margin in the event of a postulated LOCA or MSLB inside containment, by removing heat and steam from the containment atmosphere. The function of the CS system is accomplished by taking borated water from the IRWST, pumping it to the containment atmosphere through a CS heat exchanger and spray nozzles. The schematic diagram of the CS system is shown in Figure 2-11.

The CS system reduces the containment pressure and temperature and removes radioactive fission products from the containment atmosphere following a postulated LOCA or MSLB. The CS system consists of two redundant 100 percent capacity trains. Each train includes a CS pump, a CS heat exchanger, a CS miniflow heat exchanger, a main spray header with nozzles, an auxiliary spray header with nozzles, and associated valves, piping and instrumentation.

The CS system has sufficient capacity to reduce containment pressure to less than 50 percent of calculated peak pressure within 24 hours after the design basis event and to maintain the bulk contents of


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the IRWST subcooled during the feed-and-bleed mode of core cooling. Each CS pump train is designed to deliver a minimum [ ]

TS to the spray nozzles.

The CS system operation is automatically actuated by a containment high-high pressure signal (containment spray actuation signal). The actuation pressure is sufficiently high to prevent the system actuation due to a small leak. In APR1400 containment integrity analyses, the spray actuation setpoint is assumed at [ ]

TS of the containment pressure with [ ]

TS of the CS

actuation delay time for LOCA or MSLB accidents, respectively.


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Figure 2-1 Reactor Coolant System Block Diagram

Stea

mG

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No.

1

Rea

ctor

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sel

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utdo

wn

Coo

ling

LOO

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LOO

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Mai

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eam

RC

P 2A

RC

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Letd

own

(CV

CS)

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P 1A

RC

P 1B

RC

P 1A

Mai

nSt

eam

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RC

GV

S

Surg

e Li

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ne(C

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4)

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ctor

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lant

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/Sh

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RD

T


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Figure 2-2 Reactor Vessel and Internals

TS


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Figure 2-3 Steam Generator

TS


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Figure 2-4 Reactor Coolant Pump

TS


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Figure 2-5 Pressurizer

TS


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Figure 2-6 Safety Injection System (Short-Term Mode)

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Figure 2-7 Safety Injection System (Long-Term Mode)

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Figure 2-8 Safety Injection Tank and Fluidic Device

TS


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Figure 2-9 In-Containment Water Storage System

TS


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Figure 2-10 IRWST and HVT Plan View

TS


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3 METHODOLOGY DESCRIPTION FOR LOCA MASS AND ENERGY RELEASE

3.1 Acceptance Criteria

The methodology of the mass and energy release analysis incorporates the acceptance criteria described in Standard Review Plan (SRP) Section 6.2.1.3, “Mass and Energy Release Analysis for Postulated Loss-of-Coolant-Accident (LOCA)” (Reference 1). The SRP acceptance criteria are based on meeting the relevant requirements of the following regulations:

• General Design Criterion 50, as it relates to the containment and subcompartments being designed with sufficient margin, requires that the containment and its associated systems can accommodate, without exceeding the design leakage rate, and the containment and subcompartment design can withstand the calculated pressure and temperature conditions resulting from any LOCA.

• 10 CFR Part 50, Appendix K, as it relates to sources of energy during the LOCA, includes requirements to provide assurance that all the energy sources have been considered.

Conservative assumptions for the input conditions based on the acceptance criteria are considered in the mass and energy release analysis. However, any specific limitation to the results based on the acceptance criteria is not applicable to the mass and energy release, since the mass and energy release results are the process parameters at the intermediate stage of the containment analysis.

The final results of the containment analysis show the behaviors of the containment pressure and temperature. The limitation based on the acceptance criteria is applied to the calculated containment pressure as the containment design pressure of [ ]

TS. The containment design

pressure is calculated in accordance with the guideline of Reference 5.

3.2 Accident Description for LOCA Mass and Energy Release

LOCA mass and energy release analyses are categorized as the following phases: blowdown, refill, reflood, post-reflood, and decay heat period.

a. The blowdown period extends from time zero until the primary system depressurizes and equalizes with the containment pressure. During blowdown, most of the initial primary coolant is released to the containment as a two-phase mixture. Following blowdown, the water for releases is supplied from the safety injection system (SIS). There is an important distinction between hot leg breaks and cold leg breaks for LOCA post-blowdown analyses. For a hot leg break, most of the SIS-supplied water leaving the core can vent directly to the containment without passing through a steam generator. Therefore, there is no mechanism for releasing the steam generator energy to the containment for a hot leg break, and only the blowdown period is considered. In contrast, for cold leg breaks like discharge leg break and suction leg break, the water passes through a steam generator before reaching the containment. Therefore, the post-blowdown releases to the containment are considered for cold leg breaks.

b. The first post-blowdown period is refill. During refill, the SIS water refills the bottom of the reactor vessel to the bottom of the core. This period is conservatively omitted from the analysis.


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c. The second post-blowdown period is the reflood period. During reflood, SIS water floods the core. Reflood is assumed to end when the liquid level in the core is [ ]

TS below the top of

the active core. During reflood, a significant amount of the SIS water entering the core is postulated to be carried out of the core by the steaming action of the core-to-coolant heat transfer process. This fluid then passes through a steam generator where reverse (i.e., secondary to primary) heat transfer heats it before it reaches the containment. The residual steam generator secondary energy is sufficient to convert all of this fluid to superheated steam during the initial part of the reflood period. Subsequently, as the steam generators are cooled by this process, there is not enough heat transfer to boil all of the fluid passing through the tubes. This causes the break flow to change from pure steam to two-phase. As the entire NSSS cools, the flow to the containment eventually becomes subcooled because the safety injection water is subcooled. The onset of the two-phase release to the containment may or may not occur before the end of reflood; typically, it occurs close to the end of reflood. The potential release of subcooled fluid to the containment does not occur during reflood when conservative system parameters are used.

d. The third post-blowdown period is the post-reflood period. During this period, the dominant process is the continued cooling of the steam generators by the SIS water leaving the core. The release to the containment during this period becomes generally two-phase in the earlier stage of this period as the cooling of the steam generators continues. The post-reflood ends when the affected steam generator has essentially reached the containment temperature.

e. The final post-blowdown period is the decay heat period, which begins at the end of post-reflood. During the decay heat period, the dominant mechanisms for release rates are the generation of the decay heat and the cooling of all NSSS metal. The decay heat period ends when the containment pressure and the environmental pressure are essentially equal.

LOCA mass and energy releases are analyzed using the computer codes, CEFLASH-4A and FLOOD3 for the categorized phases. The CEFLASH-4A computer code is used for analysis of the blowdown period and the FLOOD3 computer code is used for analysis of the reflood period. The detailed descriptions of the codes are presented in References 2 and 3, respectively.

The M/E calculated by CEFLASH-4A and FLOOD3 is supplied as input to the GOTHIC computer program (References G-5, G-6 and G-7 in Appendix G) for the containment analysis. Mass and energy release during the decay heat period is calculated directly by GOTHIC and is integrated with the containment analysis. Detailed descriptions of the codes are presented in Appendix A.

3.3 Mass and Energy Release Data

Pipe breaks and locations are assumed to be as follows:

• Double-ended suction leg slot break (DESLSB) in the RCP suction leg.

• Double-ended discharge leg slot break (DEDLSB) in the RCP discharge leg.

• Double-ended hot leg slot break (DEHLSB) in the RCP hot leg.

The break type is assumed to be a slot break that has the break area equivalent to the double-ended break. The largest break area, i.e. the double-ended break area is limiting for a large break LOCA.


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There are five analysis cases for LOCA M/E analysis: for the suction leg break with maximum and minimum SI flow, discharge leg break with maximum and minimum SI flow, and hot leg break cases. Those cases are summarized in Table 4-1. In this report, the analysis result of the limiting case is provided. The limiting case is determined based on the containment peak pressure. Mass and energy release data for the limiting case are provided in Tables 4-2. For cold leg breaks (pump suction and discharge), some of the post-blowdown SIS water is postulated to spill to the containment floor whenever the reactor vessel annulus is full. The vessel spillage data associated with these breaks are also given in Part A of Table 4-2.

3.4 Energy Sources

The following sources of generated and stored energy in the reactor coolant system and secondary coolant system are considered:

• Primary coolant

• Secondary coolant

• Primary walls (including reactor internals)

• Secondary walls

• Safety injection water

• Core power

• Decay heat

For the conservative analysis, the assumptions of energy sources are biased to maximize stored energy.

For considering the stored energy in the coolant, the initial reactor coolant system water volumes are conservatively calculated based on maximum manufacturing tolerances for the reactor vessel and steam generator tubes. Expansion of the loop components from cold to hot operating conditions is also considered for the coolant stored energy. The initial water volume of the pressurizer includes an allowance for level instrumentation error. This maximizes the pressurizer water volume containing the maximized stored energy.

For considering the stored energy in the walls, the large specific heat and heat conductivity of carbon steel are conservatively assumed for all walls in the RCS.

The core stored energy may be increased slightly by thermal conductivity degradation (TCD). However, the effect of TCD on the mass and energy release is negligible. The details of the results are described in Reference 5.

For considering the energy in the safety injection water, the liquid break flow is assumed to be mixed with the water in the IRWST. The mixed water is taken and discharged into DVI by SI pumps. This increases the energy in the safety injection water.

The initial power level assumed in the analyses consists of the core power and an additional RCP power. The initial core power is assumed to be 102 percent of its normal power which includes the instrumentation error. The higher power level is conservative for LOCA containment pressure calculations.


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For the core decay heat curve as a fraction of the initial power level following the accident; a [ ]TS

conservatism factor is used for the first [ ]

TS, followed by a [ ]

TS factor thereafter. The

normalized decay heat curve is shown in Figure 4-1.

Initial conditions in the reactor coolant system are given in Table 4-3. It shows various stored energies in the RCS at the initial time.

3.5 Description of Blowdown Model

Blowdown mass and energy release rates are calculated using the CEFLASH-4A computer code (Reference 2). The node diagram of this code is presented in Figure 3-1 for the blowdown analysis of LOCA discharge leg break. A description of the CEFLASH-4A code including the conservatisms in modeling is given below. This section includes justification of the heat transfer correlations. The following assumptions are made in selecting input data for the code.

a. The CEFLASH-4A code model of the heat transfer in a node allows only one wall per node. Accordingly, the thickness used for the "U" factor for each node wall is selected so that the energy released from the system is conservatively modeled.

b. The CEFLASH-4A wall representation uses the total heat capacitance of all the walls in the reactor coolant system that actually face a given node. This is conservative since, in reality, some of the walls will not participate as effectively as others in the heat transfer process. For example, the geometry of the actual flow path allows some components to partially shield others from the flow. This effect is conservatively omitted from the modeling.

c. Although much of the steel facing the coolant in the reactor coolant system is stainless cladding [ ]

TS, a conservative carbon steel conductivity

[ ]TS

, is used for the entire wall. This conservatively overpredicts the energy released from all such walls.

d. Wall surface heat transfer coefficients are assumed to be infinite.

e. All primary water volumes are conservatively increased from their nominal design values in order to obtain an upper bound for the available mass and energy in the system prior to LOCA. The pressurizer water volume includes an allowance for level instrumentation error. Pressure and temperature expansion of the reactor coolant system and steam generator to the normal operating condition is included.

f. An accepted (Jens Lottes) two-phase heat transfer correlation is used for the core to coolant heat transfer whenever the flow through the core is not pure steam.

g. Heat transfer across the steam generator tubes is modeled with the same heat transfer coefficient in both the forward and reverse directions. This is conservative since it maintains a nucleate boiling heat transfer coefficient on the secondary side during the LOCA blowdown. In reality, the reactor trip following the LOCA would result in a turbine trip that would close the turbine stop valves and then the heat transfer in the secondary side would be through natural convection, in which the heat transfer coefficient has a small value. However, in this analysis, it is conservative to assume the overall heat transfer coefficient in the initial steady state at full-power operation since it maximizes the reverse heat transfer.

h. The turbine stop valves are assumed to close at [ ]TS

. This is conservative since it keeps energy within the NSSS which in turn is a source of energy for containment pressurization.


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i. The main feedwater isolation valves(MFIVs) are assumed to close only after the generation of a main steam isolation signal (MSIS) of [ ]

TS containment pressure. This

signal occurs very rapidly (~1 second). MFIV closure is assumed to take [ ]

TS, with an additional allowance of [ ]

TS for the MSIS signal delay.

Feedwater flow and enthalpy are kept at their normal values due to the short times involved. As an additional conservatism, the feedwater is assumed to be added at the end of blowdown, so that the steam generator secondary temperatures during the blowdown are not lowered by the relatively cold feedwater. Note that the feedwater is hot relative to potential peak LOCA containment temperatures so that feedwater addition at the end of blowdown is conservative both for blowdown and for reflood calculations.

j. Auxiliary feedwater flow is conservatively omitted since it is cold [ ]

TS relative to both blowdown and reflood conditions.

3.6 Description of Core Reflood Model

Reflood mass and energy release rates are calculated using the FLOOD3 computer code (Reference 3). The hydraulic network of this code is presented in Figure 3-2 for the reflood analysis of LOCA discharge leg break. Heat transfer is conservatively modeled for core, vessel walls, vessel internals, loop metal, steam generator tubes, steam generator secondaries, and steam generator secondary walls. The FLOOD3 code hydraulics calculates flow rates and pressure. The heat transfer process predicts fluid enthalpies. Fluid densities are calculated as functions of pressures and enthalpies. The conservatisms in the model are as follows:

a. The containment backpressure during reflood is assumed to be [ ]TS

and constant. It is given for the input of FLOOD3 code.

b. A one-dimensional heat transfer model is used for all wall heat transfer calculations. This is demonstrated in Reference 4 where comparisons of one-dimensional models and otherwise identical two-dimensional models show that one-dimensional modeling is more conservative.

c. A nucleate boiling heat transfer coefficient of [ ]TS

is used to model the heat transfer from the steam generator tubes to the primary coolant. This coefficient represents an upper limit, and is conservatively used at all times throughout the tubes.

d. During reflood, the behavior of steam generator liquid level is calculated. The liquid level is predicted to be decreased due to the reversed heat transfer and a part of tube area is in contact with the secondary steam. The heat transfer coefficient of steam-to-tube area is the Nusselt condensation heat transfer coefficient, and is much higher than that of liquid-to-tube, natural circulation heat transfer coefficient. Therefore, it is conservatively assumed that the whole tube heat transfer area is in contact with the secondary steam. A conservative Nusselt condensation heat transfer coefficient of [ ]

TS is used

in conjunction with the tube area.

e. The thermal resistance corresponding to the steam generator tubes is [ ]

TS. This value is also used in calculating secondary to primary

heat transfer.

f. The carryout rate fraction (CRF) used during reflood is constant (0.05) up to the 46 cm (18 in) core level, and linearly increases to 0.8 up to the 61 cm (24 in) core level, and is kept constant at 0.8 until the 3.2 m (10.5 ft) level is reached, which is 0.6096 m (2 ft) below the top of the


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active core. Other variables, such as core inlet temperature, pressure, flow rate, linear heat rate, or other experimental data are not used to determine the CRF.

g. Reflood is assumed to terminate when the 3.2 m (10.5 ft) quench level in the core is reached.

h. [ ]TS

of the standard decay heat (Figure 4-1) curve is used as a conservatism for the available energy sources.

i. During reflood, credit is taken for the condensation of steam in the annulus by the cold SIS water. As a conservatism, credit is not taken unless the reactor vessel annulus is full since the SI flow is injected directly into the annulus. Also, as an additional conservatism, credit is not taken when the SI flow rate is too low to thermodynamically condense all of the steam in the annulus. Thus, credit is not taken for the condensation after the SITs empty or the turndown to low SIT flow by the fluidic device. The percentage of the total steam flow condensed varies slightly with time for each case. For suction leg and discharge leg cases, credit is taken for the condensation of approximately [ ]

TS of the total steam flow when the annulus is full and the

thermodynamic criteria are simultaneously met.

3.7 Description of Post-Reflood Model

The post-reflood model is identical to the reflood model except that, at the end of reflood, the CRF is changed from [ ]

TS to [ ]

TS. This conservatively increases the system flow rates due to the increased

CRF. The flow rates are further enhanced by the fact that the core liquid height is now constrained at the [ ]

TS level, which maximizes the available driving head between the annulus level and the

core in the flooding equation. All heat transfer coefficients are kept at the values used for the reflood analysis. Condensation is analyzed as previously described; however, there is insufficient spillage for complete thermodynamic condensation of the steam so that credit for condensation is not taken.

3.8 Description of Decay Heat Phase Model

The final phase of the large break LOCA is a relatively stable period characterized by decay heat release and it extends from the end of the post-reflood phase. The analysis method used to determine the mass and energy released during this period is described in Appendix A.2.4, “Decay Heat Phase M/E Analysis Model.”

3.9 Single Active Failure Analysis

Two possible failures are considered as single failure in LOCA mass and energy analysis, the failure of one SI pump and the failure of one emergency diesel generator (EDG). Both failures would degrade SI flow and eventually degrade emergency core cooling system (ECCS) performance to cool down the core. In LOCA mass and energy analysis, the single failure is assumed for minimum SI flow and no failure is assumed for maximum SI flow.

Another failure in the containment system is considered as a single failure, the failure of one train of containment spray. The failure reduces the capability to suppress containment pressure, which results in higher containment pressure during the LOCA transient. In the case with maximum SI, the failure of one train of containment spray system (CSS) is assumed. In the case with minimum SI, the failure of one train of containment spray system is assumed also, due to the failure of one emergency diesel generator.

The limiting case is determined by the case analyses with the maximum and minimum safety injection flow. This case analysis with the maximum and minimum safety injection flow is performed at the three break locations.


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The cases presented in this analysis show the mass/energy source terms with maximum safety injection (no pump failure or no power source failure) and minimum safety injection (failure of one diesel generator). Since the peak containment pressure is a function of both the release rates and the containment parameters, the effects of the various single failures are estimated to find out the limiting single failure for the analysis.

3.10 Metal-Water Reaction

According to 10 CFR 50, Appendix K, the additional source energy by the metal-water reaction should be considered in LOCA mass and energy analysis. A bounding calculation, not dependent on the Appendix K metal-water correlation, assumes the maximum allowable one percent zirconium-water reaction. This bounding calculation is based on a zirconium mass in the active core of [ ]

TS Using

a molecular weight of 91.22 for zirconium and reaction energy of [ ]

TS the 1 percent metal-water reaction produces [ ]

TS

The metal-water reaction energy is added to the energy release rate and assumed to be uniformly distributed over a period no longer than 2 minutes following the end of blowdown (Reference 5). If the peak pressure time is earlier than the 2 minutes after the end of blowdown (EOB), the peak pressure time is conservatively assumed to be the end of the metal-water reaction.

The results of containment pressure and temperature (P/T) analysis with 1 percent metal-water reaction are provided in Figure 3-3 and Table 3-1. Figure 3-3 shows the results of containment P/T analysis for the limiting case (DEDLSB with maximum SI) and provides comparison with containment P/T results of the APR1400 DCD (Figure 6.2.1-3). Table 3-1 shows the results of containment P/T analysis for all cases of LOCA and comparison with those (Table 6.2.1-19) of the APR1400 DCD.

The metal-water reaction energy is less than [ ] TS

of the total amount of energy released to the time of peak containment pressure during the limiting LOCA. The total energy released to the peak pressure time at [ ]

TS based on Table 4-2 (sh. 21 of 21) is [ ]

TS This metal-

water reaction energy causes an increase of [ ]TS

of the containment peak pressure. Even if this bounding metal-water reaction is assumed to occur, the metal-water reaction energy will have little effect on the containment pressures. Detailed quantitative descriptions of the metal-water reaction analysis are provided in Reference 3.

Actually, the maximum fuel clad temperature is only [ ]TS

during the limiting LOCA transient. This fuel clad temperature is not so high enough to enable the metal-water reaction to occur in this analysis. Therefore the metal-water reaction energy is not included in the mass/energy source terms of the APR1400 DCD.

3.11 Evaluation of Effects from the Advanced Design Features

3.11.1 Effect of Fluidic Device Controlled by K-Factor

In the APR1400 design, a fluidic device (FD) is employed for control of safety injection tank (SIT) flow during a large break LOCA. The fluidic device is a safety-related component and is installed in the SIT as shown in Figure 2-8. It passively controls the SIT injection flow in two operation stages, high-flow injection initially, and then low-flow injection, according to the SIT water level.

High flow : This stageoccurs when the SIT water level is above the entrance of stand pipe of the fluidic device and SIT inventory is supplied into both the main port and the four control ports. The condensing fraction is assumed to be [ ]

TS during the period of high-flow injection from the time that the RV

annulus is full until the SIT water level decreases below the top of the stand pipe of FD.


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Low flow : Due to the sustained SIT injection, the SIT water level decreases below the entrance of the stand pipe. The inventory supply into the main port is lost. The inventory is supplied only into the four control ports, thus, the SIT fluidic device produces low-flow, which is about one third of the high-flow. This delays the SIT empty time and minimizes the spillage of the injected flow. The condensing fraction is assumed to be [ ]

TS during the period of low-flow injection.

The fluidic device is considered in this analysis. The function of flow control is implemented in the computer code, FLOOD3 as in the flow diagram, Figure 3-4.

Although the fluidic device will improve the LOCA thermal margin for fuel performance, it may have an adverse impact on the mass and energy release during a LOCA. In a conventional LOCA M/E analysis where the fluidic device is not considered, the SIT injection flow is high enough to condense the steam flows in the intact side from the time that the reactor downcomer is full until the SIT is empty. In the APR1400, the SIT-FD low-flow is so small that the flow is not credited to condense the steam flows in reactor vessel annulus. Thus the steam mass and energy release through the break is increased by the amount of the non-condensed steam. The increased mass and energy release can have a considerable impact on the peak pressure and temperature of the containment.

3.11.2 Effect of IRWST Water Temperature on Mass and Energy Release

The released mass through the break during a LOCA is accumulated on the floor inside containment in a hot liquid condition. This liquid flows into the IRWST via the holdup tank and is mixed with the IRWST water inventory, which is the source water for safety injection pumps. Mixing with the hot liquid results in a temperature increase in the IRWST inventory and this yields a temperature increase in the SI water into the RCS. Therefore, the energy of the break flow may be higher due to the deteriorating circulation of the released mass.

In the LOCA M/E analysis, the effect of the IRWST water temperature was considered. The containment analysis based on the initial M/E data can provide the data of the sump water temperature. Taking the data of the sump water temperature as the input of SI water temperature during the reflood stage, reanalysis of LOCA M/E is performed for more conservative M/E data.

3.12 Description of Containment Pressure and Temperature Analysis

The methodology for the containment response analyses to LOCA and MSLB accidents is addressed in detail in Appendix A.


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Table 3-1 Containment P/T with 1 Percent Metal-Water Reaction in LOCA

Maximum Pressure

Accident W/O Metal-Water Reaction W/ Metal-Water Reaction Peak Pr.

kg/cm2A (psia) Time

(seconds) Peak Pr.

kg/cm2A (psia) Time

(seconds) DESLSB with Max. SI DESLSB with Min. SI DEDLSB with Max. SI DEDLSB with Min. SI

DEHLSB

Maximum Temperature Accident

W/O Metal-Water Reaction W/ Metal-Water Reaction Peak T. oC (oF)

Time (seconds)

Peak T. oC (oF)

Time (seconds)

DESLSB with Max. SI DESLSB with Min. SI DEDLSB with Max. SI DEDLSB with Min. SI

DEHLSB

TS


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Figure 3-1 CEFLASH-4A Node Diagram for LOCA Discharge Leg Break

TS


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Figure 3-2 FLOOD3 Hydraulic Network Model for LOCA Discharge Leg Break

TS


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Figure 3-3 Containment P/T for DEDL Max.SI Case with 1 Percent Metal-Water Reaction

TS


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Figure 3-4 Flow Diagram for Fluidic Device Implementation

TS


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4 LOCA M/E ANALYSIS FOR THE APR1400

4.1 M/E Data for the Analyzed Cases

Five cases are analyzed for full-spectrum analysis of LOCA M/E: suction leg break with maximum and minimum SI flow, discharge leg break with maximum and minimum SI flow and hot leg break cases. These cases are summarized in Table 4-1. In this report, the analysis result of the limiting case is provided. The limiting case is determined based on the containment peak pressure, which is the case of discharge leg break with maximum SI flow.

The mass and energy release data of the limiting case are given in Part A of Table 4-2. For cold leg breaks (pump suction and discharge), some of the post-blowdown SIS water is postulated to spill to the containment floor whenever the reactor vessel annulus is full. The vessel spillage data associated with these breaks are also given in Part A of Table 4-2. In Part B and Part C of Table 4-2, the transient data of reactor vessel pressure and SI flow are provided, respectively. In Part D of Table 4-2, the chronology of events is provided.

Initial conditions in the reactor coolant system are given in Table 4-3. The ESF parameters and the analysis inputs conservatively assumed for analysis are presented in Table 4-4. The primary side resistance factors for the FLOOD3 code are shown in Table 4-5. The FLOOD3 code hydraulics calculates flow rates and pressure.

Figure 4-1 shows the normalized decay heat curve as a fraction of the initial power level.

Curves of Safety Injection Flow versus time are provided in Figure 4-2. It shows the controlled SIT flow by fluidic device in SIT as described in Subsection 3.11.1. There are two modes of the FD-controlled flow, roughly distinguished as follows: full flow during the blowdown stage and low-flow during the post-blowdown stage. The low-flow controlled by the fluidic device allows sustained SI flow from the SITs for an extended duration.

4.2 Containment Pressure and Temperature Results

The calculation results of the containment pressure and temperature response to LOCA events are described in detail in Appendix B.


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Table 4-1 Spectrum of Postulated LOCAs

Break Location SIS Condition Break Area,

m2 (ft2)

Double-Ended Suction Leg Slot (DESLS)

Maximum SIS Flow

Minimum SIS Flow

Double-Ended Discharge Leg Slot (DEDLS)

Maximum SIS Flow

Minimum SIS Flow

Double-Ended Hot Leg Slot (DEHLS)

N/A

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Table 4-2 Double-Ended Discharge Leg Slot Break - Maximum SIS Flow (1 of 21)

Part A. Mass and Energy Release Data (Blowdown Period)

Time (sec)

Break Mass Flow Rate Break Enthalpy

kg/sec lbm/sec kcal/kg Btu/lbm

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Time (sec)



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Time (sec)



Integral Mass and Energy Release at End of Blowdown

Time (sec)

Integral Mass Integral Energy

kg lbm Million kcal Million Btu

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TS


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Part A: Mass and Energy Release Data (Reflood and Post-Reflood Period)

Time (sec)



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Time (sec)



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Time (sec)



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Integral Mass and Energy Release at the End of Reflood and Post-Reflood

Time (sec)



Part A : Mass/Energy Release Data (Spillage)

Time (sec)



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Part A: Mass and Energy (Steam) Release Data (Decay Heat Period)

Time (sec)

Mass Flow Rate Break Enthalpy


Integral Mass and Energy Release (Vapor) at 24 hours and End of Analysis

Time (sec)



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Part A: Mass and Energy (Spillage) Release Data (Decay Heat Period)

Time (sec)

Mass Flow Rate Break Enthalpy


Integral Mass and Energy Release (Spillage) at 24 hours and End of Analysis

Time (sec)



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Part B : Reactor Vessel Pressure vs. Time (Blowdown Period)

Time (sec)

Reactor Vessel Pressure

kg/cm2A psia

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Time (sec)


kg/cm2A psia

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Time (sec)


kg/cm2A psia

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Part B : Reactor Vessel Pressure vs. Time (Reflood and Post-Reflood Period)

Time (sec)


kg/cm2A psia

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Part B : Reactor Vessel Pressure vs. Time (Reflood and Post-Reflood Period)

Time (sec)


kg/cm2A psia

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Part B : Reactor Vessel Pressure vs. Time (Reflood and Post-Reflood Period) Time (sec)


kg/cm2A psia

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Part B : Reactor Vessel Pressure vs. Time (Decay Heat Period)

Time (sec)


kg/cm2A psia

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Part C: Safety Injection Flow vs. Time (Blowdown Period)

Time (sec)

Safety Injection Tank Flow

kg/sec lbm/sec

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Part C: Safety Injection Flow vs. Time (Reflood and Post-Reflood Period)

Time (sec)

Safety Injection Tank Flow Safety Injection Pump Flow

kg/sec lbm/sec kg/sec lbm/sec

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Time (sec)



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Time (sec)



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Part D : Chronology of Events Time (sec) Event Values

Break occurs

Containment pressure Hi-Hi setpoint

Start safety injection tank (SIT) injection

First peak containment pressure (Blowdown phase)

End of blowdown

Start SI pump injection

SIT flow is turned down to low-flow by fluidic device in SIT

Start containment spray actuation

End of reflood

Peak containment temperature

Peak containment pressure

End of post-reflood

Safety injection tank empty

Time of depressurization of the containment at 50 %

of peak pressure

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Table 4-3 Initial Conditions for Containment Peak Pressure Analysis

(Based on a nominal core power of 3983 MWt)

---------------------------------------------------------------------------------------------------------------------- 1) At full power plus 2% uncertainty plus max. RCP power [ ]

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2) Energy is relative to 0 °C (32°F)

Parameter Value Reactor Coolant System

- Reactor power level 1), MWt

- Average coolant temperature, °C (°F)

- Mass of reactor coolant system liquid, kg (lbm)

- Mass of reactor coolant system steam, kg (lbm)

- Energy in Reactor coolant system liquid plus steam 2), 106 kcal (106 Btu)

- Energy from feedwater nozzle to MSIV per Steam Generator 2), 106 kcal (106 Btu)

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Table 4-4 ESF Systems Parameters for Containment Peak Pressure Analysis

System/Item Full Capacity Value Used for Peak Pressure Analyses 1)

Passive Safety Injection System Number of accumulators (safety

injection tanks)

Pressure setpoint, kg/cm2G (psig) Volume per accumulator, Maximum, m3 (ft3) Minimum, m3 (ft3) - Active Safety Injection System Number of divisions Number of pumps/division Flow rate per pump at 0 psig, Maximum, L/min (gpm) Minimum, L/min (gpm)

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Table 4-5 Primary Side Resistance Factors used in FLOOD3 Code

Path Resistance Factor, R′ SI unit1) (British unit2))

Core

Lower Core

Upper Core

Upper Plenum to Steam Generator, Broken Side

Upper Plenum to Tubes

Tubes to Steam Generator Outlet

Steam Generator Outlet in Broken Side of Annulus

Forward Flow

Reverse Flow

Annulus to Break

Suction Leg Break

Discharge Leg

Steam Generator Outlet in Broken Side to Break

Suction Leg Break

Discharge Leg Break

Upper Plenum to Annulus, Intact Side

Upper Plenum to Tubes

Tubes to Annulus

Break Resistances

1.0 Break

-------------------------------------------

1) Units of R’ are 10 x

kg

m sec

kg

cm

kg

6

3

2

2

2

2) Units of R’ are

10 x

sec

ft sec

lbm

psi 6

3

2

2

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Figure 4-1 Normalized Decay Heat Curve (0 second ~ EOPR)

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Figure 4-2 Double-Ended Discharge Leg Slot Break (DEDLSB) – Max. SI Pump Flow vs. Time

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5 SUMMARY

The purpose of this report is to explain the methodology used and demonstrate its applicability to the APR1400 LOCA mass and energy release analysis.

The analysis includes evaluation of the advanced design features, four EDGs for four trains of ECCS, the fluidic device installed inside the SIT, and the IRWST.

The limiting case of the mass and energy release calculation was the case of double-ended discharge leg slot break with maximum SI pump flow.

The analysis shows that the results, including the effects of the advanced design features meets the acceptance criteria of Standard Review Plan.

The methodology as presented will be used when determining maximum containment pressure and temperature following a loss-of-coolant accident for the APR1400.


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6 REFERENCES

1. NUREG-0800, Standard Review Plan, Section 6.2.1.3, Mass and Energy Release Analysis for Postulated Loss-of-Coolant Accidents, Rev. 3, U.S. Nuclear Regulatory Commission, March 2007.

2. CENPD-133P, "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis", Combustion Engineering Inc., August 1974 (Proprietary).

CENPD-133P, Supplement 2, "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis (Modifications)," Combustion Engineering Inc., February 1975 (Proprietary).

CENPD-133, Supplement 4-P, "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis," Combustion Engineering Inc., April 1977 (Proprietary).

CENP-133, Supplement 5-P, "CEFLASH-4A, A FOKTRAN 77 Digital Computer Program for Reactor Blowdown Analysis," Combustion Engineering Inc., June 1985 (Proprietary).

3. "FLOOD-MOD2 - A Code to Determine the Core Reflood Rate for a PWR Plant with Two Core Vessel Outlet Legs and Four Core Vessel Inlet Legs", Interim Report, Aerojet Nuclear Company, November 2, 1972.

4. Kreith, F. "Principles of Heat Transfer," International Textbook Company, 1958.

5. ANSI/ANS-56.4-1983, “Pressure and Temperature Transient Analysis for Light Water Reactor Containments,” American Nuclear Society, December 1983


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LOCA Mass and Energy Release Methodology APR1400-Z-A-NR-14007-NP, Rev. 0

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LOCA Mass and Energy Release Methodology

Appendices A through G

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ABSTRACT

Appendices A through G to the technical report “LOCA [Loss-of-Coolant Accident] Mass and Energy Release Methodology” for the Advanced Power Reactor 1400 (APR1400) describe the methodology used to predict the containment maximum pressure and temperature response to a spectrum of high-energy line breaks in the large, dry APR1400 containment building. The containment response calculations use the GOTHIC computer program with plant-specific inputs.

The plant-specific short-term (from the accident initiation to the end of post-reflood [EOPR]) mass and energy (M/E) release data required for the GOTHIC containment analysis are calculated using the CEFLASH-4A and FLOOD3 codes for LOCA events. The methodology for the short-term M/E release analysis is described in Section 3 of the technical report.

The successful application of the methodology described in these appendices demonstrates adherence to the following acceptance criteria described in NUREG-0800, Standard Review Plan, Section 6.2.1.A:1

● The containment design pressure should provide at least a 10 percent margin above the calculated peak containment pressure following LOCA or main steam line break (MSLB) accidents.

● The containment pressure should be reduced to less than 50 percent of the peak calculated pressure for the design basis LOCA within 24 hours after the postulated accident.

● The calculated external pressure of the containment structure caused by pressure and temperature changes inside the containment due to inadvertent operation of containment heat removal systems is lower than the external design pressure with a 10 percent margin.

● The highest temperature of the structures within the containment is lower than the containment design temperature for structure integrity.

Appendix A describes the modeling approach and analysis methodology for containment post-accident response analysis. It includes basic assumptions, key modeling characteristics, and analysis methods for containment response analyses to both LOCA and MSLB accidents. A description of the noding structure and use of the various GOTHIC components and models is presented including the LOCA long-term phase M/E analysis model.

Appendix B presents the application of the GOTHIC containment model to the APR1400 containment pressure and temperature response analyses. The appendix contains five subsections: LOCA peak pressure, MSLB peak temperature, MSLB peak pressure, maximum in-containment refueling water storage tank (IRWST) water temperature, and maximum passive heat sink temperature. The results presented in this section are consistent with those described in the APR1400 Design Control Document (DCD), Subsection 6.2.1.1.2

Appendix C presents the results of various sensitivity studies designed to define the postulated limiting breaks and, subsequently, the bases for the assumptions, initial conditions and modeling characteristics from a containment peak pressure and temperature standpoint. It provides bounding analyses to determine conservative initial conditions, containment model noding sensitivity, break flow droplet discharge duration, time step control, and containment spray effectiveness.

1 NUREG-0800, “Standard Review Plan,” Sections 6.2.1 and 6.2.1.1.A, “PWR Dry Containments, including Subatmospheric

Containments,” Rev. 3, U.S. Nuclear Regulatory Commission, March 2007.

2 APR1400-K-X-FS-14002, “APR1400 Design Control Document Tier 2,” KEPCO and KHNP, November 2014.


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Appendix D presents the containment external pressure loading analysis. This analysis demonstrates that the containment building and related structures are designed to accommodate the maximum external pressure load resulting from an inadvertent operation of the containment heat removal systems (i.e., spray, purge, and fan cooler systems).

Appendix E provides the conclusions of the containment post-accident response methodology and analyses described in Appendices A through D.

Appendix F describes the APR1400 containment response analysis methodology as compared to the requirements in NUREG-0800 and ANSI/ANS-56.43 guidelines.

Appendix G contains the references.

3 ANS-56.4-1983, “Pressure and Temperature Transient Analysis for Light Water Reactor Containments,” American Nuclear Society,

December 1983.


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Appendix A Methodology Description

for Containment Response Analysis


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TABLE OF CONTENTS

A. METHODOLOGY DESCRIPTION FOR CONTAINMENT RESPONSE ANALYSIS.... 1 A.1 APR1400 GOTHIC Containment Model ............................................................................... 1 A.1.1 GOTHIC Code Description ..................................................................................................... 1 A.1.2 Modeling Approach ................................................................................................................ 2 A.1.3 Analysis Approach ................................................................................................................. 2 A.2 LOCA Containment Response Analysis ............................................................................. 3 A.2.1 Assumptions .......................................................................................................................... 3

A.2.2 Key Modeling Characteristics ................................................................................................. 4

A.2.2.1 Break Flow Model .................................................................................................................. 4

A.2.2.2 Heat Transfer Model .............................................................................................................. 4

A.2.2.3 Containment Sprays ............................................................................................................... 5

A.2.2.4 Boundary Submodels ............................................................................................................. 6

A.2.2.5 Initial Conditions..................................................................................................................... 6

A.2.3 LOCA Containment Modeling ................................................................................................. 7

A.2.3.1 Control Volumes .................................................................................................................... 8

A.2.3.2 Flow Paths ........................................................................................................................... 12

A.2.3.3 Thermal Conductors ............................................................................................................. 13

A.2.3.4 Components ........................................................................................................................ 14

A.2.3.5 M/E Boundary Conditions Input ............................................................................................ 15

A.2.3.6 Forcing Functions ................................................................................................................ 16

A.2.3.7 Control Variables ................................................................................................................. 16

A.2.3.8 Other GOTHIC Modeling Parameters ................................................................................... 17

A.2.4 Decay Heat Phase M/E Analysis Model ................................................................................ 18

A.2.4.1 Sources of Energy ............................................................................................................... 18

A.2.4.2 M/E Release Calculation ...................................................................................................... 19

A.3 MSLB Containment Response Analysis ........................................................................... 20

A.3.1 Assumptions ........................................................................................................................ 20

A.3.2 Key Modeling Characteristics ............................................................................................... 21

A.3.2.1 Break Flow Model ................................................................................................................ 21

A.3.2.2 Heat Transfer Model ............................................................................................................ 21

A.3.2.3 Containment Sprays ............................................................................................................. 21

A.3.2.4 Boundary Conditions ............................................................................................................ 21

A.3.2.5 Initial Conditions................................................................................................................... 21

A.3.3 MSLB Containment Modeling ............................................................................................... 22


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A.3.3.1 Control Volumes .................................................................................................................. 22

A.3.3.2 Flow Paths ........................................................................................................................... 22

A.3.3.3 Thermal Conductors ............................................................................................................. 22

A.3.3.4 Components ........................................................................................................................ 22

A.3.3.5 M/E Boundary Conditions Input ............................................................................................ 22

A.3.3.6 Forcing Functions ................................................................................................................ 23

A.3.3.7 Control Variables ................................................................................................................. 23

A.3.3.8 Other GOTHIC Modeling Parameters ................................................................................... 23


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LIST OF TABLES

Table A-1 LOCA Containment Model Summary Description .......................................................... 24 Table A-2 Description of LOCA Containment Flow Paths .............................................................. 25 Table A-3A Passive Heat Sink Data ................................................................................................ 26 Table A-3B Material Properties of Passive Heat Sinks ..................................................................... 28 Table A-4A Water Volume of Each Component in RCS ................................................................... 28 Table A-4B Metal Mass of Each Component In RCS ....................................................................... 29 Table A-4C Metal Energy Release of the SGs Secondary Side........................................................ 29 Table A-4D Coolant Energy Release of the SGs Secondary Side .................................................... 30 Table A-5 Active Heat Sink Data ................................................................................................... 30 Table A-6 Forcing Functions for LOCA Containment Analyses ...................................................... 31 Table A-7 Control Variables for LOCA Containment Analyses ....................................................... 32 Table A-8 Decay Heat Table Used for Decay Heat Phase ............................................................. 33 Table A-9 Summary of MSLB Containment Model Description ...................................................... 36 Table A-10 Description of MSLB Containment Flow Paths .............................................................. 36 Table A-11 Forcing Functions for MSLB Containment Analyses ...................................................... 36


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LIST OF FIGURES

Figure A-1 Computer Codes Interface Diagram (LOCA) ................................................................. 38

Figure A-2 Computer Codes Interface Diagram (MSLB) ................................................................. 39

Figure A-3 LOCA Containment Response Analysis Flow Diagram.................................................. 40

Figure A-4 MSLB Containment Response Analysis Flow Diagram.................................................. 41

Figure A-5 Fluid Phases of a LOCA Discharge Flow (Cold Leg Break) ........................................... 42

Figure A-6 Noding Diagram for LOCA Containment Analyses ........................................................ 43

Figure A-7 Decay Heat Curve (Decay Heat Phase) ........................................................................ 44

Figure A-8 Noding Diagram for MSLB Containment Analyses ........................................................ 45


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A. METHODOLOGY DESCRIPTION FOR CONTAINMENT RESPONSE ANALYSIS

A.1 APR1400 GOTHIC Containment Model

A.1.1 GOTHIC Code Description

GOTHIC (Reference 5)4 is a general-purpose thermal-hydraulic computer code often used in the design, licensing, safety, and analysis of nuclear power plant containments. The code was developed by Numerical Applications (NAI), a Division of Zachry Nuclear Engineering, Inc., with support from the Electric Power Research Institute (EPRI). The GOTHIC computer code is maintained under a 10 CFR Part 50 Appendix B Quality Assurance Program.

GOTHIC performs containment integrity analysis by solving the mass, energy, and momentum conservation equations for the multi-component, multi-phase flow. Conservation equations are solved for three fields: vapor (steam/gases mixture), continuous liquid, and droplets. The vapor field is composed of steam and non-condensing gases. The gases and steam are assumed to be homogeneously mixed and in thermal equilibrium within a computational control volume (Reference 6).

The phase balance equations are coupled by mechanistic models implementing energy, mass, and momentum interchange between the vapor, continuous liquid, and droplet phases. The phase interface models allow for the possibility of thermal non-equilibrium and non-homogeneous fluid velocity.

GOTHIC solves the conservation equations of mass, energy, and momentum in lumped-parameter mode or in a multidimensional mode that considers the boundary geometry effects and the multidimensional characteristics of the flow. The GOTHIC containment model described in this document uses multiple lumped-parameter volumes. The M/E balances are maintained among multiple lumped-parameter volumes with interconnecting flow paths. A flow path models the hydraulic connection between two volumes or a volume and boundary condition.

GOTHIC models the heat transfer between solid structures and/or contacting liquid or steam/air mixtures. Containment passive heat sinks, referred to in GOTHIC as thermal conductors, are modeled as one-dimensional heat structures. Nodalization of a conductor in the direction of heat transfer allows for variation of material properties. Boundary conditions may be specified for determining the heat transfer from/to the structure.

GOTHIC provides a variety of heat transfer options for thermal conductor modeling. The models can consider condensation, boiling, natural convection, forced convection, and radiation in calculating the heat transfer at the conductor surfaces exposed to the containment environment. These include DIRECT, FILM, Tagami, and user-specified heat transfer coefficient options. The Uchida, Gido-Koestel, and diffusion layer model (DLM) options are also available to represent condensing heat transfer on the conductor surface. The various heat transfer modes (convection, radiation, boiling, condensation) are combined to get the overall heat transfer rate using the DIRECT or FILM heat transfer options in GOTHIC.

The GOTHIC code includes an extensive set of models to simulate operating equipment. These items, collectively referred to as components, include pumps and fans, valves and doors, heat exchangers, vacuum breakers, spray nozzles, coolers and heaters, hydrogen recombiners, igniters, etc. Additional resources available to expand the modeling flexibility include control functions, control systems, and trips.

GOTHIC version 8.0 is used to perform all the calculations in this report. Each version of GOTHIC is qualified against a wide range of tests (both experimental and analytical) and the results are documented in the code Qualification Report (Reference 7). The containment model and analysis methods used in the

4 References are listed in Appendix G.


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APR1400 containment accident response calculations are not restricted to a specific GOTHIC version; future containment analyses using these models on updated GOTHIC code versions may be revised, as necessary, to execute in newer GOTHIC versions with acceptable regulatory inputs, methods, and assumptions.

A.1.2 Modeling Approach

The GOTHIC code is used to develop the APR1400 containment pressure and temperature response analyses for postulated LOCA and MSLB accidents.

The LOCA containment analysis M/E release during the blowdown, reflood, and post-reflood phases is calculated by the CEFLASH-4A and FLOOD3 codes using preliminary inputs for containment pressure and the IRWST temperature. Figure A-1 shows the computer code used in each transient phase and the data interface between the M/E calculation and the containment response analysis. The results of the CEFLASH-4A and FLOOD3 M/E calculations are input to GOTHIC from the initiation of the event to the end of post-reflood (EOPR). As described in Subsection 3.10.2, the GOTHIC calculated containment pressure and IRWST water temperature are used to update the CEFLASH-4A and FLOOD3 inputs to produce more accurate short-term M/E release data. Feedback of the containment response results to the CEFLASH-4A and FLOOD3 M/E calculations is only performed once, since the impact of the GOTHIC updated containment conditions on the recalculated M/E data is negligible.

Following the EOPR, GOTHIC computes the long-term M/E release using its own reactor coolant system (RCS) simulation. During this decay heat phase, the decay heat and all of the residual energy stored in RCS and SGs metal and coolant are released into containment in the form of saturated steam through the coolant boil-off process.

For the MSLB containment analysis, the break’s M/E release data are calculated using the SGN-III computer code and the results provided as input to the GOTHIC model. In the MSLB M/E analysis, there is no feedback of the calculated containment pressure and temperature response since conservative containment conditions are used in the SGN-III calculations. Figure A-2 shows the data interface between the SGN-III and GOTHIC codes. The MSLB M/E release continues for 1,800 seconds after the accident coincident with the time that auxiliary feedwater to the affected SG is assumed terminated by operator action. Afterwards, the containment temperature and pressure rapidly decrease due to continuous operation of the containment spray system (CSS).

A.1.3 Analysis Approach

SRP Section 6.2.1.1.A describes the containment functional design requirements for PWR dry containments. Peak calculated containment pressure following a LOCA or secondary system piping rupture should be less than the containment design pressure with at least a 10 percent margin and the pressure at 24 hours after the postulated design basis accident (DBA) LOCA should be less than one-half of calculated peak pressure.

The APR1400 LOCA containment analysis is categorized into two phases depending on whether (1), the break flow is provided to GOTHIC as a boundary condition or (2), the M/E release is calculated within the code via an RCS model; the (1) short-term analysis is performed until the EOPR and the (2) long-term analysis is performed for the remainder of the accident.

Figure A-3 shows the analytical procedures for

the containment response analysis divided into short-term and long-term phases based on the EOPR.

In the GOTHIC short-term LOCA analysis period, the principal parameters and processes that impact the containment peak pressure and temperature include the externally calculated boundary condition’s M/E release data, the containment free volume, the containment heat sinks, the CSS performance and the containment initial conditions. The size of the containment’s free volume as well as the heat removal capabilities of the active and passive heat sinks are designed so that the calculated containment


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maximum pressure and temperature will remain below the structural design limits in accordance with the acceptable margins in Reference 1.

The GOTHIC long-term LOCA containment response calculations are performed similarly but the M/E release calculations are performed within the code using internal models to represent the core decay heat and the residual metal and fluid energies in the RCS and SGs.

Similar to the LOCA event, the major parameters and processes that impact the containment peak temperature and pressure for an MSLB are the M/E release from the secondary system pipe rupture, the containment heat sinks, the containment free volume, the CSS, and the containment initial conditions. Unlike the LOCA containment analysis, the M/E release from the MSLB break terminates when the SG secondary side inventory is depleted. Figure A-4 shows the MSLB containment response calculation analytical flow diagram.

A.2 LOCA Containment Response Analysis

A.2.1 Assumptions

The LOCA containment pressure and temperature response analysis is based on the assumption of loss of offsite power (LOOP) and the most severe single failure in the emergency power system, containment heat removal systems, or the core cooling systems (10 CFR 50, Appendix A, GDC 38 and 50 and SRP 6.2.1.1.A). For the APR1400 LOCA containment response analysis, a LOOP is assumed to conservatively delay the actuation of the CSS by the time required to start the emergency diesel generator (EDG).

The only active containment engineered safety feature (ESF) system for post-LOCA containment pressure and temperature considerations is the CSS. The spray system operation could be impacted by the loss of a CS pump or as a consequence from an EDG failure. In either case, one CS train will be lost. Therefore, loss of one CS train is assumed as an active single failure for the LOCA containment analyses.

The major assumptions used in the LOCA containment response analysis are listed below;

Input parameters are biased to enforce the assumptions described above. A conservative prediction is produced by considering the upper or lower bounding values of containment initial conditions, geometric parameters, and thermodynamic properties to maximize containment peak pressure and temperature as well as IRWST water temperature. The bounding initial conditions for the LOCA containment peak pressure and temperature calculations are developed with supporting sensitivity studies as described in Appendix C.1.

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A.2.2 Key Modeling Characteristics

The principal submodels used in the LOCA containment response analysis include the treatment of the break fluid, containment sprays, and the heat and mass transfer models applied to the surface of passive heat sinks and water pool surfaces. The following subsections describe the characteristics of these models.

A.2.2.1 Break Flow Model

The phase separation of the LOCA break fluid is treated depending on the transient’s discharge period, i.e., blowdown and post-blowdown phases. During the blowdown phase, the break fluid is released directly into the containment atmosphere as superheated (relative to the containment pressure) liquid and steam. Following the blowdown, the break flow is released as steam to the containment atmosphere and liquid spillage flowing to the IRWST liquid region.

By using the appropriate combination of break droplet discharge and liquid discharge during the blowdown and post-blowdown phases, a conservative containment post-LOCA response is calculated. Appendix C.3 describes sensitivity analyses performed to determine the duration of the LOCA droplet discharge. Per Appendix C.3, the droplet discharge is assumed to end at the EOB phase. Figure A-5 shows the discharge fluid phases applied to the LOCA cold leg piping breaks.

A.2.2.2 Heat Transfer Model

Wall Heat Transfer Coefficients

Containment structures in contact with the containment atmosphere are subject to condensation heat transfer following the LOCA. The direct/DLM condensation heat transfer option is used to calculate the condensation heat transfer rate on the surface of each heat structure. The use of the direct/DLM option

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for condensation heat transfer has been accepted by the NRC for peak containment pressure and temperature analysis (References 8 and 9).

The GOTHIC natural convection option is chosen for wall convection heat transfer on containment passive heat sinks.

Forced convection and radiation heat transfer from the containment atmosphere to the containment structures is conservatively excluded.

Liquid / Vapor Heat Transfer Coefficients

A.2.2.3 Containment Sprays

The containment spray droplet size is one of the primary variables that impact the spray heat removal effectiveness. GOTHIC makes use of a single mean spray drop diameter. The spray drop size can be specified from manufacturer’s data for plant-specific spray nozzles or set to a conservatively acceptable value for a particular analysis.

The drop sizes from the APR1400 specific containment spray nozzles are in the range of 90 microns to 500 microns, and the Sauter Mean Diameter (SMD) is estimated at 294 microns. The SMD is obtained from the following equation:

( )( )

3

2

i ii

i ii

n dSMD

n d

×=

×

∑∑

µm

Where:

ni: number of droplets that are in the i-th droplet diameter

di: diameter of the i-th droplet

The APR1400 containment analysis uses an assumed droplet SMD of 1,000 microns, which is typically used for containment sprays in the GOTHIC Qualification Report (Reference 7). This is a conservative assumption as larger droplets minimize droplet cooling.

A parametric study for the spray effectiveness based on spray temperature, flow rate, and fall height was performed. The results of the sensitivity study, provided in Table C-5, indicate that these spray parameters, as assumed in the APR1400 GOTHIC containment analysis CSS model, are conservative.

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A.2.2.4 Boundary Submodels

The boundary submodels used in the LOCA containment model can be grouped into three categories as shown below.

● M/E release through the break

● Containment structures (wall, dome, and base floor)

● Containment spray heat exchangers (containment heat removal system)

The approach used to develop these boundary submodels is described below.

M/E Release

GOTHIC flow boundary conditions are used to model the LOCA M/E until the EOPR. Thereafter, the GOTHIC model internally calculates the break flow from the RCS. Section 3 of this report describes the analysis method for the LOCA M/E calculation until the EOPR. The phase separation model applied to the break fluid is described in Subsection A.2.2.1.

Containment Structures

The control volume for LOCA containment analyses includes the containment’s dome, cylinder wall, and base floor. Typically, during a LOCA or secondary system pipe rupture, a large amount of energy is absorbed in these passive heat sinks and some of this heat is lost to the outside environment. However, in the APR1400 containment model, the outer surfaces of the containment shell (wall and dome) are conservatively modeled as adiabatic and the inner surface of the base floor is excluded as a passive heat sink. Subsection A.2.3.3 provides a detailed description of each passive heat sink type.

CS Heat Exchangers

Containment heat may be removed by the CS heat exchangers and containment fan coolers. In the APR1400 containment analysis, the CS heat exchanger is the only active heat sink credited. A CS heat exchanger is modeled using a GOTHIC heat exchanger component with specified inlet flow and temperature on the shell (cooling) side. Subsection A.2.3.4 details the design specification of the CS heat exchanger and the secondary side inlet conditions.

A.2.2.5 Initial Conditions

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A.2.3 LOCA Containment Modeling

The APR1400 GOTHIC containment analysis is based on a lumped-parameter model. This lumped-parameter modeling approach is acceptable for containment LOCA or MSLB response because these large breaks are expected to result in a nearly well mixed containment. Comparisons of lumped- parameter modeling and a three-dimensional model for the Carolinas-Virginia Tube Reactor (CVTR) (References 7 and 10) show that the lumped-model results in conservatively high peak pressures. GOTHIC lumped-parameter models for an extensive set of experiments have shown good agreement between the GOTHIC calculations and the experimental data (Reference 7).

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A.2.3.1 Control Volumes

GOTHIC requires volume, elevation, height, and hydraulic diameter input values for each control volume in the model. The volume is the net free volume that can be filled with steam or non-condensing gases. The reference elevation of the APR1400 containment is 100 ft,

or the level of the IRWST upper slab. The

height of each control volume is calculated from the respective plant drawings. The hydraulic diameter is calculated as four times the free volume divided by the total wetted surface area. The liquid/vapor interface area, representing the pool surface in each node, is biased to yield conservative results as required.

The volume input data for each of the model’s control volumes is summarized below.

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A.2.3.2 Flow Paths

GOTHIC flow paths connect control volumes and boundary conditions to one another. The APR1400 LOCA containment model includes a total of 17 flow paths summarized in Table A-2.

GOTHIC required input values for each flow path include elevation, height, flow area, hydraulic diameter, inertia length, and friction coefficient, which are determined from plant drawings and specifications or determined appropriately considering the impact to the model sensitivity and analysis results.

Nominal values of 1 ft2, 1 ft, 1 ft, and 0 ft are used for the GOTHIC flow area, hydraulic diameter, inertia length, and friction length, respectively for liquid flow paths. For vapor, two-phase mixture, or undetermined phase flow paths, nominal values of 100 ft2, 10 ft, 1 ft, 0 ft are used for the flow area, hydraulic diameter, inertia length, and friction length, respectively. These nominal values are chosen to minimize the flow resistance through the flow path and are deemed not to impact the results significantly.

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For model stability, non-zero K values are arbitrarily assumed for selected flow paths associated with the RCS model at the beginning of the decay heat phase.

A.2.3.3 Thermal Conductors

Passive Heat Sinks

Following a LOCA or a secondary system pipe rupture, some of the break’s energy is dissipated into the passive heat sinks within containment. In the APR1400 containment GOTHIC model, most of the structures exposed to the containment atmosphere are considered as passive heat sinks and modeled as one-dimensional thermal conductors. The lower surface area with nominal thickness of each heat structure, which considers appropriate uncertainties based on odd shapes, is determined to maximize the containment pressure and temperature response. The surface area and thickness of each passive heat sink is developed from plant drawings and specifications.

The containment structures considered as passive heat sinks are tabulated in Table A-3A. As shown in Figure A-6, a total of 18 passive heat structures are considered in the APR1400 containment model. Conservatively, structures located below the elevation of the IRWST upper slab (100 ft) are excluded. The floor surface of the containment floor is assumed to be insulated since it may be flooded during an accident. Other conservative assumptions used in modeling the containment heat structures are described below:

● The outer surface of the containment shell (wall and dome) is modeled as adiabatic to ignore heat rejection to the outside environment.

● The surface of the containment floor, including the basemat and the IRWST outside upper slab, is conservatively assumed to be insulated, i.e., no heat transfer is assumed to this internal surface, which may be flooded during an accident.

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● All internal structures that are not exposed to the containment atmosphere prior to the accident are excluded from passive heat sink modeling.

Each thermal conductor is made up of multiple material layers as appropriate. The containment shell concrete, air gap, steel liner plate, and painted surfaces are taken into account. The interface resistance between the concrete and the liner plate is set conservatively high by assuming conduction through the air gap to underestimate the heat absorption by the passive heat sinks. In addition, the heat structure’s conduction calculation mesh spacing is set small enough to provide reasonable assurance of accurate calculation of the thermal gradient through the structure.

The heat sink material thermophysical data used in the analysis is listed in Table A-4. Thermal properties for each material are minimum values based on the appropriate temperatures expected during the postulated accident.

Heat Transfer Coefficients

The heat transfer coefficients applied to each thermal conductor surface are described in Subsection A.2.2.2.

Thermal Conductor for RCS Metal Energy Release

In the APR1400 containment model, the RCS, including the primary side of the SGs, is modeled as a single lumped-parameter volume with a thermal conductor and three heater components. See the noding diagram of the RCS model in Figure A-6.

A.2.3.4 Components

GOTHIC components are used to model pumps, spray nozzles, heat exchangers, valves, and heaters within the APR1400 containment model. All the component inputs are based on plant-specific data. Valve components do not necessarily represent actual valves in the plant but rather used to control (on/off) fixed flows through modeled systems as appropriate.

Pumps

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Spray Nozzles

A GOTHIC spray nozzle component is used to model the discharge of CS liquid droplets to the containment for containment atmosphere cooling. In lumped-parameter modeling, the spray droplets immediately and uniformly fill the entire containment vapor region after spray initiation. The spray drop size primarily influences the heat transfer rate between the drops and the containment vapor. Smaller droplets have a larger total interfacial heat transfer area and reach thermal equilibrium with the containment vapor more quickly than larger droplets.

CS Heat Exchanger

The APR1400 CS heat exchanger is modeled using a GOTHIC heat exchanger component. The CS heat exchanger model uses a design fixed heat transfer coefficient (UA) value and a flow boundary condition for the shell side coolant flow at a conservatively assumed component cooling water flow rate and temperature. The CS heat exchanger input data are developed from the design specification sheets of the APR1400 CS heat exchanger and summarized in Table A-5.

Heater

A GOTHIC heater component (1H) is used to model the decay heat release. Decay heat is added directly to the RCS liquid volume and the coolant is maintained at saturated conditions by modeling the downcomer that supplies the minimum makeup required for core boil-off. The ANS 5.1-1979 (Reference 3) decay heat curve plus two sigma uncertainty is used for core decay heat power fraction and specified as a GOTHIC time-dependent forcing function (Table A-6). Additional information on the decay heat release model is presented in Subsection A.2.4, “Decay Heat Phase M/E Analysis Model.”

A.2.3.5 M/E Boundary Conditions Input

The LOCA short-term M/E release is calculated by other computer codes (CEFLASH-4A and FLOOD3) until the EOPR (or EOB for hot leg break) and provided to the GOTHIC containment model as flow boundary condition inputs. The short-term M/E release calculation methodology and results are presented in Section 4 of this report.

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The M/E release during the long-term boil-off phase (decay heat phase) is generated from the RCS model (Volume 3) incorporated within the containment model. Boundary condition (6F) is added to represent the release of the remaining SIT water not completely depleted by the EOPR. This remaining SIT water volume is released to the downcomer and eventually flows into the IRWST. The amount of SI fluid from boundary condition (6F) is relatively small and not deemed to impact the containment P/T response.

A.2.3.6 Forcing Functions

Tabulated forcing functions are used to provide input data for various boundary conditions or to control GOTHIC components. There are 20 forcing functions in the LOCA containment model and each is briefly described in the Table A-6.

A.2.3.7 Control Variables

The APR1400 LOCA containment model has 82 control variables. Control variables are used to calculate the flow rates and accumulated M/E through each flow path, the amount of M/E contained in each volume, the core heat rate, the containment spray heat exchanger (CSHX) heat removal rate, and various values for plotting. Each of these control variables is briefly described in Table A-7.

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A.2.3.8 Other GOTHIC Modeling Parameters

Revaporization Fraction

In GOTHIC, the revaporization fraction is the fraction of condensate that vaporizes from the surface of passive heat sinks. For all the APR1400 LOCA analyses, the containment atmosphere is at saturated conditions at the peak containment temperature as shown in Table B-1G. Therefore, the GOTHIC revaporization option has no impact on the analyses results. This option is set to DEFAULT. The DEFAULT option is used in the GOTHIC code qualification calculations (Reference 7) and has been approved for PWR containment analysis (Reference 8).

Minimum Heat Transfer Coefficient

The GOTHIC RUN option for minimum heat transfer coefficient specifies the lower limit on convection heat transfer that applies to liquid-vapor interfacial heat transfer at a pool surface. The parameter also sets the minimum mass transfer coefficient at the same boundary. It is set to 0.0 Btu/hr-ft2-oF to avoid the cooling effect by the relatively cold IRWST water. The effective area for heat and mass transfer rate at the pool surfaces in the APR1400 model is set to zero, and this parameter therefore has no impact on the results.

Reference Pressure

The reference pressure option is set to IGNORE (default) to use the local pressure and temperature to calculate density in the gravitational force term of the vapor momentum equation. This is the recommended, default, setting.

Forced Entrainment Drop Diameter

The diameter of entrained drops for specified entrainment modeling is in subdivided volumes. This parameter is set to DEFAULT (0.1 in) since the containment model is based on the lumped-parameter approach. There is no specified entrainment modeling in the APR1400 model so this parameter has no impact on the results.

Vapor Phase Head Correction

This option is set to INCLUDE. This is the code recommend setting. When set to INCLUDE, the static head of the pool liquid that is above the cell centerline is subtracted from the vapor phase pressure, and the calculated results are physically more realistic than results when this parameter is set to IGNORE. In the APR1400 containment model, this parameter has no impact on the containment atmosphere volume since the pool liquid level is significantly lower than the volume centerline.

Kinetic Energy

The kinetic energy transport and storage in the fluid energy equations is set to IGNORE. The kinetic energy in the break fluid is already included in the M/E release calculations.

Vapor, Liquid, and Drop Phase Options

The vapor phase, including non-condensing gases, liquid, and droplets are all modeled in the APR1400 containment response analysis. Thus, vapor, liquid, and droplet conservation equations are solved for each fluid phase. These options are set to INCLUDE.


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Force Equilibrium

If this option is set to INCLUDE, the interphase heat and mass transfer coefficients are set to a large value to force all phases to be in thermal equilibrium. This option is set to IGNORE to allow the code to perform non-equilibrium calculations of interfacial heat transfer as appropriate.

Drop-Liquid Conversion

The INCLUDE option enables the droplet entrainment, agglomeration, and deposition. The value is set to INCLUDE to allow break drops to appropriately transport to the containment floor and flow to the IRWST.

A.2.4 Decay Heat Phase M/E Analysis Model

The GOTHIC computer code is used to calculate the M/E release rate during long-term boil-off. The analysis for this decay heat phase utilizes a simplified RCS model to calculate the M/E release to containment.

The RCS is modeled to calculate the maximum water boil-off from the decay heat and release of stored energy. The steam is released directly to the containment, conservatively neglecting any interaction with the subcooled SI water.

A.2.4.1 Sources of Energy

The following energy sources are accounted for in the decay heat phase M/E calculation:

● Core decay energy

● RCS metal stored energy

● RCS coolant stored energy

● SGs secondary side metal stored energy

● SGs secondary side coolant energy

Core Decay Energy

Core decay heat is modeled using a GOTHIC heater component with a time-dependent heat rate. The decay heat is calculated using ANSI/ANS 5.1-1979 with 2σ uncertainties for full reactor power (References 3 and 8). The decay heat contribution from actinides other than U-239 and Np-239 is taken into account in accordance with Reference G-13. Table A-8 lists the core power fraction as a function of time and associated decay heat curves are shown in Figure A-7.

The reactor power used in the decay heat rate calculation is based on the full core power plus 2 percent uncertainty (3,983 MW x 1.02). The decay heat data are input via a forcing function (18T, Table A-6) and the forcing function multiplied by the full-power core heat generation rate (plus uncertainty) to yield the decay heat power as a function of time.

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RCS Metal Stored Energy

The energy in the RCS primary side metal at the EOPR is taken into account as an initial metal stored energy for the decay heat phase. A thermal conductor is used to model the stored metal energy; see Subsection A.2.3.3. The temperature of the conductor at the beginning of the decay heat phase is assumed to be the same as the RCS coolant temperature at EOPR.

RCS Coolant Stored Energy

The RCS coolant volume/level at the EOPR is kept steady during the entire decay heat phase by the downcomer model. The water volume in the RCS is listed in Table A-4A. The RCS coolant initial temperature is set at the EOPR temperature and the stored energy calculated by GOTHIC.

A.2.4.2 M/E Release Calculation

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The water boil-off generated from the decay heat and metal stored energy is at saturated conditions and released to containment through the break. The long-term energy release rate calculated from the GOTHIC code is given as follows:

This decay heat phase M/E is now simulated by modeling of the simplified RCS in GOTHIC as previously described in this section.

A.3 MSLB Containment Response Analysis

This section describes the major assumptions and key modeling characteristics applied to the containment response analyses for MSLB events. It also describes the noding structure, component models, and the other GOTHIC modeling parameters used in the MSLB containment model.

A.3.1 Assumptions

In general, the MSLB M/E release, with the assumption of offsite power available, leads to more severe containment conditions than with LOOP. Without a LOOP, the peak containment pressure and temperature exceed those with a LOOP even under assumption of loss of one CS train and delayed spray actuation caused by EDG startup. For the APR1400, offsite power is assumed to be available in the MSLB containment analyses to maximize heat transfer to the secondary side of the affected SG. Thus, unlike the LOCA analysis that assumes a 20-second delay on CS initiation for EDG startup, the MSLB analysis assumes no CS initiation time delay. The single active failures considered in MSLB containment analyses are the loss of one CS train caused by a CS pump failure or an MSIV failure to close.

All other assumptions used in the MSLB containment analyses are the same as those described in Subsection A.2.1 for the LOCA event with the exception that there is no nitrogen release from the SITs.

Input parameters are determined based on both plant-specific data and the assumptions described above. The initial conditions for the MSLB containment peak temperature calculation are determined based on the results of sensitivity analyses described in Appendix C.1.

s gm h×�

energy release rate to the containment (Btu/sec)

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A.3.2 Key Modeling Characteristics

A.3.2.1 Break Flow Model

The break fluid from the affected SG is discharged into containment as superheated steam and immediately mixes within the vapor region until the water inventory is depleted.

A.3.2.2 Heat Transfer Model

A.3.2.3 Containment Sprays

The Subsection A.2.2.3 discussion on the APR100 containment spray nozzles droplet diameter also applies to the MSLB analysis.

A.3.2.4 Boundary Conditions

The MSLB containment analysis boundary conditions include the M/E release, energy removal through the CS heat exchangers, and energy rejection through the outside surface of the containment building, which is ignored for conservatism.

The M/E releases following an MSLB accident are introduced to the containment atmosphere via a GOTHIC flow boundary condition. The containment thermal structures and CS heat exchanger models are the same as those for the LOCA analysis described in Subsection A.2.2.4.

A.3.2.5 Initial Conditions

The superheated steam break flow for the MSLB accident results in higher containment peak temperature compared to a LOCA event. Conversely, the MSLB containment peak pressure is lower than that from the limiting LOCA event.

The analyses described in Appendix B.3 demonstrate that all of the MSLB containment response calculations using initial conditions biased to give peak pressure are bounded by the maximum pressure determined in design basis LOCA containment analyses. Therefore, the MSLB containment response analyses are biased to maximize the containment temperature. The upper- and lower-bound values used as initial conditions for the MSLB containment response sensitivity analyses, cover the entire range of LCOs specified in the plant Technical Specifications as presented in Table B-2A. The outside ambient conditions are not relevant due to the adiabatic assumption used for the containment building outside wall heat transfer.

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A.3.3 MSLB Containment Modeling

A.3.3.1 Control Volumes

For the MSLB containment response analyses, the containment atmosphere region (Volume 1) and IRWST (Volume 2) are the same as in the LOCA containment model. The input data for these two volumes are described in Subsection A.2.3.1.

A.3.3.2 Flow Paths

The MSLB containment model includes three flow paths as described in Table A-10. Flow path J15 represents the broken pipe steam flow discharged from the affected SG to the containment vapor region. All the GOTHIC flow path parameters are the same as those described in Subsection A.2.3.2 for LOCA.

A.3.3.3 Thermal Conductors

The containment passive heat sinks and heat transfer coefficients discussion provided in Subsection A.2.3.3 for the LOCA containment analyses also applies to the MSLB.

A.3.3.4 Components

GOTHIC built-in components such as a pump, spray nozzle and a heat exchanger are used in the MSLB containment analyses model. Input values for these components are based on plant-specific data. For the CS pump, automatic start operation from a high-high containment pressure setpoint includes a time delay of 90 seconds. The 20-second delay for EDG startup in the LOCA analysis is not required for the MSLB analysis. All other input parameters used for the CS pump, spray nozzle, and CS heat exchanger are exactly the same as described in Subsection A.2.3.4.

A.3.3.5 M/E Boundary Conditions Input

The MSLB break fluid from the secondary side of the affected SG is essentially superheated steam. Therefore, the break fluid from boundary condition (7F) is discharged to the containment atmosphere as 100 percent steam during the entire discharge period.

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A.3.3.6 Forcing Functions

Tabulated forcing functions are used to provide input data for various boundary conditions or to control GOTHIC components. There are three forcing function tables in the APR1400 MSLB containment model and each is described in Table A-11.

A.3.3.7 Control Variables

The MSLB containment model has 27 control variables. These control variables are used to calculate the flow rates and accumulated M/E through each flow path, the amount of M/E contained in each volume, the CSHX heat removal rate, and various values for plotting. Each of these control variables is described in Table A-12.

A.3.3.8 Other GOTHIC Modeling Parameters

The Subsection A.2.3.8 GOTHIC modeling parameter discussion for the LOCA containment analyses also applies to MSLB analysis except for the GOTHIC revaporization option described below.

Revaporization Fraction

The revaporization fraction is the fraction of condensate that vaporizes from the surface of passive heat sinks when the containment atmosphere is superheated. For MSLB containment analyses, DIRECT/DLM option is chosen as the wall condensation heat transfer model that includes revaporization. The revaporization fraction is set to DEFAULT to allow GOTHIC to calculate the re-evaporation. The default revaporization option was used for all of the validation cases in the GOTHIC Qualification Report (References 7 and 8) and found acceptable by the NRC in Reference 9.


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Table A-1. LOCA Containment Model Summary Description

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Table A-1 (cont.). LOCA Containment Model Summary Description

Table A-2. Description of LOCA Containment Flow Paths

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Table A-3A. Passive Heat Sink Data

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Table A-3A (cont.). Passive Heat Sink Data

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Table A-3B. Material Properties of Passive Heat Sinks

Materials

Density Specific Heat Thermal Conductivity

kg/m3 lbm/ft3 J/kg-°C Btu/lbm°F W/m-°C Btu/hr-ft-°F

Concrete 2,242.6 140 879.2 0.21 1.592 0.8

Carbon steel 7,817.0 488 460.5 0.11 46.383 26.8

Stainless steel 7,817.0 488 460.5 0.11 15.940 9.21

Inorganic zinc paint 5,462.3 341 891.8 0.213 1.004 0.58

Epoxy paint 1,417.6 88.5 1,272.8 0.304 0.277 0.16

Air 0.961 0.06 720.1 0.172 0.030 0.0174

The properties for each material are assumed constant at the specified values based on the accident expected temperatures.

Table A-4A. Water Volume of Each Component in RCS

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Table A-4B. Metal Mass of Each Component In RCS

Table A-4C. Metal Energy Release of SGs Secondary Side

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Table A-4D. Coolant Energy Release of SGs Secondary Side

Table A-5. Active Heat Sink Data

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Table A-6. Forcing Functions for LOCA Containment Analyses

No. Description Dependent

Variable

1T Mass release rate during the blowdown phase lbm/s

2T Specific enthalpy of the mass during the blowdown phase Btu/lbm

3T Heat transfer coefficient of the CS heat exchanger (constant) Btu/hr-ft2-oF

4T Number of operating CS pumps, multiplier 1 or 2

5T Mass flow rate for B/C (3F) to adjust the RV water level lbm/s

6T Atmosphere pressure (14.7 psia, constant) psia

7T Atmosphere temperature (120 oF, constant) oF

8T Containment design pressure (60 psig, constant) psia

9T Number of operating SIPs, multiplier 3 or 4

10T SIT N2 gas release rate after EOB multiplier

11T Vapor mass release rate during the reflood and post-reflood phases lbm/s

12T Vapor enthalpy during the reflood and post-reflood phases Btu/lbm

13T Cold spillage mass release rate during the reflood and post-reflood phases lbm/s

14T Cold spillage enthalpy during the reflood and post-reflood phases Btu/lbm

15T Hot spillage mass release rate during the reflood and post-reflood phases Btu/lbm

16T Hot spillage enthalpy during the reflood and post-reflood phases Btu/lbm

17T Mass release rate of the remaining SIT water after the EOPR lbm/s

18T Normalized decay energy release rate decay rate

19T SGs secondary side metal sensible energy release rate 106 Btu/s

20T SGs secondary side coolant energy release rate 106 Btu/s


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Table A-7. Control Variables for LOCA Containment Analyses

No. Description

1C Break flow rate (J13), reflood/post-reflood

2C Accumulated vapor flow (J13)

3C SIP mass flow rate (J4)

4C Accumulated SIP mass flow (J4)

5C Break flow (Cold spillage) rate (J5)

6C Accumulated cold spillage (J5)

7C Containment Steam partial pressure

8C Containment relative humidity

9C Containment total pressure

10C IRWST liquid temperature

11C IRWST total pressure

12C RCS total pressure (Measured)

13C RCS Pr. (Used prior to the decay heat phase)

14C Break flow rate (J11), long-term phase

15C Accumulated break flow (J11)

16C Spillage flow rate (J9), long-term phase

17C Liquid mass in the IRWST

18C Vapor mass in the IRWST

19C Droplet mass in the IRWST

20C Liquid mass in the containment

21C Vapor mass in the containment

22C Droplet mass in the Containment

23C Liquid energy in the IRWST

24C Vapor energy in the IRWST

25C Droplet energy in the IRWST

26C Liquid energy in the containment

27C Vapor energy in the containment

28C Droplet energy in the containment

29C Total mass (L/V/D) in the containment

30C Total energy (L/V/D) in the containment

31C Total mass (L/V/D) in the IRWST

32C Total energy (L/V/D) in the IRWST

33C Total mass in the containment and IRWST

34C Total energy in the containment and IRWST

35C RCS liquid temperature


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No. Description

36C Liquid mass in the RCS

37C SIT liquid flow rate (J14) after EOPR

38C RCS metal temperature

39C Liquid mass in the downcomer

40C Liquid energy in the RCS

41C Liquid mass in RCS and downcomer

42C Liquid energy in RCS and Downcomer

43C Decay energy release rate

44C Accumulated decay energy

45C Liquid energy in the Downcomer

46C Total mass (L/V/D) in the entire volumes

47C Total energy (L/V/D) in the entire volumes

48C Accumulated nitrogen mass to containment

49C SIT nitrogen energy release rate (J12)

50C Accumulated nitrogen energy to containment

51C Break flow (Hot spillage) rate (J3)

52C Accumulated hot spillage to containment(J3)

53C CSHX heat transfer rate (Btu/s)

54C Accumulated heat to the HX shell side (Btu)

55C Spillage (hot+cold) flow rate (J3+J5)

56C Accumulated spillage (hot+cold) (J3+J5)

57C Accumulated break flow (J11), (=15C)

58C Vapor energy release rate from RCS (J11)

59C Accumulated vapor energy from RCS(J11)

60C Spillage energy release rate (J9)

61C Accumulated spillage energy from RCS (J9)

62C SIP mass flow rate (J4) (=3C)

63C Accumulated SIP mass flow (J4) (=4C)

64C SIP energy flow rate (J4)

65C Accumulated SIP energy flow (J4)

66C Mass flow rate (Downcomer→RCS) (J10)

67C Accumulated mass flow (Downcomer→RCS)

68C Energy flow rate (Downcomer→RCS) (J10)

69C Accumulated energy flow (downcomer→RCS)

70C SIT liquid flow rate (J14), (=37C)

71C Accumulated SIT liquid flow after EOPR


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No. Description

72C SIT liquid energy rate (J14) after EOPR

73C Accumulated SIT liquid energy after EOPR

74C SIT nitrogen mass release rate (J12)

75C Break flow rate (J15) during blowdown phase

76C Accumulated break flow (J15)

77C Energy content of RCB atmosphere

78C Energy contents of containment passive heat sinks

79C Energy contents of RCS metal

80C Saturation temperature at RCS total pressure

81C 0 or 1 (1: Before EOPR, 0: After EOPR)

82C 0 or 1 (1: Before EOB, 0: After EOB)


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Table A-8. Decay Heat Table Used for Decay Heat Phase

Time (sec)

Decay Heat Rate

Time (sec)

Decay Heat Rate

ANS 5.1 1979+2σ

ANS 5.1 1979+2σ with Actinide (1)

ANS 5.1 1979+2σ

ANS 5.1 1979+2σ with Actinide (1)

10 0.0538760 0.0544148 15,000 0.0100970 0.0101980

15 0.0504010 0.0509050 20,000 0.0093500 0.0094435

20 0.0480180 0.0484982 40,000 0.0077780 0.0078558

40 0.0424010 0.0428250 60,000 0.0069580 0.0070276

60 0.0392440 0.0396364 80,000 0.0064240 0.0064882

80 0.0370650 0.0374357 100,000 0.0060210 0.0060812

100 0.0354660 0.0358207 150,000 0.0053230 0.0058553

150 0.0327240 0.0330512 200,000 0.0048470 0.0053317

200 0.0309360 0.0312454 400,000 0.0037700 0.0041470

400 0.0270780 0.0273488 600,000 0.0032010 0.0035211

600 0.0249310 0.0251803 800,000 0.0028340 0.0031174

800 0.0233890 0.0236229 1,000,000 0.0025800 0.0028380

1,000 0.0221560 0.0223776 2,000,000 0.0019090 0.0020999

1,500 0.0199210 0.0201202 4,000,000 0.0013550 0.0014905

2,000 0.0183150 0.0184982 6,000,000 0.0010910 0.0012001

4,000 0.0147810 0.0149288 8,000,000 0.0009270 0.0010197

6,000 0.0130400 0.0131704 10,000,000 0.0008080 0.0008888

8,000 0.0120000 0.0121200 100,000,000 0.0008080 0.0008888

10,000 0.0112620 0.0113746 (1) Decay heat contribution from actinides other than U-239 and Np-239 is additionally considered in accordance with

Reference 13.


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Table A-9. Summary of MSLB Containment Model Description

Table A-10. Description of MSLB Containment Flow Paths

Table A-11. Forcing Functions for MSLB Containment Analyses

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Table A-12 Control Variables for MSLB Containment Analyses

No. Description No. Description

1C Break flow rate (J13) 24C Vapor energy in the IRWST

2C Accumulated vapor flow (J13) 25C Droplet energy in the IRWST

7C Containment steam partial pressure 26C Liquid energy in the containment

8C Containment relative humidity 27C Vapor energy in the containment

9C Containment total pressure 28C Droplet energy in the containment

10C IRWST liquid temperature 29C Total mass (L/V/D) in the containment

11C IRWST total pressure 30C Total energy (L/V/D) in the containment

17C Liquid mass in the IRWST 31C Total mass (L/V/D) in the IRWST

18C Vapor mass in the IRWST 32C Total energy (L/V/D) in the IRWST

19C Droplet mass in the IRWST 33C Total mass in the containment and IRWST

20C Liquid mass in the containment 34C Total energy in the containment and IRWST

21C Vapor mass in the containment 53C CSHX heat transfer rate (Btu/s)

22C Droplet mass in the containment 54C Accumulated heat to the HX shell side (Btu)

23C Liquid energy in the IRWST Most of the control variables used in the MSLB containment model are essentially the same as those used in the LOCA containment model presented in Table A-7. However, only 27 control variables, as described above, are used for MSLB containment model.


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Figure A-1. Computer Codes Interface Diagram (LOCA)


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Figure A-2. Computer Codes Interface Diagram (MSLB)


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Figure A-3. LOCA Containment Response Analysis Flow Diagram

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Figure A-4. MSLB Containment Response Analysis Flow Diagram

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Figure A-5. Fluid Phases of a LOCA Discharge Flow (Cold Leg Break)

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Figure A-6. Noding Diagram for LOCA Containment Analyses

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102 103 104 105 106 107 1080.00

0.01

0.02

0.03

0.04

ANS 5.1 1979 + 2 sigmas

(with Actinide decay heat power)ANS 5.1 1979 + 2 sigmas

Decay heat Phase

De

cay

Heat

(Fra

ctio

n of

Ope

ratin

g Po

wer)

Time after Shutdown(sec)

EOPR (LOCA)

Figure A-7. Decay Heat Curve (Decay Heat Phase)


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Figure A-8. Noding Diagram for MSLB Containment Analyses

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Appendix B Calculations Using

the APR1400 Containment Model


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TABLE OF CONTENTS

B. CALCULATIONS USING THE APR1400 CONTAINMENT MODEL ........................ 1 B.1 LOCA Peak Pressure Analysis ............................................................................................ 1 B.1.1 Accident Description .............................................................................................................. 1 B.1.1.1 Blowdown .............................................................................................................................. 1 B.1.1.2 Reflood .................................................................................................................................. 2 B.1.1.3 Post-Reflood .......................................................................................................................... 2 B.1.1.4 Decay Heat Boil-Off ............................................................................................................... 2 B.1.2 Input Parameters ................................................................................................................... 2 B.1.2.1 Initial Conditions..................................................................................................................... 2 B.1.2.2 M/E Release Data .................................................................................................................. 3 B.1.3 Calculations ........................................................................................................................... 3 B.1.3.1 Double-Ended Suction Leg Slot Break (DESLSB) with Maximum SI ....................................... 3 B.1.3.2 Double-Ended Suction Leg Slot Break (DESLSB) with Minimum SI ........................................ 4

B.1.3.3 Double-Ended Discharge Leg Slot Break (DEDLSB) with Maximum SI ................................... 4

B.1.3.4 Double-Ended Discharge Leg Slot Break (DEDLSB) with Minimum SI .................................... 5

B.1.3.5 Double-Ended Hot Leg Slot Break (DEHLSB)......................................................................... 5

B.1.4 Results .................................................................................................................................. 5

B.2 MSLB Peak Temperature ..................................................................................................... 6 B.2.1 Description of Accident........................................................................................................... 6 B.2.2 Input Parameters ................................................................................................................... 6 B.2.2.1 Initial Conditions..................................................................................................................... 6 B.2.2.2 M/E Release Data .................................................................................................................. 6 B.2.3 Calculations ........................................................................................................................... 7 B.2.4 Results .................................................................................................................................. 7 B.3 MSLB Peak Pressure ........................................................................................................... 8 B.3.1 Input Parameters ................................................................................................................... 8 B.3.2 Calculation and Results.......................................................................................................... 8 B.4 IRWST Maximum Water Temperature ................................................................................. 8 B.4.1 Calculation and Results.......................................................................................................... 9 B.5 Maximum Passive Heat Sink Temperature ......................................................................... 9 B.6 Conclusions ......................................................................................................................... 9


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LIST OF TABLES

Table B-1A Initial Conditions – LOCA Peak Pressure ...................................................................... 10 Table B-1B Sequence of Events – LOCA (DESLSB with Maximum SI Flow) ................................... 11 Table B-1C Sequence of Events – LOCA (DESLSB with Minimum SI Flow) .................................... 12 Table B-1D Sequence of Events – LOCA (DEDLSB with Maximum SI Flow) ................................... 12 Table B-1E Sequence of Events – LOCA (DEDLSB with Minimum SI Flow) .................................... 13 Table B-1F Sequence of Events – LOCA (DEHLSB with Maximum SI Flow) ................................... 13 Table B-1G Summary of LOCA Peak Pressure ................................................................................ 14 Table B-2A Initial Conditions – MSLB Peak Temperature ................................................................ 14 Table B-2B Spectrum of Postulated MSLB Accidents ...................................................................... 15 Table B-2C Sequence of Events – MSLB (102 Percent Power, Loss of a CSS Train) ...................... 15 Table B-2D Sequence of Events – MSLB (102 Percent Power, MSIV Single Failure) ....................... 16 Table B-2E Sequence of Events – MSLB (75 Percent Power, Loss of a CSS Train) ........................ 16 Table B-2F Sequence of Events – MSLB (75 Percent Power, MSIV Single Failure) ......................... 17 Table B-2G Sequence of Events – MSLB (50 Percent Power, Loss of a CSS Train) ........................ 17 Table B-2H Sequence of Events – MSLB (50 Percent Power, MSIV Single Failure) ......................... 18 Table B-2I Sequence of Events – MSLB (20 Percent Power, Loss of a CSS Train) ........................ 18 Table B-2J Sequence of Events – MSLB (20 Percent Power, MSIV Single Failure) ......................... 19 Table B-2K Sequence of Events – MSLB (0 Percent Power, Loss of a CSS Train) .......................... 19 Table B-2L Sequence of Events – MSLB (0 Percent Power, MSIV Single Failure)........................... 20 Table B-2M Summary of MSLB Peak Temperature .......................................................................... 21 Table B-3A Initial Conditions – MSLB Peak Pressure ...................................................................... 22 Table B-3B Summary of MSLB Peak Pressure ................................................................................ 23 Table B-4A Initial Conditions for Maximum IRWST Water Temperature ........................................... 24 Table B-4B Summary of Maximum IRWST Water Temperature (LOCA) .......................................... 25 Table B-5A Maximum Surface Temperature of Each Passive Heat Sink

(LOCA – DEDLSB with Max. SI Flow) ........................................................................... 26 Table B-5B Maximum Surface Temperature of Each Passive Heat Sink

(MSLB – 75 Percent Power, Loss of a CSS Train) ........................................................ 27 Table B-6 Summary of Containment Integrity Analyses ................................................................. 28


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LIST OF FIGURES

Figure B-1A LOCA Mass Release (DESLSB with Max. SI) ............................................................... 29 Figure B-1B LOCA Mass Release (DESLSB with Max. SI) (Long Term) ........................................... 30 Figure B-1C LOCA Containment P/T Transients (DESLSB with Max. SI) .......................................... 31 Figure B-1D LOCA Containment P/T Transients (DESLSB with Max. SI)

(Long Term) ................................................................................................................. 32 Figure B-2A LOCA Mass Release (DESLSB with Min. SI) ................................................................ 33 Figure B-2B LOCA Mass Release (DESLSB with Min. SI) (Long Term) ............................................ 34 Figure B-2C LOCA Containment P/T Transients (DESLSB with Min. SI) ........................................... 35 Figure B-2D LOCA Containment P/T Transients (DESLSB with Min. SI) (Long Term) ....................... 36 Figure B-3A LOCA Mass Release (DEDLSB with Max. SI) ............................................................... 37 Figure B-3B LOCA Mass Release (DEDLSB with Max. SI) (Long Term) ........................................... 38 Figure B-3C LOCA Containment P/T Transients (DEDLSB with Max. SI) .......................................... 39 Figure B-3D LOCA Containment P/T Transients (DEDLSB with Max. SI)

(Long Term) ................................................................................................................. 40 Figure B-4A LOCA Mass Release (DEDLSB with Min. SI) ................................................................ 41 Figure B-4B LOCA Mass Release (DEDLSB with Min. SI) (Long Term) ............................................ 42 Figure B-4C LOCA Containment P/T Transients (DEDLSB with Min. SI) ........................................... 43 Figure B-4D LOCA Containment P/T Transients (DEDLSB with Min. SI) (Long Term) ....................... 44 Figure B-5A LOCA Mass Release (DEHLSB with Max. SI) ............................................................... 45 Figure B-5B LOCA Mass Release (DEHLSB with Max. SI) (Long Term) ........................................... 46 Figure B-5C LOCA Containment P/T Transients (DEHLSB with Max. SI) .......................................... 47 Figure B-5D LOCA Containment P/T Transients (DEHLSB with Max. SI)

(Long Term) ................................................................................................................. 48 Figure B-6A MSLB Mass Release (102 % Power, Loss of a CSS Train) ........................................... 49 Figure B-6B MSLB Containment P/T Response (102 % Power, Loss of a CSS Train)....................... 50 Figure B-7A MSLB Mass Release (102 % Power, MSIV Failure) ...................................................... 51 Figure B-7B MSLB Containment P/T Response (102 % Power, MSIV Failure).................................. 52 Figure B-8A MSLB Mass Release (75 % Power, Loss of a CSS Train) ............................................. 53 Figure B-8B MSLB Containment P/T Response (75 % Power, Loss of a CSS Train) ........................ 54 Figure B-9A MSLB Mass Release (75 % Power, MSIV Failure) ........................................................ 55 Figure B-9B MSLB Containment P/T Response (75 % Power, MSIV Failure).................................... 56 Figure B-10A MSLB Mass Release (50 % Power, Loss of a CSS Train) ............................................. 57 Figure B-10B MSLB Containment P/T Response (50 % Power, Loss of a CSS Train) ........................ 58 Figure B-11A MSLB Mass Release (50 % Power, MSIV Failure) ........................................................ 59 Figure B-11B MSLB Containment P/T Response (50 % Power, MSIV Failure).................................... 60 Figure B-12A MSLB Mass Release (20 % Power, Loss of a CSS Train) ............................................. 61


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Figure B-12B MSLB Containment P/T Response (20 % Power, Loss of a CSS Train) ........................ 62 Figure B-13A MSLB Mass Release (20 % Power, MSIV Failure) ........................................................ 63 Figure B-13B MSLB Containment P/T Response (20 % Power, MSIV Failure).................................... 64 Figure B-14A MSLB Mass Release (0 % Power, Loss of a CSS Train) ............................................... 65 Figure B-14B MSLB Containment P/T Response (0 % Power, Loss of a CSS Train) .......................... 66 Figure B-15A MSLB Mass Release (0 % Power, MSIV Failure) .......................................................... 67 Figure B-15B MSLB Containment P/T Response (0 % Power, MSIV Failure) ..................................... 68 Figure B-16 MSLB Temperature Transients (Peak Temperature Analyses) ...................................... 69 Figure B-17 MSLB Pressure Transients (Peak Pressure Analyses) .................................................. 70 Figure B-18A Surface Temperature Transients of Passive Heat Sinks

(LOCA – DEDLSB with Max. SI Flow) ........................................................................... 71 Figure B-18B Surface Temperature Transients of Passive Heat Sinks

(MSLB – 75 % Power, Loss of a CSS Train) ................................................................. 72 Figure B-19A IRWST Water Temperature Transients (LOCAs, MSLBs).............................................. 73 Figure B-19B Maximum IRWST Water Temperature (LOCA – DEDLSB with Min. SI) ......................... 74


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B. CALCULATIONS USING THE APR1400 CONTAINMENT MODEL

This section demonstrates the use of the GOTHIC containment model described in Appendix A for the APR1400 containment post-accident response analysis. The section contains five subsections: LOCA peak pressure, MSLB peak temperature, MSLB peak pressure, maximum IRWST water temperature, and maximum passive heat sink temperature.

A LOCA containment analysis is performed to determine the containment peak pressure. The MSLB is analyzed for containment peak temperature and it is shown that the calculated containment peak pressure for the LOCA event is bounding over the MSLB peak pressure.

Sections B.1 and B.2 present the results of containment response calculations to the LOCA peak pressure and MSLB peak temperature, respectively. Section B.3 demonstrates that the results from MSLB pressure analyses using initial conditions biased for maximum peak pressure are bounded by the LOCA maximum peak pressure.

Calculations for maximum IRWST water temperature are presented in Section B.4 and describe the conservatism in initial conditions set to maximize the IRWST water temperature. The calculation results are provided to estimate the minimum net positive suction head available (NPSHa) for the SIPs and CS pumps.

Section B.5 is presented to demonstrate that the highest surface temperature of heat structures in containment is lower than the containment design temperature limit, thus providing reasonable assurance of the containment integrity.

The LOCA and MSLB containment response analyses documented in Sections B.1 and B.2 are consistent with those described in the APR1400 Design Control Document (DCD), Subsection 6.2.1.1 (Reference 2).

B.1 LOCA Peak Pressure Analysis

This section addresses the containment pressure response following a LOCA event. All calculations were performed using the containment model described in Appendix A. The following subsections contain a description of the LOCA scenario, input parameters, calculations, and results.

B.1.1 Accident Description

As described in Section 3, cold leg reactor coolant pump (RCP) suction, cold leg RCP discharge, and hot leg breaks are considered as large break LOCA events for containment response analysis. The LOCA M/E release for a cold leg break is divided into four distinct phases: blowdown, reflood, post-reflood, and decay heat boil-off. The following is a description of each phase from the containment response perspective.

B.1.1.1 Blowdown

The blowdown phase encompasses the rapid release of two-phase reactor coolant inventory into containment and ends when most of the RCS inventory has been released and the RCS pressure approaches containment pressure. The blowdown phase is typically completed within 20 seconds of a large break LOCA and the initial containment peak pressure occurs toward the end of this phase.

The CS pump starts the supply of subcooled water to the CS nozzles with a time delay of 110 seconds after the containment pressure reaches the high-high setpoint. Liquid droplets from the CS nozzles rapidly condense the containment atmosphere’s steam, reducing the containment temperature and pressure.


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B.1.1.2 Reflood

The reflood phase follows the end of blowdown. Subcooled water from the SITs refills the RV until the core liquid level reaches the height at which core liquid entrainment ends. During the reflood phase, the SI water is injected into the upper part of the downcomer and the active core is quenched as the RV water level increases. The reflood process generates a large amount of steam and entrained water, which is then released into containment through the break.

For the RCP suction leg break, the containment peak pressure occurs before the end of the reflood phase. This suction break’s peak pressure also occurs earlier than for an equivalent discharge leg break as the SG secondary side energy is released to containment at a higher rate compared to the RCP discharge leg break.

B.1.1.3 Post-Reflood

The post-reflood phase starts after the end of the reflood phase and continues until the temperatures of the RCS and the steam generators are essentially equal (Reference 4). For the RCP discharge pipe break accidents, the containment maximum pressure occurs during this period. After the peak containment pressure is reached, the continuous flow of subcooled water CS spray droplets into the containment atmosphere gradually decreases containment pressure and temperature.

B.1.1.4 Decay Heat Boil-Off

The final post-blowdown phase, referred to as the decay heat phase, is a relatively stable period characterized by the release of reactor core decay heat and RCS/SGs metal sensible energy to containment. During this period, M/E release to containment continues through the boil-off of RCS coolant by heat addition from the remaining stored energy in RCS metal and coolant as well as reactor decay heat. The containment heat removal system, including the CS nozzles and heat exchangers, continues to condense steam and cool the containment atmosphere. Heat is removed from the CSS by the component cooling water system on the secondary side of the CS heat exchanger. Consequently, the containment pressure and temperature monotonically decrease during this phase of the transient.

For the hot leg pipe break, flow is cut off to the broken-loop steam generator and there is no heat transfer between the primary and secondary sides of the affected steam generator. Hence, for the hot leg break, only the blowdown and decay heat phases require modeling. During the decay heat phase, all the SG stored energy, including the broken-loop SG, as well as the RCS stored energy at the EOB, are considered in the long-term M/E release model.

B.1.2 Input Parameters

B.1.2.1 Initial Conditions

A peak pressure sensitivity analysis for containment initial conditions was performed and documented in Appendix C. As shown in Table B-1A, higher initial pressure and temperature and lower initial relative humidity were shown to yield maximum peak containment pressure. The assumed IRWST initial water volume is the minimum IRWST volume considering the water accumulated in containment that cannot return to the tank. In addition, the IRWST initial water temperature is assumed as the highest IRWST water temperature at normal operating conditions TS. These conservatisms in the LOCA event initial conditions are further described in Section C.1.


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B.1.2.2 M/E Release Data

As described in Appendix A, the LOCA M/E release data until the EOPR is calculated separately and specified as flow boundary conditions inputs to GOTHIC. Thereafter, the RCS M/E release is directly calculated by GOTHIC. The time-dependent M/E release data for each LOCA event case are depicted in Figures B-1(A,B) through B-5(A,B). The ME analysis methodology up to the EOPR period is described in Section 3 of this report. Section A.2.4 of Appendix A describes the LOCA M/E methodology during the decay heat phase.

B.1.3 Calculations

The containment response to five LOCA events was considered: two cases of the RCP suction leg break with maximum and minimum SI, two cases of RCP discharge leg break with maximum and minimum SI flow, and one hot leg break with maximum SI. The containment response analyses are extended up to 1 million seconds (11.6 days) to include all the relevant aspects of the transient. The containment LOCA pressure, temperature, and IRWST water temperature histories are illustrated in Figures B-1(C,D) through B-5(C,D).

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B.2 MSLB Peak Temperature

This section describes the APR1400 containment temperature and pressure response analysis following a main steam line break event. The MSLB containment analysis model described in Section A.3 is used.

B.2.1 Description of Accident

Following a postulated MSLB inside containment, steam and water from the affected SG shell (secondary) side will be released into containment. An MSLB can release sufficient amounts of high-energy fluid to produce high temperature and pressure conditions inside containment.

The bulk of the unaffected SG contents will be isolated by the main steam isolation valves (MSIVs). The containment pressurization rate following a secondary side piping break is dependent on the amount of break fluid entering the containment atmosphere as steam. A break size spectrum analysis is performed to determine the break size that results in the worst-case containment response. The quantitative nature of the MSLB release is dependent upon break size, plant operating conditions, and the potential configurations of plant steam systems.


B.2.2.1 Initial Conditions

A bounding MSLB, although releasing less total M/E than a LOCA, discharges nearly 100 percent steam with a relatively higher specific energy than a LOCA break. Therefore, it results in higher containment peak temperature but lower peak pressure as compared to the bounding LOCA event. As indicated in Section B.3, all MSLB containment analyses based on initial conditions biased to maximize peak pressure result in containment peak pressure bounded by the design LOCA event. In this section, the MSLB containment analysis’ initial conditions were biased for maximum peak temperature calculation. Section C.1 documents the initial conditions’ sensitivity analysis performed in support of MSLB containment peak temperature calculations.

The selected bounding initial conditions for MSLB peak temperature analysis are the same as those used in the LOCA peak pressure analysis except for containment initial pressure; see Table B-2A. A minimum initial pressure is more conservative since lower pressure decreases the initial cold air mass, resulting in a rapid increase in containment temperature. The minimum initial pressure also delays the time to reach the containment spray system actuation setpoint. The conservatism used in selecting bounding initial conditions for the MSLB containment analysis is described in Section C.1.

B.2.2.2 M/E Release Data

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B.2.3 Calculations

A total of 10 MSLB cases covering various power levels and break sizes were postulated. The M/E release from the break continues until the operator manually terminates the auxiliary feedwater flow to the affected SG. All the MSLB containment response analyses are performed for 1 million seconds.

Following an MSLB, containment temperature and pressure rise rapidly due to the large influx of steam from the break. Break flow drops abruptly approximately 10 seconds into the accident as the intact SG steam release is terminated by MSIV closure. The CS pump begins to deliver subcooled water to the spray nozzles, with a 90-second time delay, after the containment high-high setpoint is reached. Liquid droplets from the CS nozzles rapidly condense the steam atmosphere and cool the superheated containment to saturation temperature. The containment thus reaches its peak temperature just prior to CS flow initiation and cools to saturated conditions for the remainder of the transient.

Assumed operator action terminates auxiliary feedwater flow to the affected SG 1,800 seconds into the event followed by a rapid reduction in containment pressure and temperature.

B.2.4 Results

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B.3 MSLB Peak Pressure In general, the maximum containment peak pressure and temperature occur in LOCA and MSLB accidents, respectively. Accordingly, the LOCA containment response analyses were performed to determine the maximum containment peak pressure and the maximum peak temperature was estimated from MSLB peak temperature analyses. Hence, it is necessary to demonstrate that the maximum peak pressure from the postulated MSLB accidents is no higher than that determined from the design basis LOCA. The same cases chosen for MSLB peak temperature calculations described in Section B.2 are used for the peak pressure analysis.


Table B-3A presents the initial conditions used for the MSLB peak pressure calculations. Each of these initial values is biased to maximize the containment peak pressure. All of the initial values are identical to those for the LOCA analyses described in Section B.1 except those for the safety injection tank (SIT), which does not contribute to the MSLB events.

The M/E release data used for this calculation are the same as for the MSLB peak temperature calculations described in Subsection B.2 and depicted in Figures B-6(A) through B-15(A).

B.3.2 Calculation and Results

B.4 IRWST Maximum Water Temperature

This section addresses the calculation of the maximum IRWST water temperature during LOCA and MSLB accidents. This maximum temperature may be used as input to minimum NPSHa calculations for pumps that use IRWST water as the working fluid.

A primary input to the maximum IRWST water temperature analysis is the IRWST initial water volume. A lower IRWST initial water volume leads to a higher IRWST water temperature in the long-term cooling phase and thus, a higher SI water supply temperature to the RCS.

In the APR1400 containment, it is assumed that the liquid falling to the containment bottom from the containment spray and break flow is returned to the IRWST except for the amount of water that fills the “dead” volumes such as the reactor cavity and the HVT. Therefore, the IRWST initial water volume is conservatively determined from a minimum water level considering the maximum amount of water trapped in these “dead” volumes during LOCA events.

The impact of initial conditions with respect to the maximum IRWST water temperature calculation is described in Section C.1. Table B-4A lists the bounding initial conditions chosen for the maximum IRWST water temperature analysis. The same assumptions and M/E release data used for the LOCA peak pressure analysis described in Section B.1 are applied to this calculation.

The sensitivity of the maximum IRWST maximum water temperature to the effective containment pool surface area as well as from the single-node and two-node containment models was determined and documented in Section C.2. Per Section C.2, the two-volume containment model and the assumption of zero heat transfer at the pool surface produced the maximum IRWST water temperature.

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B.4.1 Calculation and Results

B.5 Maximum Passive Heat Sink Temperature

B.6 Conclusions

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Table B-1A. Initial Conditions – LOCA Peak Pressure

Initial Conditions

Values

Design LOCA

Containment building

Atmosphere pressure, kgf/cm2A (psia) 0.997~1.133

(14.18~16.12) 1.133

(16.12) (1)

Atmosphere temperature, °C (°F) 10.0~48.9 (50 ~ 120)

48.9 (120)

Atmosphere relative humidity (%) 0 ~ 100 0.0

Component cooling water temperature, °C (°F) 43.3 (110) (2) 43.3 (110)

IRWST water temperature, °C (°F) 10.0~48.9 (50 ~ 120)

48.9 (120)

Outside air temperature, °C (°F) 10.0~48.9 (50 ~ 120)

N/A (3)

Containment inner wall temperature, °C (°F) 10.0~48.9 (50 ~ 120)

48.9 (120)

Water and N2 gas in IRWST and SITs

(1) The maximum initial pressure is determined based on LCO described in Technical Specifications Subsection 3.6.4 with uncertainty, (14.7 + 1.0 (LCO max. value) + 0.42 (pressure indicator uncertainty) ) psia.

(2) The CCW system and service water systems are not explicitly modeled. A conservatively high CCW water temperature of 110 °F (43.3 °C) was used for the containment peak pressure and temperature analysis during the entire transient.

(3) Heat loss to the environment through the containment enclosure is conservatively neglected. (4) Range of water volume in the IRWST at normal operating conditions. (5) Minimum water volume estimated from the lowest IRWST water level during LOCAs; considers water trapped in

containment dead volumes such as ICI cavity, HVT, and sumps.

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Table B-1B. Sequence of Events – LOCA (DESLSB with Maximum SI Flow)

Time (sec) Events Setpoint

0.0 Break occurs

4.38 Containment pressure Hi-Hi setpoint

14.01 Start SIT (SIT) injection

19.20 End of blowdown

Start SIP injection

19.21 First peak containment pressure (blowdown phase)

63.70 SIT flow is turned down to low flow by fluidic device in SIT

101.91 Peak containment temperature

102.51 Peak containment pressure

110.20 End of reflood

114.38 Start containment spray actuation

125.00 End of post-reflood

368.10 SIT empty

57,660.1 Time of depressurization of the containment at 50 % of peak pressure

TS


KEPCO & KHNP B12

Non-Proprietary

Table B-1C. Sequence of Events – LOCA (DESLSB with Minimum SI Flow)


0.0 Break occurs


14.01 Start SIT injection


Start SIP injection








368.30 SIT empty


Table B-1D. Sequence of Events – LOCA (DEDLSB with Maximum SI Flow)


0.0 Break occurs





Start SIP injection







372.05 SIT empty


TS

TS


KEPCO & KHNP B13

Non-Proprietary

Table B-1E. Sequence of Events – LOCA (DEDLSB with Minimum SI Flow)


0.0 Break occurs





Start SIP injection






369.65 SIT empty



Table B-1F. Sequence of Events – LOCA (DEHLSB with Maximum SI Flow)


0.0 Break occurs


6.80 Start SIP injection

13.01 13.01


Peak containment pressure



TS

TS


KEPCO & KHNP B14

Non-Proprietary

Table B-1G. Summary of LOCA Peak Pressure

Break Location

DESLSB(1) DEDLSB(2) DEHLSB(3)

MaximumSI Flow

Minimum SI Flow

Maximum SI Flow

Minimum SI Flow

Maximum SI Flow

Total Break Area, m² (ft²)

Peak pressure, kgf/cm²A (psia) [psig]

Time of peak pressure (sec)

Peak temperature, °C (°F)

Time of peak temperature (sec)

Max. saturated temperature, °C (°F)

Pressure margin (%)

(1) DESLSB: double-ended suction leg slot break (2) DEDLSB: double-ended discharge leg slot break (3) DEHLSB: double-ended hot leg slot break

Table B-2A. Initial Conditions – MSLB Peak Temperature

Initial Conditions

Values

Design MSLB



(14.18~16.12) 0.997

(14.18) (1)


48.9 (120)


Component cooling water temperature, °C (°F) 43.3 (110) 43.3 (110)


48.9 (120)


N/A

Containment inner wall temperature, °C (°F)

IRWST water volume, m³ (ft³)

(1) The MSLB limiting temperature case conservatively uses the minimum initial pressure (14.7- 0.1 (LCO min. value) - 0.42 (indicator uncertainty) ) psia.

(2) Minimum water volume estimated from the lowest IRWST water level during an MSLB. This volume was calculated by subtracting all the water trapped in holdup volumes from the initial IRWST water volume and adding the water volume of the SITs (Reference 12).

TS

TS


KEPCO & KHNP B15

Non-Proprietary

Table B-2B. Spectrum of Postulated MSLB Accidents

Main Steam Line Break ESF Single-Failure

Condition Break Area,

m² (ft²)

MSLB at 102 % power Loss of one CSS train

MSIV failure


MSIV failure


MSIV failure


MSIV failure


MSIV failure

Table B-2C. Sequence of Events – MSLB (102 % Power, Loss of a CSS Train)

Time (sec) Events Values

0.0 Break occurs

3.89 Containment pressure reaches reactor trip analysis setpoint and main steam isolation signal analysis setpoint

5.20 High containment pressure reactor trip signal

5.30 Reactor trip breakers open, turbine admission valves closed

10.40 Main steam isolation valves closed

15.40 Main feedwater isolation valves closed


125.51 Start containment spray injection



1,800.0 End of blowdown

3,503.2 Time of depressurization of the containment to 50 % of peak pressure

TS

TS


KEPCO & KHNP B16

Non-Proprietary

Table B-2D. Sequence of Events – MSLB (102 % Power, MSIV Single Failure)


0.0 Break occurs












Table B-2E. Sequence of Events – MSLB (75 % Power, Loss of a CSS Train)


0.0 Break occurs












TS

TS


KEPCO & KHNP B17

Non-Proprietary

Table B-2F. Sequence of Events – MSLB (75 % Power, MSIV Single Failure)


0.0 Break occurs







112.21 112.21

Start containment spray injection





Table B-2G. Sequence of Events – MSLB (50 % Power, Loss of a CSS Train)


0.0 Break occurs












TS

TS


KEPCO & KHNP B18

Non-Proprietary

Table B-2H. Sequence of Events – MSLB (50 % Power, MSIV Single Failure)


0.0 Break occurs







22.61 Containment pressure Hi- Hi setpoint





Table B-2I. Sequence of Events – MSLB (20 % Power, Loss of a CSS Train)


0.0 Break occurs












TS

TS


KEPCO & KHNP B19

Non-Proprietary

Table B-2J. Sequence of Events – MSLB (20 % Power, MSIV Single Failure)


0.0 Break occurs












Table B-2K. Sequence of Events – MSLB (0 % Power, Loss of a CSS Train)


0.0 Break occurs












TS

TS


KEPCO & KHNP B20

Non-Proprietary

Table B-2L. Sequence of Events – MSLB (0 % Power, MSIV Single Failure)


0.0 Break occurs







119.11 119.11

Start containment spray injection





TS


KEPCO & KHNP B21

Non-Proprietary

Table B-2M. Summary of MSLB Peak Temperature

Break Location 102 % Power 75 % Power 50 % Power

CSS Failure

MSIV Failure

CSS Failure

MSIV Failure

CSS Failure

Total break area, m² (ft²)






Pressure margin (%)


MSIV Failure

CSS Failure

MSIV Failure CSS Failure MSIV

Failure







Pressure margin (%)

TS

TS


KEPCO & KHNP B22

Non-Proprietary

Table B-3A. Initial Conditions – MSLB Peak Pressure

Initial Conditions

Values

Design MSLB



(14.18~16.12) 1.133

(16.12)


48.9 (120)




48.9 (120)


N/A

Containment inner wall temperature, °C (°F)


(1) Water volume estimated from the minimum IRWST water level during an MSLB considering water trapped in containment volumes such as ICI cavity, HVT, and sumps.

TS


KEPCO & KHNP B23

Non-Proprietary

Table B-3B. Summary of MSLB Peak Pressure


CSS Failure

MSIV Failure CSS Failure

MSIV Failure

CSS Failure







Pressure margin (%)


MSIV Failure

CSS Failure

MSIV Failure CSS Failure MSIV

Failure




Peak temperature, °C(°F)


Max. saturated temperature, °C(°F)

Pressure margin (%)

TS

TS


KEPCO & KHNP B24

Non-Proprietary

Table B-4A. Initial Conditions for Maximum IRWST Water Temperature

Initial Conditions

Values

Design Analysis



(14.18~16.12) 0.997

(14.18)


48.9 (120)

Atmosphere relative humidity (%) 0 ~ 100 100



48.9 (120)


N/A

Containment inner wall temperature, °C (°F) 10.0~48.9 (50 ~ 120)

48.9 (120)

Water and N2 gas in IRWST and SITs


N2 gas in SI tanks, kg (lb)

(1) Water volume estimated from the minimum IRWST water level during an MSLB considering water trapped in containment volumes such as ICI cavity, HVT, and sumps

TS


KEPCO & KHNP B25

Non-Proprietary

Table B-4B. Summary of Maximum IRWST Water Temperature (LOCA)

Break Location DESLSB DEDLSB DEHLSB

Maximum SI Flow

Minimum SI Flow

Maximum SI Flow

Minimum SI Flow

Maximum SI Flow

Max. IRWST water temperature, °C (°F)

Time at the max. temperature, seconds

TS


KEPCO & KHNP B26

Non-Proprietary

Table B-5A. Maximum Surface Temperature of Each Passive Heat Sink (LOCA – DEDLSB with Max. SI Flow)

No. Description Time (sec)

Max. Temp. °C (°F)

1 Containment cylinder wall above El. 100ʹ 425.3 127.6 (261.7)

2 Containment dome 400.8 128.2 (262.7)

3 Containment basement 0.0 48.9 (120.0)

4 Concrete embedded carbon steel 893.0 124.8 (256.6)

5 Concrete without embedment plate (C, A) 360.2 128.2 (262.7)

6 Refuel pool with SS liner 366.2 134.3 (273.7)

7 IRWST outside upper slab and HVT 0.0 48.9 (120.0)

8 Polar crane and bridge, M 466.3 132.8 (271.0)

9 Safety injection tank 1,746.2 125.8 (258.4)

10 Group A (structural steel and PWR), C 581.8 131.6 (268.9)

11 Group B (grating and metal decking), C 362.2 134.4 (273.8)

12 Group C (pipe support and HVAC), P, C, M 363.4 134.2 (273.6)

13 Group D (miscellaneous steel), A, J, M 361.5 134.4 (273.9)

14 Group E (electrical components), E 343.3 134.5 (274.1)

15 Group F (miscellaneous steel from P, A, J, N [CS]) 373.6 134.0 (273.2)

16 Group G (miscellaneous steel from P, N, J [SS]) 371.4 134.2 (273.6)

17 Group J (CS from NSSS DOOSAN) 397.3 133.8 (272.9)

18 Group K (SS from NSSS DOOSAN) 3,056.8 123.3 (254.0)


KEPCO & KHNP B27

Non-Proprietary

Table B-5B. Maximum Surface Temperature of Each Passive Heat Sink (MSLB – 75 % Power, Loss of a CSS Train)

No. Description Time (sec)

Max. Temp. °C (°F)

1 Containment cylinder wall above El. 100ʹ 1,799.1 128.0 (262.4)

2 Containment dome 545.7 128.4 (263.1)

3 Containment basement 0.0 48.9 (120.0)

4 Concrete embedded carbon steel 1,799.1 127.7 (261.8)

5 Concrete without embedment plate (C, A) 502.9 128.2 (262.8)

6 Refuel pool with SS liner 484.8 133.4 (272.2)

7 IRWST outside upper slab and HVT 0.0 48.9 (120.0)

8 Polar crane and bridge, M 580.2 132.6 (270.7)

9 Safety injection tank 1,799.1 129.6 (265.4)

10 Group A (structural steel and PWR), C 694.0 131.9 (269.5)

11 Group B (grating and metal decking), C 478.1 133.5 (272.3)

12 Group C (pipe support & HVAC), P, C, M 490.3 133.4 (272.1)

13 Group D (miscellaneous steel), A, J, M 476.6 133.5 (272.3)

14 Group E (electrical components), E 468.7 133.57 (272.43)

15 Group F (miscellaneous steel from P, A, J, N [CS]) 502.9 133.3 (271.9)

16 Group G (miscellaneous steel from P, N, J [SS]) 493.8 133.4 (272.1)

17 Group J (CS from NSSS DOOSAN) 524.2 133.2 (271.7)

18 Group K (SS from NSSS DOOSAN) 1,820.2 126.4 (259.5)


KEPCO & KHNP B28

Non-Proprietary

Table B-6. Summary of Containment Integrity Analyses

Component Peak condition Accident Unit

Calculated Value

Design Limit Margin

Containment

IRWST

(1) Containment design pressure (2) Ratio of the containment pressure @ 24 hours to the calculated peak pressure. (3) Maximum containment atmosphere temperature that permits EQ for safety-related equipment in containment. (4) Containment design temperature (allowable maximum surface temperature of the inner structures within containment)

TS


KEPCO & KHNP B29

Non-Proprietary

0 100 200 300 400 500 600100

101

102

103

104

105

106

Blowdown

End of post-reflood (EOPR)

End of reflood (EOR)

Hot Spillage

Steam

Brea

k flo

w (lb

m/s

)

Time (sec)

End ofblowdown(EOB)

Cold Spillage

Figure B-1A. LOCA Mass Release (DESLSB with Max. SI)


KEPCO & KHNP B30

Non-Proprietary

0.0 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106100

101

102

103

104

105

106

Decay Heat Phase

Spillage

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam

End of post-reflood

Figure B-1B. LOCA Mass Release (DESLSB with Max. SI) (Long Term)


KEPCO & KHNP B31

Non-Proprietary

0 100 200 300 400 500 60010.0

20.0

30.0

40.0

50.0

60.0

70.0

Containment Temperature

Containment Pressure

Pres

sure

(psia

)

Time (sec)

114.38 secSpray actuation

@ 102.51 secPmax= 4.546 kgf/cm2A(64.66 psia)

@ 101.91 secTmax= 133.84 °C(272.91 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-1C. LOCA Containment P/T Transients (DESLSB with Max. SI)


KEPCO & KHNP B32

Non-Proprietary

100 101 102 103 104 105 1060.0

10.0

20.0

30.0

40.0

50.0

60.0

70.0

IRWST Liquid Temperature


Containment PressurePr

essu

re (p

sia)

Time (sec)

@ 19,411 secTmax= 117.24 °C(243.03 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-1D. LOCA Containment P/T Transients (DESLSB with Max. SI) (Long Term)


KEPCO & KHNP B33

Non-Proprietary

0 100 200 300 400 500 600100

101

102

103

104

105

106

Blowdown

EOPR

EOR

Hot Spillage

Steam

Brea

k flo

w (lb

m/s

)

Time (sec)

EOB

Cold Spillage

Figure B-2A. LOCA Mass Release (DESLSB with Min. SI)


KEPCO & KHNP B34

Non-Proprietary

0.0 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106100

101

102

103

104

105

106

Decay Heat Phase

Spillage

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam

End of post-reflood

Figure B-2B. LOCA Mass Release (DESLSB with Min. SI) (Long Term)


KEPCO & KHNP B35

Non-Proprietary

0 100 200 300 400 500 60010.0

20.0

30.0

40.0

50.0

60.0

70.0



Pres

sure

(psia

)

Time (sec)



@ 104.21 secTmax= 133.97 °C(273.14 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-2C. LOCA Containment P/T Transients (DESLSB with Min. SI)


KEPCO & KHNP B36

Non-Proprietary

100 101 102 103 104 105 1060.0

10.0

20.0

30.0

40.0

50.0

60.0

70.0




essu

re (p

sia)

Time (sec)

@ 18,710 secTmax= 116.34 °C(241.42 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-2D. LOCA Containment P/T Transients (DESLSB with Min. SI) (Long Term)


KEPCO & KHNP B37

Non-Proprietary

0 100 200 300 400 500 600100

101

102

103

104

105

106

EOPREOR

Hot Spillage

Steam

Brea

k flo

w (lb

m/s

)

Time (sec)

EOB

Cold Spillage

Blowdown

Figure B-3A. LOCA Mass Release (DEDLSB with Max. SI)


KEPCO & KHNP B38

Non-Proprietary

0.0 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106100

101

102

103

104

105

106

Decay Heat Phase

Spillage

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam

End of post-reflood

Figure B-3B. LOCA Mass Release (DEDLSB with Max. SI) (Long Term)


KEPCO & KHNP B39

Non-Proprietary

0 100 200 300 400 500 60010.0

20.0

30.0

40.0

50.0

60.0

70.0



Pres

sure

(psia

)

Time (sec)



@ 323.82 secTmax= 134.59 °C(274.27 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-3C. LOCA Containment P/T Transients (DEDLSB with Max. SI)


KEPCO & KHNP B40

Non-Proprietary

100 101 102 103 104 105 1060.0

10.0

20.0

30.0

40.0

50.0

60.0

70.0




essu

re (p

sia)

Time (sec)

@ 17,409 secTmax= 118.93 °C(246.07 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-3D. LOCA Containment P/T Transients (DEDLSB with Max. SI) (Long Term)


KEPCO & KHNP B41

Non-Proprietary

0 100 200 300 400 500 600100

101

102

103

104

105

EOPREOR

Hot Spillage

Brea

k flo

w (lb

m/s

)

Time (sec)

EOB

Cold Spillage

Blowdown

Steam

Figure B-4A. LOCA Mass Release (DEDLSB with Min. SI)


KEPCO & KHNP B42

Non-Proprietary

0.0 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106100

101

102

103

104

105

106

Decay Heat Phase

Spillage

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam

End of post-reflood

Figure B-4B. LOCA Mass Release (DEDLSB with Min. SI) (Long Term)


KEPCO & KHNP B43

Non-Proprietary

0 100 200 300 400 500 60010.0

20.0

30.0

40.0

50.0

60.0

70.0



Pres

sure

(psia

)

Time (sec)



@ 320.42 secTmax= 134.52 °C(274.13 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-4C. LOCA Containment P/T Transients (DEDLSB with Min. SI)


KEPCO & KHNP B44

Non-Proprietary

100 101 102 103 104 105 1060.0

10.0

20.0

30.0

40.0

50.0

60.0

70.0




essu

re (p

sia)

Time (sec)

@ 16,307 secTmax= 118.53 °C(245.35 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-4D. LOCA Containment P/T Transients (DEDLSB with Min. SI) (Long Term)


KEPCO & KHNP B45

Non-Proprietary

0 20 40 60 80 100100

101

102

103

104

105

106

Blowdown

Brea

k flo

w (lb

m/s

)

Time (sec)

EOB

Figure B-5A. LOCA Mass Release (DEHLSB with Max. SI)


KEPCO & KHNP B46

Non-Proprietary

0.0 2.0x105 4.0x105 6.0x105 8.0x105 1.0x106100

101

102

103

104

105

106

Decay Heat Phase

Spillage

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam

End of blowdown

Figure B-5B. LOCA Mass Release (DEHLSB with Max. SI) (Long Term)


KEPCO & KHNP B47

Non-Proprietary

0 100 200 300 400 500 60010.0

20.0

30.0

40.0

50.0

60.0

70.0



Pres

sure

(psia

)

Time (sec)



@ 13.01 secTmax= 132.99 °C(271.38 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-5C. LOCA Containment P/T Transients (DEHLSB with Max. SI)


KEPCO & KHNP B48

Non-Proprietary

100 101 102 103 104 105 1060.0

10.0

20.0

30.0

40.0

50.0

60.0

70.0




Pres

sure

(psia

)

Time (sec)

@ 26,618 secTmax= 115.20 °C(239.36 °F)

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-5D. LOCA Containment P/T Transients (DEHLSB with Max. SI) (Long Term)


KEPCO & KHNP B49

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 2000100

101

102

103

104

105

106

End of discharge

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam

Figure B-6A. MSLB Mass Release (102 % Power, Loss of a CSS Train)


KEPCO & KHNP B50

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 200010.0

20.0

30.0

40.0

50.0

60.0

70.0


Pres

sure

(psia

)

Time (sec)

Spray actuation125.51 sec

Tmax= 164.85 °C(328.72 °F)@ 125.51 sec

Pmax= 4.173 kgf/cm2A(59.36 psia)@ 345.41 sec


Aux. Feedwater Stop1800 sec

100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-6B. MSLB Containment P/T Response (102 % Power, Loss of a CSS Train)


KEPCO & KHNP B51

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 2000100

101

102

103

104

105

106

End of discharge

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam

Figure B-7A. MSLB Mass Release (102 % Power, MSIV Failure)


KEPCO & KHNP B52

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 200010.0

20.0

30.0

40.0

50.0

60.0

70.0


Pres

sure

(psia

)

Time (sec)


Tmax= 167.45 °(333.41 °FC)@ 111.71 sec




100

150

200

250

300

350

400

Tem

pera

ture

(o F)

Figure B-7B. MSLB Containment P/T Response (102 % Power, MSIV Failure)


KEPCO & KHNP B53

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 2000100

101

102

103

104

105

106

End of discharge

Brea

k flo

w (lb

m/s

)

Time (sec)

Steam



KEPCO & KHNP B54

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 200010.0

20.0

30.0

40.0

50.0

60.0

70.0


Pres

sure

(psia

)

Time (sec)


Tmax= 163.52 °C(326.34 °F)@ 126.31 sec




100

150

200

250

300

350

400

Tem

pera

ture

(o F)



KEPCO & KHNP B55

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 2000100

101

102

103

104

105

106

End of discharge

Brea

k flo

w (l

bm/s

)

Time (sec)

Steam



KEPCO & KHNP B56

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 200010.0

20.0

30.0

40.0

50.0

60.0

70.0


Pres

sure

(psia

)

Time (sec)


Tmax= 165.28 °C(329.51 °F)@ 112.21 sec




100

150

200

250

300

350

400

Tem

pera

ture

(o F)



KEPCO & KHNP B57

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 2000100

101

102

103

104

105

106

End of discharge

Brea

k flo

w (l

bm/s

)

Time (sec)

Steam



KEPCO & KHNP B58

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 200010.0

20.0

30.0

40.0

50.0

60.0

70.0


Pres

sure

(psia

)

Time (sec)


Tmax= 161.90 °C(323.43 °F)@ 127.51 sec




100

150

200

250

300

350

400

Tem

pera

ture

(o F)



KEPCO & KHNP B59

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 2000100

101

102

103

104

105

106

End of discharge

Brea

k flo

w (l

bm/s

)

Time (sec)

Steam



KEPCO & KHNP B60

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 200010.0

20.0

30.0

40.0

50.0

60.0

70.0

Containment TemperaturePres

sure

(psia

)

Time (sec)


Tmax= 163.77 °C(326.79 °F)@ 22.51 sec




100

150

200

250

300

350

400

Tem

pera

ture

(o F)



KEPCO & KHNP B61

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 2000100

101

102

103

104

105

106

End of discharge

Brea

k flo

w (l

bm/s

)

Time (sec)

Steam



KEPCO & KHNP B62

Non-Proprietary

0 200 400 600 800 1000 1200 1400 1600 1800 200010.0

20.0

30.0

40.0

50.0

60.0

70.0

Containment TemperaturePres

sure

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Tmax= 157.37 °C(315.27 °F)@ 97.81 sec




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100 101 102 103 104100.0

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Figure B-16. MSLB Temperature Transients (Peak Temperature Analyses)


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100 101 102 103 10410.0

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Figure B-17. MSLB Pressure Transients (Peak Pressure Analyses)


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Figure B-18A. Surface Temperature Transients of Passive Heat Sinks (LOCA – DEDLSB with Max. SI Flow)

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Figure B-18B. Surface Temperature Transients of Passive Heat Sinks (MSLB – 75 % Power, Loss of a CSS Train)

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10-1 100 101 102 103 104 10550.0

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MSLBs

IRW

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MSLB (0% Power, an MSIV single-failure)Max. temperature: 226.77 oF

Max. temperature: 246.47 oFLOCA (Double-ended discharge leg slot break with Min. SI)

Figure B-19A. IRWST Water Temperature Transients (LOCAs, MSLBs)


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Figure B-19B. Maximum IRWST Water Temperature (LOCA – DEDLSB with Min. SI)

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Appendix C Case Studies for Modeling Characteristics


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TABLE OF CONTENTS

C. CASE STUDIES FOR MODELING CHARACTERISTICS ........................................ 1

C.1 Bounding Analysis for Initial Conditions ............................................................................... 1 C.1.1 Initial Conditions for Peak Pressure ........................................................................................... 1 C.1.2 Initial Conditions for Peak Temperature ..................................................................................... 1 C.1.3 Initial Conditions for IRWST Maximum Water Temperature ........................................................ 2 C.2 Comparison of Nodalization for Containment P/T and IRWST .............................................. 2 C.3 Sensitivity Analysis for Duration of Droplet Discharge ......................................................... 3 C.4 Sensitivity Analysis for Time Step Control ............................................................................ 4 C.5 Parametric Study for Containment Spray Effectiveness ....................................................... 4


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LIST OF TABLES

Table C-1A Containment Analysis Bounding Initial Conditions ........................................................... 6 Table C-1B LOCA Bounding Initial Conditions for Peak Pressure ...................................................... 6 Table C-1C MSLB Bounding Initial Conditions for Peak Temperature ................................................ 6 Table C-1D LOCA Bounding Initial Conditions for Maximum IRWST Temperature ............................. 6 Table C-2A Comparison of Volume Size for Containment Models ...................................................... 7 Table C-2B Results of Noding Structure Sensitivity Analysis .............................................................. 7 Table C-3A Duration of Droplet Discharge ......................................................................................... 8 Table C-3B Peak Pressure Results vs. Duration of Droplet Discharge ............................................... 8 Table C-4A Peak Pressures vs. Various Time Step Sizes .................................................................. 8 Table C-4B Peak Temperatures vs. Various Time Step Sizes ............................................................ 9 Table C-4C Time Step Size Applied to Each Time Domain ................................................................ 9 Table C-5 List of Parameters for Containment Spray Effectiveness ............................................... 10


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LIST OF FIGURES

Figure C-1A Pressure Transients to Various Initial Conditions (LOCA) .............................................. 11 Figure C-2A Comparison of Noding Structure (Single Node / Two Nodes) ........................................ 14 Figure C-2B Comparison of Containment Pressure (LOCA – DEDLSB with Max. SI) ........................ 15 Figure C-2C Comparison of Containment Temperature

(MSLB – 102 Percent Power, MSIV Failure) ................................................................. 16 Figure C-2D Comparison of IRWST Water Temperature

(LOCA – DEDLSB with Max. SI) ................................................................................... 17


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C. CASE STUDIES FOR MODELING CHARACTERISTICS

This appendix presents sensitivity analyses performed to support conservative GOTHIC containment modeling characteristics. They include initial conditions, noding sensitivity, duration of droplet discharge for break flow, time step control, and spray effectiveness.

C.1 Bounding Analysis for Initial Conditions In order to provide assurance that the LOCA/MSLB containment response analyses provide conservative results, containment initial conditions are considered in sensitivity analyses designed to maximize containment pressure or temperature, or IRWST water temperature.

The key containment initial conditions affecting peak transient conditions include pressure, temperature, and relative humidity. An appropriate combination of upper or lower bounding values of these parameters should be used to maximize containment peak pressure or temperature, or IRWST water temperature following the postulated break.

The bounding limits for each of the considered initial conditions are those documented in the limiting conditions for operation (LCO) specified in the DCD Technical Specifications. The upper and lower limits of each initial condition are tabulated in Table C-1A.

C.1.1 Initial Conditions for Peak Pressure

The containment response to a LOCA results in a higher peak pressure than for MSLB accidents. The highest peak pressures in LOCA and MSLB accidents are summarized in Tables B-1G and B-3B, respectively. The impact of each initial condition on containment peak pressure is individually evaluated for the limiting LOCA event (double-ended discharge leg slot break with maximum SI flow).

Four cases that vary the containment’s initial condition limiting values are chosen to determine their effect on LOCA peak pressure. As documented in Table C-1B, the results of this LOCA peak pressure sensitivity study indicate that the combination of maximum initial pressure and temperature with minimum relative humidity results in the highest containment peak pressure. Figure C-1A depicts the calculated pressure transients for each set of initial conditions described in Table C-1B.

The upper-bound limit of containment initial pressure maximizes the mass of non-condensing gases and steam in containment at the accident initiation. In the determination of containment initial temperature, the upper-bound value nominally results in a higher peak pressure from the competing effects of the initial amount of non-condensing gases and the thermal absorption of passive heat sinks as demonstrated by the sensitivity analysis. A lower relative humidity is conservative from the peak pressure standpoint since it increases the initial ratio of non-condensing gases to steam volume, which ultimately reduces the condensation rate when steam is released to the containment.

The IRWST initial water volume is also considered a parameter affecting the containment peak pressure. A larger IRWST water volume reduces the containment initial atmosphere space (which includes the IRWST air volume) and results in a slightly higher peak containment pressure. In the APR1400 containment model, a combination of the minimum IRWST water and air volume, calculated from the lowest and highest water levels, respectively, is conservatively chosen as an initial condition for all the analyses.

C.1.2 Initial Conditions for Peak Temperature

As shown in the summary results of containment response to LOCA and MSLB events listed in Tables B-1G and B-2M, the highest peak temperature occurs following an MSLB accident. Hence, the effect of


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each containment’s initial condition on peak temperature is determined for the worst MSLB accident (102 percent power level with an MSIV single failure).

As documented in Table C-1C, the combination of minimum initial pressure, maximum initial temperature, and minimum relative humidity results in the maximum containment peak temperature following an MSLB. Figure C-1B depicts the calculated temperature transients for each set of initial conditions described Table C-1C.

The containment minimum initial pressure and maximum initial temperature minimize the initial mass of air and steam in containment and reduce heat losses to the containment heat structures. Furthermore, the lower initial pressure delays reaching the containment high-high pressure setpoint and consequently delays spray actuation. The minimum relative humidity is conservative for peak temperature determinations since a higher air mass fraction reduces the condensation rate and decreases the initial heat capacity of containment vapor region.

C.1.3 Initial Conditions for IRWST Maximum Water Temperature

The maximum IRWST water temperature following a postulated break in containment is used to calculate the net positive suction head (NPSH) of emergency core cooling system (ECCS) pumps as well as to verify that the IRWST water remains subcooled during the entire transient.

Comparing the IRWST water temperature for LOCAs and MSLBs, as shown in Figure B-19A, the maximum IRWST water temperature occurs during a LOCA event. Therefore, the impact of each of the containment’s initial conditions on IRWST water temperature is determined using the limiting LOCA event for maximum IRWST water temperature (DEDLSB with minimum SI flow).

As documented in Table C-1D, containment initial conditions of minimum pressure, maximum temperature, and maximum relative humidity result in the highest IRWST water temperature. In the peak temperature sensitivity analysis documented above, initial conditions of minimum pressure and maximum temperature lead to increased containment temperature, whereas higher relative humidity increases the heat capacity of the containment atmosphere and increases energy transfer to the IRWST from condensate on the containment walls. The combined effects maximize the IRWST water temperature following the LOCA event.

Figure C-1C shows the calculated IRWST water temperature transients under the various combinations of initial condition values listed in Table C-1D.

C.2 Comparison of Nodalization for Containment P/T and IRWST

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C.3 Sensitivity Analysis for Duration of Droplet Discharge Phase separation of the LOCA break fluid can impact the containment pressure and temperature analysis. In general, the assumption of droplets for break flow is limited to the period when the liquid, after phase separation, produces continuous containment heating. Sensitivity analyses on the duration of droplet injection were performed to establish its impact on containment response. The drop diameter is assumed as 100 microns and sufficiently small to reach thermal equilibrium with the containment atmosphere before reaching the floor. Table C-3A lists four sensitivity cases addressing droplet injection during the blowdown and post-blowdown phases. The calculated containment peak pressure for each of the analyses is tabulated in Table C-3B.

Cases 1 and 2 produce identical peak pressures for all the LOCA events analyzed. The Case 1 and 2 peak pressures are also higher than the corresponding pressures for the Case 3 and 4 events. Case 4, which assumes droplets for hot and cold spillage during post-blowdown, results in a lower containment peak pressure than Case 1.

The results of this study confirm that droplet/steam discharge during blowdown provides atmospheric heating due to liquid superheat relative to the transient atmospheric conditions, while continuous discharge of the spillage as droplets following blowdown produces an atmospheric cooling effect. Therefore, the assumed combination of droplet/steam discharge during blowdown and liquid/stream discharge following blowdown is limiting for the containment response analysis to LOCA events.

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C.4 Sensitivity Analysis for Time Step Control The GOTHIC solver automatically adjusts the time step during the calculation, increasing the time step up to the user-specified maximum when the transient is progressing slowly and decreasing the time step during periods of rapid change.

The minimum time step should be set small enough to allow the solver to continue calculating during periods of rapid change. In the APR1400 containment analyses, a value of 10 micro (1.0e-5) seconds is chosen for the minimum time step.

For the maximum time step, an appropriate value is determined by GOTHIC considering the rate of change of key parameters as well as the Courant limit for numerical stability. The APR1400 containment model additionally restricts the maximum allowable time steps to maintain an accurate and stable solution. These maximum allowable time steps are confirmed from the results of time step sensitivity studies

Tables C-4A and C-4B show comparisons of containment peak pressure and temperature obtained for various maximum time steps using the limiting LOCA event. Note that maximum time step sensitivity on the peak pressure and temperature becomes apparent at a maximum time step greater than 0.01 second. Therefore, the maximum time step size shall be limited to 0.01 second until the containment peak pressure and temperature are reached.

In the APR1400 containment analyses, the maximum time step is limited to 0.01 second or less until 600 seconds after accident initiation. As indicated in Tables B-1G and B-2L, this maximum time step is maintained throughout the transient period prior to reaching peak conditions for both the LOCA and MSLB events. To reduce computation time, a maximum time step of 0.1 second is used from 600 to 1,000 seconds and 1.0 second thereafter. Table C-4C shows the minimum and maximum time step sizes applied in the APR1400 containment analyses.

C.5 Parametric Study for Containment Spray Effectiveness

A parametric study is performed to estimate the minimum fall height of the spray droplets required to produce 100 percent containment spray effectiveness. Thirteen cases including combinations of droplet diameter, spray flow rate, containment temperature, and spray water temperature are considered.

As documented in Subsection A.2.2.3, the mean droplet diameter is estimated as 294 microns using the empirical data of plant-specific nozzles. A diameter of 1,000 microns is conservatively chosen for the APR1400 analysis. From the results of the parametric study as documented in Table C-5, the parameters yielding the largest fall height or worst case are as follows:

(1) Minimum containment atmosphere temperature

(2) Minimum droplet temperature

(3) Maximum spray flow rate

The definition of spray thermal effectiveness (ε) is the ratio of heat transferred from the containment atmosphere to the subcooled spray water, to that heat transfer corresponding to thermal equilibrium.

ε = hsf − hsihca − hsi

Where,

hsf: spray enthalpy after being heated by the containment atmosphere

hsi: spray water enthalpy at nozzles


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hca: enthalpy of the spray water at thermal equilibrium with the containment atmosphere

From Table C-5, it is observed that the spray effectiveness approaches 100 percent before the spray droplets reach a level of less than 40 ft from the spray nozzles. Since the minimum spray fall height in the APR1400 containment building is 105.4 ft (distance to top the pressurizer wall from the average nozzle height), the CSS spray effectiveness is assumed as 100 percent.


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Table C-1A. Containment Analysis Bounding Initial Conditions

Input Parameter Unit Min. Value Max. Value

Pressure psia 14.18 16.12

Temperature °F 50.0 120.0

Relative Humidity % 0.0 100.0

Table C-1B. LOCA Bounding Initial Conditions for Peak Pressure

Event Case

Min. or Max. (Bounding values) Peak Pr.

(psia) Temp.

(°F) Time (sec) Pressure Temp. Humidity

LOCA (l-03) (1)

1 max max min 65.79 274.3 323.8

2 min max min 63.26 274.2 323.8

3 max min min 60.48 263.8 17.3

4 max max max 64.96 276.0 323.8

LOCA limiting case (maximum containment pressure, DEDLSB with max. SI flow)

Table C-1C. MSLB Bounding Initial Conditions for Peak Temperature

Event Case

Min. or Max. (Bounding values) Pressure

(psia) Peak T.

(°F) Time (sec) Pressure Temp. Humidity

MSLB (m-02) (1)

1 min max min 59.32 333.41 111.7

2 max max min 61.78 329.6 106.0

3 min min min 52.41 293.2 112.1

4 min max max 58.56 333.3 114.2

(1) MSLB limiting case (maximum containment temp., 102 % power with an MSIV single failure)

Table C-1D. LOCA Bounding Initial Conditions for Maximum IRWST Temperature

Event Case

Min. or Max. (Bounding values) Max. Water Temp. (°F)

Time (sec) Pressure Temp. Humidity

LOCA (l-04) (1)

1 min max max 246.47 16,007

2 max max max 246.37 16,107

3 min min max 220.09 23,512

4 min max min 245.44 16,307

LOCA limiting case (Maximum IRWST water temperature, DEDLSB with min. SI flow)


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Table C-2A. Comparison of Volume Size for Containment Models

Table C-2B. Results of Noding Structure Sensitivity Analysis

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Table C-3A. Duration of Droplet Discharge

Table C-3B. Peak Pressure Results vs. Duration of Droplet Discharge

Accident

Peak Pressure (psia)

Case 1 Case 2 Case 3 Case 4

DESLSB Max. SI 64.659 64.659 64.659 64.616

Min. SI 64.824 64.824 64.824 64.791

DEDLSB Max. SI 65.792 65.792 65.785 62.536

Min. SI 65.667 65.667 65.661 63.585

DEHLSB Max. SI 62.980 62.980 62.980 62.980

Table C-4A. Peak Pressures vs. Various Time Step Sizes

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Table C-4B. Peak Temperatures vs. Various Time Step Sizes

Table C-4C. Time Step Size Applied to Each Time Domain

Time Int.

Time Steps (sec) End Time seconds △T min △T max

1

0.00001

0.0001 0 < time ≤ 0.1

2 0.0001 0.1 < time ≤ 1.0

3 0.0010 1.0 < time ≤ 10.0

4 0.0100 10.0 < time ≤ 600.0

5 0.1000 600.0 < time ≤ 1,000.0

6 1.0000 1,000.0 < time ≤ End

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Table C-5. List of Parameters for Containment Spray Effectiveness

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10-1 100 101 102 103 104 10510.0

20.0

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40.0

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70.0LOCA - DEDLSB with Max. SI Flow

Case 3

Case 2

Case 1 (65.79 psia)

Case 4

PRES

SURE

(PSI

A)

TIME (SECOND)

Figure C-1A. Pressure Transients to Various Initial Conditions (LOCA)


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MSLB - 102% Power with MSIV Single-failure

Case 3

Case 2

Case 4

Case 1 (333.41 oF)

Figure C-1B. Temperature Transients to Various Initial Conditions (MSLB)


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Case 2

Max. IRWST Water Temperature

LOCA - DEDLSB with Min. SI Flow

Case 4Case 3

TEM

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Case 1 (246.47 oF)

Figure C-1C. IRWST Water Temperature Transient to Various Initial Conditions (LOCA)


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Figure C-2A. Comparison of Noding Structure (Single Node / Two Nodes)

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0 100 200 300 400 500 600 700 800 900 100050.0

55.0

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65.0

70.0

Pmax=65.79 psia(4.626 kgf/cm2A)

LOCA - DEDLSB with Max. SI Flow

Cont

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(psia

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Time (sec)

Two-node Modeling

Single-node ModelingPmax=65.46 psia(4.602 kgf/cm2A)

Figure C-2B. Comparison of Containment Pressure (LOCA – DEDLSB with Max. SI)


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0 100 200 300 400 500 600 700 800 900 1000200.0

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350.0

400.0

Tmax=331.14 oF (166.19 oC)

Tmax=333.41 oF (167.45 oC)

MSLB - 102% Power, 9.134 ft2, MSIV Single-Failure

Cont

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empe

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Time (sec)

Two-node Modeling

Single-node Modeling

Figure C-2C. Comparison of Containment Temperature (MSLB – 102 Percent Power, MSIV Failure)


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0.1 1 10 100 1000 10000 100000100.0

120.0

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220.0

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260.0

Two-node Modeling


Tmax=247.07 oF (119.48 oC)

Tmax=247.19 oF (119.55 oC)

LOCA - DEDLSB with Max. SI Flow

IRW

ST W

ater

Tem

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Two-node Modeling


Figure C-2D. Comparison of IRWST Water Temperature (LOCA – DEDLSB with Max. SI)


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Appendix D Containment External Pressure Analysis for Inadvertent Operation of Containment Spray


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TABLE OF CONTENTS

D. CONTAINMENT EXTERNAL PRESSURE ANALYSIS FOR INADVERTANT OPERATION OF CONTAINMENT SPRAY ............................................................. 1

D.1 Assumptions............................................................................................................................ 1 D.2 Initial Conditions ..................................................................................................................... 1 D.3 Analysis Method ...................................................................................................................... 2


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LIST OF TABLES

Table D-1 Initial Conditions for Containment External Pressure Analysis ......................................... 4 Table D-2 Calculation Results of Containment External Pressure Analysis ...................................... 4


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D. CONTAINMENT EXTERNAL PRESSURE ANALYSIS FOR INADVERTENT OPERATION OF CONTAINMENT SPRAY

This appendix documents the APR1400 containment external pressure loading analysis. The analysis determines the containment external pressure load and demonstrates that the containment building and related structures are designed to accommodate the maximum external pressure resulting from an inadvertent operation of containment heat removal systems such as the spray system, purge system, and fan cooler system.

The containment pressure reduction resulting by an inadvertent actuation of the reactor containment fan cooler (RCFC) units is deemed negligible because the lowest RCFC cooling water temperature is higher than the IRWST minimum water temperature assumed for containment sprays. During an inadvertent operation of the purge system (i.e., operation of the exhaust train with the supply train isolated), the maximum feasible internal vacuum is limited to a few inches of water (gauge) based on the exhaust fan design characteristic curve.

The decrease in containment internal pressure from an inadvertent operation of the CSS with the containment purge valves open is negligible. However, a significant containment pressure reduction can occur following CS spray actuation into a sealed containment with all the purge valves closed. Consequently, the worst postulated event for the determination of maximum containment external pressure is the inadvertent operation of the CSS.

The calculation results presented in this section are consistent with DCD Subsection 6.2.1.1.

D.1 Assumptions

In calculating the maximum containment external pressure load, an analytical model based on the ideal gas equation and Dalton’s partial pressure law is used. Computerized transient calculations are not necessary since the absolute minimum containment pressure is of interest and the transient response is not required. The following conservative assumptions are used to simplify the analytical model:

● The air and steam in containment reach thermal equilibrium at the containment spray liquid droplets temperature.

● The containment is completely isolated at the time of spray actuation, that is, all of the purge valves are closed and leakage into containment is not considered.

● There is no heat transfer from/to the containment structures and no volume reduction from the addition of spray water.

● All heat sources within containment are disregarded.

D.2 Initial Conditions

The key initial conditions for this analysis include pressure, temperature, relative humidity, and spray water temperature. The minimum initial pressure and maximum initial temperature minimize the containment steam and air mass at the start of the transient. A maximum relative humidity decreases the initial ratio of non-condensing gases to steam. The minimum spray water temperature, which determines the containment atmosphere temperature after spray actuation, produces higher negative pressure. The assumed initial conditions for the containment external pressure analysis are listed in Table D-1.


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D.3 Analysis Method The approach used to determine the pressure reduction as a result of the inadvertent spray actuation is as follows:

● Determine the initial containment total pressure as a sum of air and steam partial pressure.

● Calculate the final containment air partial pressure (after spray actuation) using the ideal gas law and the corresponding equation of state.

● Add the containment steam pressure (corresponding to the spray water temperature) to the final air partial pressure to obtain the final containment total pressure.

● Compare the initial and final containment total pressures to determine the pressure difference across the containment wall, which acts as the containment external pressure load.

The containment pressure reduction from the subcooled spray water injection can be represented by the following equation:

The APR1400 containment is designed to withstand an external pressure loading of 0.281 kgf/cm2 (4.0 psi) relative to ambient pressure. Therefore, it is demonstrated that the design external pressure load on the containment provides more than a 10 percent margin (11.43 % [1 – 3.54/4.0]) in accordance with the

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requirements of Section 6.2.1.1.A of NUREG-0800. Table D-2 presents all of the calculated values for each parameter in the above equations.


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Table D-1. Initial Conditions for Containment External Pressure Analysis

Parameter Initial value

Initial pressure 0.997 kgf/cm2A (14.18 psia) (1)

Initial temperature 48.9 °C (120 °F)

Relative humidity 100 %

Spray droplet temperature 10 °C (50 °F)

(1) The minimum pressure is determined from the value, 14.7- 0.1 (LCO min. value) – 0.42 (indicator uncertainty) psia.

Table D-2. Calculation Results of Containment External Pressure Analysis

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Appendix E Conclusions


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E. CONCLUSIONS

Appendices A through C describe the GOTHIC containment analysis methodology used to determine the containment pressure and temperature response to a spectrum of high-energy pipe breaks in the containment building. The methodology used to calculate the LOCA M/E release during the long-term boil-off phase is also described. This long-term LOCA M/E release is calculated within the GOTHIC code using an RCS/SGs model.

The APR1400 GOTHIC containment model was conservatively used in bounding LOCA and MSLB postulated accidents and the results shown to be bounded by the APR1400 containment design limits with sufficient margin.

This documented APR1400 GOTHIC containment analysis methodology is deemed conservative and appropriate for the determination of containment response to postulated breaks in containment including the calculation of maximum containment pressure and temperature and maximum IRWST water temperature following postulated LOCA and MSLB events inside containment.


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Appendix F Regulatory Requirements

for Containment Response Analysis


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F. REGULATORY REQUIREMENTS FOR CONTAINMENT RESPONSE ANALYSIS

This section compares the proposed APR1400 containment response analysis methodology to NUREG-0800 (Reference 1) requirements and ANSI/ANS-56.4 (Reference 4) guidelines. A summary of the APR1400 containment analysis methodology with respect to the corresponding requirement/guideline is outlined in the following table.

SRP 6.2.1.1.A PWR Dry Containments, Acceptance Criteria

No. Requirements APR1400 Methodology

1 GDC 16 and GDC 50 To satisfy the requirements of GDC 16 and 50 regarding sufficient design margin, the containment design pressure should provide at least a 10 % margin above the accepted peak calculated containment pressure following a loss-of-coolant accident, or a steam or feedwater line break.

2 GDC 38 To satisfy the requirements of GDC 38 to rapidly reduce the containment pressure, the containment pressure should be reduced to less than 50 % of the peak calculated pressure for the design basis loss-of-coolant accident within 24 hours after the postulated accident.

3 GDC 38 and GDC 50 To satisfy the requirements of GDC 38 and 50 with respect to the containment heat removal capability and design margin, the loss-of-coolant accident analysis should be based on the assumption of loss of offsite power and the most severe single failure in the emergency power system (e.g., a diesel generator failure), the containment heat removal systems (e.g., a fan, pump, or valve failure), or the core cooling systems (e.g., a pump or valve failure). The selection made should result in the highest calculated containment pressure.

4 GDC 38 and GDC 50 To satisfy the requirements of GDC 38 and 50 with respect to the containment heat removal capability and design margin, the containment response analysis for postulated secondary system pipe ruptures should be based on the most severe single active failure in the containment heat removal systems (e.g., a fan, pump, or valve failure) or the secondary system isolation provisions (e.g., main steam isolation valve failure or feedwater line isolation valve failure). The analysis should also be based on a spectrum of pipe break sizes and reactor power levels. The accident conditions selected should result in the highest calculated containment pressure or temperature.

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ANS 56.4-1983 Requirements for Containment Integrity Analysis

Subsection Requirements APR1400 Methodology

4.2.1 Postulated Accidents The analysis shall include a spectrum of break sizes, locations, and power levels as required to ensure that the breaks yielding the maximum pressure and temperature transients are identified.

4.2.2 Duration of Analysis Each analysis shall be carried out for a sufficient duration to ensure that the maximum pressure and temperature have been ascertained.

For the case yielding highest maximum pressure, the analysis shall be extended to demonstrate that the vapor region pressure is reduced below 50 % of the peak calculated pressure within 24 hours and maintained below this pressure for the duration of the accident.

4.2.3 Dry Primary Containment Analysis Model The model shall be based on a solution of the conservation equations for M/E for the dry primary containment system. In developing the containment model, two distinct regions, the atmosphere and sump regions, are typically identified to exist in the dry primary containment volume and are, therefore, identified as such in the analytical model.

4.2.3.1.1 Dry Containment Atmosphere Region The steam and non-condensable components making up this region shall be considered to be homogeneously mixed and in thermal equilibrium with each other.

The existence of water droplets in the dry containment atmosphere may be included if the treatment of their thermodynamic and mechanical behavior is justified

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4.2.3.1.2 Dry Containment Sump Region The sump region is defined as that volume of accumulated water on the dry primary containment floor. This region has no vapor space. The sump region may be assumed to be in pressure equilibrium with the containment atmosphere region.

4.2.3.2.1 Pipe Break Blowdown As a minimum, the M/E release from the break shall be assumed to go directly to the containment atmosphere region for subsequent distribution, except that portion defined as spillage mass immediately directed to the sump region. For that portion of M/E released to the vapor region, a phase-separation model shall be used. This model shall produce a steam addition rate at least as large as that which is computed using the assumption of flashing to the saturation temperature at transient dry primary containment atmosphere steam partial pressure.

4.2.3.2.2 Energy Source Terms Consideration shall be given to all credible energy sources not previously accounted for in the generation of M/E release data for their influence on atmosphere region pressure and temperature.

4.2.3.2.3 Structural Heat Transfer The structural heat sinks are to be modeled conservatively with a lower-bound estimate of the number and heat transfer area.

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The rate of heat transfer between the dry primary containment regions and their structural heat sinks is determined by the physical arrangement of the conducting masses, their thermal properties, surface properties, heat transfer coefficients, and heat sink boundary conditions. Within a heat sink, the temperature profile shall be determined by an appropriate solution of the transient heat conduction equation. An appropriate value of contact resistance will be modeled where distinct material layers interface within the heat sink.

The thermal properties of the heat sink materials shall be chosen to provide a conservatively low estimate of thermal capacitance and transmission capabilities.

Convection, condensation, and (where appropriate) radiation heat transfer will be addressed at the heat sink surfaces.

4.2.3.2.4 Containment Spray System Energy removal from the primary containment vapor region due to sprays may be included. If the atmosphere is saturated, the sprays will no longer evaporate but the sprays can still absorb energy from the atmosphere. If the dry primary containment vapor region is superheated, the evaporation rate of the falling spray droplets may be assumed unlimited. Unevaporated spray water shall be assumed to go to the sump at a temperature in equilibrium with the containment vapor region. In computing the time-dependent behavior of the dry primary containment, the spray droplet residence time in the vapor region may be assumed to be zero.

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4.2.3.2.5 CHRS Energy Removal Terms Credit may be taken for removal of energy from the dry primary containment system by means of the containment heat removal system (CHRS) components. However, in such cases, consideration shall be given to the mechanism and efficiency of energy removal for each component and the effects incorporated in the modeling of this energy removal.

Sources of coolant for the various CHRS components shall be assumed to be at their highest credible temperature throughout the accident. A transient analysis may be performed to conservatively predict the source temperature and thereby relax this constraint.

Modeling of heat removal components shall include effects of fouling, condensate buildup or any conditions that may be expected to degrade the performance of the component.

4.2.3.2.6 Atmosphere – Sump Interface M/E transfer across the atmosphere-sump interface by evaporation need not be treated unless the sump region bulk temperature exceeds the saturation temperature at the total pressure of the atmosphere region. Mass and energy transfer across the atmosphere-sump interface shall be considered for this "pool boiling" case.

4.2.3.3 Modeling Considerations Selection of time step size and heat sink geometric nodalization shall be sufficient to ensure a physically representative solution that describes the dry primary containment system.

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4.2.4 Initial Conditions Initial conditions shall be chosen to yield a conservatively high peak containment atmosphere region pressure and temperature. In selecting the initial dry primary containment atmospheric conditions and structural temperatures, consideration shall be given to the competing effects of the initial air mass and the active and passive heat sink thermal capacities.

4.2.5 Single Failure Criteria Dry containment response analyses shall incorporate the effects of the most severe single failure. The failure chosen shall result in the highest calculated dry primary containment atmosphere pressure and temperature for the postulated break and shall be consistent with that chosen for the generation of mass and energy release data. In addition, the loss of all nonemergency electric power to the plant shall be postulated concurrent with the pipe break if such an occurrence yields more severe consequences.

4.4 Minimum Containment Pressure Analysis The minimum containment pressure analysis shall be performed to determine the worst-case negative pressure differential across the containment structure. The methods described for maximum pressure and temperature analysis are applicable to this analysis.

4.4.1 Duration of Analysis The analysis shall be of sufficient duration to ensure that the absolute minimum dry containment atmosphere pressure has been found.

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4.4.2 Initial Conditions For inadvertent spray actuation transients, the upper-bound initial containment atmosphere temperature and lower-bound initial containment pressure shall be used. For a dry containment that does not employ vacuum breaking devices, the initial relative humidity shall be chosen to produce the lowest dry containment atmosphere pressure.

4.4.3 Structural Heat Sinks In taking credit for the availability of structural heat sinks as an energy source, a lower-bound estimate of their number and surface area shall be chosen. In addition, a lower-bound surface heat transfer coefficient shall be used for these heat sinks.

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Appendix G References


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1. NUREG-0800, “Standard Review Plan,” Sections 6.2.1 and 6.2.1.1.A, “PWR Dry Containments, including Subatmospheric Containments,” Rev. 3, U.S. Nuclear Regulatory Commission, March 2007.

2. APR1400-K-X-FS-14002-P, “APR1400 Design Control Document Tier 2,” KEPCO and KHNP, December 2014.

3. ANSI/ANS-5.1-1979, “American National Standard for Decay Heat Power in Light-Water Reactors,” American Nuclear Society, August 1979.

4. ANS-56.4-1983, “Pressure and Temperature Transient Analysis for Light Water Reactor Containments,” American Nuclear Society, December 1983.

5. GOTHIC Thermal Hydraulic Analysis Package User Manual, Version 8.0 (QA), NAI 8907-02, Rev. 20, Numerical Applications Inc., January 2012.

6. GOTHIC Thermal Hydraulic Analysis Package Technical Manual, Version 8.0 (QA), NAI 8907-06, Rev. 19, Numerical Applications Inc., January 2012.

7. GOTHIC Containment Analysis Package Qualification Report, Version 8.0 (QA), NAI 8907-09, Rev. 12, Numerical Applications Inc., January 2012.

8. WCAP-16608-NP, “Westinghouse Containment Analysis Methodology,” Westinghouse Electric Company LLC, August 2006.

9. Final Safety Evaluation for FRAMATOME ANP Topical Report BAW-10252(P), Revision 0, “Analysis of Containment Response to Postulated Pipe Ruptures Using GOTHIC,” TAC NO. MC3783, August 31, 2005.

10. Schmitt, R. C., et al., “Simulated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment Final Report,” IN-1403, Idaho Nuclear Corporation, Idaho Falls, ID, 1970.

11. Letter, J. G. Lamb (U.S. NRC) to M. Reddeman (NMC), “Kewanee Nuclear Power Plant – Review for Kewanee Reload Safety Evaluation Methods Topical Report WPSRSEM-NP, Revision 3 (TAC No. MB0306),” September 10, 2001.

12. APR1400-E-N-NR-14001-P, Rev. 0 (Proprietary), “APR1400 Design Features to Address GSI-191,” 2014.

13. Gauld, I. C., Raap, B. Schmittroth, F., and Wilson, W.B., “Proposed Revision of the Decay Heat Standard ANSI/ANS-5.1-2005,” Oak Ridge National Laboratory, January 2010.

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