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PCTRAN ABWR VerificationAnalyses

May 2008

Micro-Simulation Technology10 Navajo Court

Montville, New Jersey 07045USA

http://www.microsimtech.com

[email protected]
http://www.microsimtech.com/http://www.microsimtech.com/


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Overview

PCTRAN is a reactor transient and accident simulation software program that operates on apersonal computer. Since its first release in 1985, Micro-Simulation Technology has beenconstantly upgrading its performance and expanding its capabilities. Numerous versions and plant

models have been installed in many countries around the world.

Advancement in modern 32-bit microprocessors and the windows-based graphic user interface(GUI) has completely revolutionized the simulation technology. It is now possible to automatethe preparation work and actual exercise on a desktop computer. Since 1998, the source code ofPCTRAN has been converted into Microsoft Visual Basic 6.0. Operation of the GUI adheresstrictly to the specifications of the Microsoft Windows environment. Data input/output are in MSOffices Access database format. Reports and data can be transferred conveniently through allWindows-based software products over the entire exercise network.

The plant model is the General Electric (GE) designed Advanced Boiling Water Reactor(ABWR) plant of 1300 MWe. The Standard ABWR Final Safety Analysis Report (SFAR) and

Taiwan Power Companys Fourth Nuclear Power Plant Lungmen (Dragon Gate) Project wasused for PCTRAN-ABWR model input. Chapter 1 is a description of the Lungmen plant systemscovered in this simulation system. Noticeable differences between the Standard ABWR andLungmen Project are marked. For example, Lungmen is designed to sustain a turbine trip or loadrejection without a reactor trip when sufficient bypass valves open. In addition, the feedwatersystem has both turbine-driven and motor-driven pumps; and there is a specific loading patternduring a feedwater pump trip transient. The second chapter includes instructions for operating theWindows version. There is an on-line "Help" button in the menu bar that provides instantinstructions. The verification benchmark analysis is in Chapter 4. Next, the mathematical theory isdescribed in Chapter 5.

Upon successful completion of the NSSS model simulation, it formulates a solid foundation andconfidence in further expansion. Chapter 6 expands into the following areas:

- Radiological release pathways and source terms for offsite dose dispersion- Development of emergency planning drill scenarios- Severe accident phenomenon study and management

By covering both plant operations and dose release, PCTRAN-Windows formulates a completesimulation system for emergency planning. It will continue to improve and expand whenapplication and experience accumulate in the industry.

Version 4.0 released May 2002 has data editing capability and real time/transient time display inthe bottom status bar. These are powerful tool in conducting analysis and emergency exercise.

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Table of Contents

1.0 Overview

2.0 ABWR Plant Systems

2.1 Reactor Internal Pumps2.2 Drywell and Suppression Pool2.3 Safety Relief Valves and the Automatic Depressurization System (ADS)2.4 Reactor Protection System2.5 Reactor Coolant Isolation Cooling (RCIC)2.6 Low Pressure Flooder System2.7 High Pressure Flooder System2.8 Reactor Dome Pressure and Turbine Header Pressure Control2.9 Feedwater System2.10 Containment Over-pressure Protection System

3.0 Operations of the Windows Graphic User Interface

3.1 General Description3.2 Installation3.3 Help File3.4 Input Files and Parameters3.5 Initial Condition Files3.6 Changing Malfunction Status3.7 Component Operation3.8 Post-Plot Utility

4.0 Benchmark Analysis

4.1 Run-out of One Feedwater Pump4.2 Turbine Trip (All Bypass Valves Available)4.3 Turbine Trip (Failure of All Bypass Valves)4.4 Closure of All Main Steam Isolation Valves (Direct Scram)4.5 Trip of all Reactor Internal Pumps4.6 Fast Runout of All Reactor Internal Pumps4.7 ATWS MSIV Closure with ARI4.8 HPCF Cold Water Injection4.9 ATWS Loss of Feedwater, FMCRD Run-in

4.10 ATWS Turbine Trip with Bypass, SLCS Initiation4.11 Loss of Coolant Accidents4.11.1 Maximum Steam Line Break Inside Containment4.11.2 Feedwater Line Break4.11.3 High Pressure Core Flooder Line Break4.11.4 Maximum Steam Line Break Outside Containment

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2.0 ABWR Plant Systems

The plant model is the General Electric (GE) designed ABWR plant of 1300 MWe. Thefollowing systems have been modeled in the simulation system:

2.1 Reactor Internal Pumps

Ten RIPs are separated into two groups: four will be tripped on reactor trip; the remaining six(which are tied to a MG set) will be tripped on low reactor water level L2 or high dome pressureset point. During normal operation the operator can manually set the RPM to change the coreflow.Load change can be controlled by combination of recirculation pumps master RPM demand androd position control. The following table represents the RIP operation strategy:

RIP Trip Configuration

RIPs with M/G set (6) RIPs without M/G Set (4)_________________ ____________________

FW Pump trip Run back Run back

RX Water L3 - Trip

RX Water L2 3 trip at 0 sec -trip at 6 sec

TCV fast closure/ - TripTSV closure TBVFailure

High dome press - Trip(1140 Asia)

TCV fast closure/ Run back Run backTSV closure________________________________________________________________

2.2 Drywell and Suppression Pool

The upper and lower drywell is modeled in the containment model. The wetwell air space andsuppression pool is properly modeled. Horizontal vents connect the drywell and wetwell. Thereare vacuum breakers to balance the pressures of the two compartments after a major blowdown

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2.3 Safety Relief Valves and the Automatic Depressurization System (ADS)

Eighteen Safety/relief valves (SRVs) are modeled in either the relief mode or the safety mode.Users have control to switch from one to another. In the relief mode there are six opening andclosing bands, and in the safety mode there are five bands. For Automatic DepressurizationSystem (ADS) actuation, eight valves will be opened at reactor water level below L1 (15.3 cmTAF) and high drywell pressure signal (1.25 bar) with 29 seconds delay or L1 only with 480seconds delay.

2.4 Reactor Protection System

The following RPS signals have been modeled:

Reactor low water level L3 = 3.75 meters TAF (Top of Active Fuel)High reactor dome pressure = 78.7 barMSIV closureHigh flux (125% of nominal or flux-flow map)

For a malfunction of anticipated transient without scram (ATWS), reactor trip will be bypassedfor the above conditions. Then four of the RIPs will be tripped and the Alternate Rod Insertion(ARI) will be started with an insertion time of 25 seconds. The ARI button in the RPS panel willbe lit automatically upon its initiation. In the event of ARI is also lost for the malfunction, theoperator can choose (manual action) either to drive in the rods by Fine Motion Control Rod Drive(FMCRD) run-in in135 seconds or start the Standby Liquid Control (SLC) boron injectionpumps. The remaining six RIPs will be tripped on L2 condition. It further reduces the reactorpower for ATWS mitigation.

2.5 Reactor Coolant Isolation Cooling (RCIC)

The Reactor Coolant Isolation Cooling (RCIC) system will be initiated at (2.43m TAF) or highdrywell pressure with 29 seconds time delay. The flow is a constant 38.8 Kg/sec for the reactorpressure between 10.55 to 82.75 bars.

2.6 Low Pressure Flooder System

Two of the 3 RHR pumps in LPFL model will be initiated automatically on L1 or high drywellpressure. Suction will be from the suppression pool and discharge into the reactor vessel. Thepumps flow rate is provided by a head curve provided in the FSAR. The discharge valve to thereactor vessel will be opened at 35.2 bar to allow water entering the vessel. The pumps will beautomatically shut if the reactor water level recovers above L8 (4.91m). During an extremeemergency or exercise when the suppression pool cannot provide water, operator can manuallyline up suction from the plant fire water system.

The same RHR pumps are used for normal shutdown cooling, containment spray and wetwellcooling. They are initiated manually either by pressing the mode switch in the RHR panel or bycontrolling the individual pumps and valves for a given mode. For shutdown cooling, water isdrawn from the reactor vessel and cooled by the RHR heat exchangers before returning back to

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the vessel. This can only be conducted when the reactor pressure is below 9.2 bars. Forcontainment spray or suppression pool cooling mode, water is drawn from the suppression pooland running through the RHR heat exchangers. The operator for spray mode operation opensseparates spray discharge valves into the drywell and wetwell air space. For the suppressionpool-cooling mode, the water is returned into the suppression pool

2.7 High Pressure Flooder System

Two HPFL pumps will be initiated automatically on L1.5 = 0.987m TAF with 36 seconds timedelay. The pumps flow rate is provided by the FSAR. The pumps will be stopped on reactorlevel above L8.

2.8 Reactor Dome Pressure and Turbine Header Pressure Control

Turbine control valves are set automatically to control the reactor dome pressure to the set point.After a turbine trip that is indicated by closing of the turbine stop valve, the bypass valves to thecondenser will be opened to adjust the turbine header pressure. Their malfunction for stuck open

or close can be simulated by either entering into the Malfunction List or by failing the component.

2.9 Feedwater SystemThe feedwater system is controlled by pump speed so that the reactor water level is maintained ata set value according to the three-element control logic. The three elements are feed flow,steam flow and reactor water level. Feedwater pump run-out can be accommodated by a pumpmalfunction of greater than 100% rated flow. The other normal pump will still follow the controllogic and mitigate the transient. This has been successfully modelled to handle feedwater tran-sients.

There are three turbine-driven pumps and one motor-driven pump. During normal operation twoturbine-driven pumps are running at 50% each with the third one in standby. The motor-drivenpump has a capacity of about 25% of total capacity and is used for start-up. In the event of oneof the operating turbine-driven pump trip, the motor-driven pump will be started automatically.After the combined feedwater flow reaches approximately 75% and the reactor power controlsystem adjusts it accordingly, the third turbine-driven pump will be started and the motor-drivenpump dropped.

2.10 Containment Over-pressure Protection System

Two rupture disks have been modelled for over-pressure protection. They will be broken uponthe suppression pool air space pressure over 7 bars absolute.

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Figure 3.1 PCTRAN/ABWR Windows Mimic

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3.2 Installation

To install PCTRAN/ABWR for Windows (total of 5 disks), please follow the following steps:

1) Your computer has to be running the Microsoft Window 95, 98, 2000 or NT operating

system. Windows 3.x or OS/2 will not work.

2) The MS Office 97 Applications Access and Excel have to be installed in your computer.

3) Your monitor should be set to a resolution of 1024 x 768 SVGA with small fonts. If not,the full PCTRAN mimic will not be displayed. To reset the monitor, you can go to MyComputer, Control Panel, Display, Settings and choose 1024 x 768 for the DesktopArea and a Font Size of Small Fonts. It may take a few seconds to reset your screen orrequire rebooting Windows 95 (or 98).

4) Close all unnecessary Windows programs running in the background, e.g. those for viruschecking, LAN connection, etc. Since they may cause conflict during installation of the DLL

files of PCTRAN.

5) To install PCTRAN, place Disk 1 in the disk drive and run Setup.exe using the Runcommand. The screen display will prompt you for the succeeding steps. The defaultdirectory is C:\Program Files\Pctran. You can change it to any name you like. Theexecutable program PCTRAN will be displayed in the Start Menu. You can make a short cutor copy it to your desktop using Windows 95 (or 98).

6) After starting PCTRAN, the MST banner will be displayed with the background of theGeneral Electric 1300 MWE ABWR mimic. Click on the MST banner to close it. You areready to run RCTRAN.

7) To get a complete tutorial, we suggest you to click on the Help button in the tool bar. Acomplete sequence of instructions will be displayed by clicking by the >> key.

3.3 Help File

Software in Windows is usually equipped with a Help feature that provides over-allinstructions. They are organized in hyper-linked pages. Keep pressing the >> button, thecomplete instructions for Windows operation is provided in Appendix 1.

3.6 Changing Malfunction Status

Malfunctions can be turned on by going through Code Control and Malfunctions in the topmenu bar, or by clicking at the lower left No Malfunction status button. The list is thendisplayed. After entering the ramp time, delay time and severity, the status should be checked to"Active" and close the window. Up to twenty malfunctions are displayed:

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3.7 Component Operation

All of the components on the two mimic diagrams are functional. Red indicates OPEN or ON,green indicates CLOSED or OFF.

Pumps are started and stopped by pressing the left mouse button when the cursor is on the pump.Pumps can be locked or failed to a specific state by pressing the right mouse button when thecursor is on the pump symbol. When this happens a Red square is placed around the pump. Thepump state can then be set to ON or OFF by operating the pump as discussed above to simulatecontrol signal failures.

Valves are opened and closed by pressing the left mouse button when the cursor is on the valve.Valves can be failed to any position from 0-100% open by pressing the right mouse button whenthe cursor is on the valve symbol. When this happen a window pops up to allow the user to enterthe failure fraction for the valve. An active valve malfunction is shown by a Red rectangle aroundthe valve symbol. Three buttons are displayed in the window:

Failure Fraction is the percent of full open at which the valve is failed. The valuecan be set to any integer from 0-100.

Status is the current state of the failure. Pressing this button toggles thestate between Inactive and Active.

Close is used to close the equipment failure window to accept the failurefraction and malfunction state.

A Green box displays a heat exchanger with a Red pipe coil passing through it. Pressing the rightmouse button when the mouse cursor is over the symbol can reduce heat removal by the heat

exchanger. Selection of the failure fraction is the same as discussed above for valve failures.When a heat exchanger malfunction is active a Red box replaces the Green box.

3.8 Post-Plot Utility

At completion of a run, the user can get the transient report in text form and the transientvariables in either Excel or Access form. Table 3.1 is the transient report for a pump trip event.After saving the plot in Access, the user can use the plotting utility "PostPlot.exe" to generatepre-achieved data. Just click at PostPlot and open the saved data, the title, variable selection,scales, range, curve thickness and data interval are all adjustable. There are even selection ofCartesian and semi-log scales. The latter is appropriate for radiological data since wide range ofvariation is expected. This gives great flexibility in output for report preparation. The user canuse query to select the NSSS variables in PlotData.mdb database and radiological source term inDoseData.mdb database to write reports. A sample one for dose release is attached.

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ID Label Units Name

39 Concentration Containment H2 by Vol (%) % CNH2

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ID Label Units Name

71 Concentration DW I-131 Eq (uCi/cc) uCi/cc DW131

70 Concentration DW Kr-87 Eq (uCi/cc) uCi/cc DW87

66 Concentration RC I-131 Eq (uCi/gm) uCi/gm RC131

67 Concentration RC Kr-87 Eq (uCi/gm) uCi/gm RC87

68 Concentration WW I-131 Eq (uCi/cc) uCi/cc WW131

69 Concentration WW Kr-87 Eq (uCi/cc) uCi/cc WW87

81 Dose Rate EAB Thyroid (mSv/hr) mSv/hr DTHY

82 Dose Rate EAB Whole Body (mSv/hr) mSv/hr DWB

87Dose Rate LPZ Thyroid (mSv/hr) mSv DTHL

88 Dose Rate LPZ Whole Body (mSv/hr) mSv DWBL

65Enthalpy RCS Break (kJ/kg) J/gm HBK

10Flow Bypass Valve (%) % BPFL

58Flow Containment Leakage (kg/s) kg/s WCLK

13Flow Core Inlet (%) % COFL

49Flow Ctmt Overpressure Protection (kg/s) kg/s WDWV

48 Flow Drywell to Wetwell Drain (kg/s) kg/s WDWD

61 Flow Drywell to Wetwell Vent Air (kg/s) kg/s WVTA

60Flow Drywell to Wetwell Vent Vapor (kg/s) kg/s WVTG

5Flow Feedwater (%) % FWFL

62Flow Low Pressure Injection (kg/s) kg/s WLPI

56Flow RCIC (kg/s) kg/s WRIC

57Flow RCS Break (kg/s) kg/s WBK

9Flow Relief Valve (%) % RVFL

44Flow Standby Liquid Control (kg/s) kg/s WSLC

46Flow Total Drywell Spray (kg/s) kg/s WDWS

41Flow Total Feed (kg/s) kg/s WMFW

55Flow Total High Pressure Flooding (kg/s) kg/s WHPI

14Flow Total Pump (%) % PMPF

42Flow Total Recirculation (kg/s) kg/s WREC

54Flow Total RHR Pumps (kg/s) kg/s WRHR

40Flow Total Steam (kg/s) kg/s WSTM

43 Flow Total Steam Relief Valves (kg/s) kg/s WVP

47Flow Total Wetwell Spray (kg/s) kg/s WWWS

11Flow Turbine Steam (%) % TBFL

6Flow Vessel Steam (%) % VSTF

4Flux Average Surface Heat (%) % ASHX

2Flux Neutron (%) % NTFX

26Fraction Core Uncovery - UNC

37Fraction of Clad Failure (%) % FRCL

63Heat Removal RHR Heat Exchanger (MW) MW ESDC

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ID Label Units Name

83 Integrated Dose EAB Thyroid (mSv) mSv DTHYI

84 Integrated Dose EAB Whole Body (mSv) mSv DWBI

85 Integrated Dose LPZ Thyroid (mSv) mSv DTHLI

86 Integrated Dose LPZ Whole Body (mSv) mSv DWBLI

12Level (Ref-Separator Skirt) (in) in LVIN

64Level Boron Injection Tank (M3) m^3 TBIJ

24Level Condensation Pool Water (M) m LWW

23Level Drywell Water (M) m LDW

21 Level Reactor Water Downcomer (cm) cm LEV

22 Level Reactor Water Two-Phase (cm) cm LCOR

38 Mass Total H2 Generated by Zr-H2O (kg) kg MH2

25Mass Total Liquid in Rx Vessel (ton) ton TOTM

45Position Rod Withdrawn (%) % RDPOS

8Pressue Turbine Press Rise (psi) psi TBPR

52Pressure Drywell (kg/cm2) bar PDW

50Pressure Reactor Dome (kg/cm2) bar P

51Pressure Turbine Header (kg/cm2) bar PHDR

7Pressure Vessel Press Rise (psi) psi VSPR

53Pressure Wetwell (kg/cm2) bar PWW

72Rad Monitor Lower DW (p/s) p/s RMLDW

75Rad Monitor Steam Line N16 (p/s) p/s RMSL

76Rad Monitor Turbine Bldg (p/s) p/s RMTB

73Rad Monitor Upper DW (p/s) p/s RMUDW

74Rad Monitor Wet Well (p/s) p/s RMWW

59Ratio Change in Critical Power (%) % DCPR

19Reactivity Boron ($) dollar BORK

16Reactivity Doppler ($) dollar DPRK

17Reactivity Scram ($) dollar SCRK

18Reactivity Total ($) dollar TLRK

15Reactivity Void ($) dollar VDRK

79 Release Rate RX Bldg Iodine (uCi/s) uCi/s STTBI

80Release Rate RX Bldg NG (uCi/s) uCi/s STTBG

77Release Rate Turbine Bldg Iodine (uCi/s) uCi/s STRBI

78 Release Rate Turbine Bldg NG (uCi/s) uCi/s STRBG

28 Temperatue Saturation in Rx Vessel (C) C TSAT

31Temperature Average Fuel (C) C TF

35Temperature Condensation Pool (WW) (C) C TWW

33Temperature Lower Drywell (C) C TDW

3 Temperature Peak Centerline Fuel (%) % PFCT

32Temperature Peak Clad (C) C TPCT

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ID Label Units Name

30Temperature Peak Fuel (C) C TFPK

27 Temperature Recirculation Loop Average(C)

C TAVG

29Temperature Submerged Fuel Average (C) C TFSB

34Temperature Upper Drywell (C) C TDWU36Temperature Wetwell Gas Space (C) C TWWG

1Time (sec) sec TIME

20Void Fraction Core Average (%) % VOID

_________________________________________________________________________

Table 3.1 Transient Report

Reset to IC #1000002.5 sec, 4 Reactor Internal Pumps Trip000003.5 sec, Turb Bypass Valve #1 Position Change: 100%000008.5 sec, 6 Reactor Internal Pumps Trip000013.5 sec, Turb Bypass Valve #1 Position Change: 0%000013.5 sec, Reactor Trip Control Change: True000014.0 sec, Reactor Manual Scram000014.0 sec, Reactor Trip000014.5 sec, Turb Bypass Valve #1 Position Change: 100%000016.0 sec, Turb Bypass Valve #1 Position Change: 0%000016.5 sec, Turb Bypass Valve #1 Position Change: 100%000017.5 sec, Turb Bypass Valve #1 Position Change: 0%000018.0 sec, Turbine Trip000018.5 sec, Turb Stop Valve Position Change: 0%

000018.5 sec, Turb Bypass Valve #1 Position Change: 100%000020.5 sec, Safety Relief Valve #0 Position Change: 100%000024.0 sec, Reactor Water Level Below L3: 374.41 CM000024.0 sec, Cleanup Valve Position Change: 0%000026.0 sec, Safety Relief Valve #0 Position Change: 0%___________________________________________________________________________

The transient parameters calculated by ABWR are written into the PlotData Access database. GEtypically groups 4 or 5 parameters in dimensionless percentage of its nominal full power values.For instance, the first five: neutron flux, peak fuel centerline temperature, average surface heatflux, feedwater flow and vessel steam flow are always displayed in the same frame for figures in

Chapter 15 Accident Analysis. For the ABWR Project, the metric units are supposed to be used.But however, conventional British units such as psi and inches still appear in their documentation.They are described as follows:

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In the selected transient runs, the figure numbers in the Standard FSAR were used for closecomparison. Each case is presented with a brief description. A table is used for the initial condition(IC) and malfunction (MF) selection and manual interaction steps. The key transient parameters ofinterest and the run time are also listed. Transient plots comparable to the original study followthem. No attempt is made to fine-tune the input parameters to be completely agreeable with theoriginal case, since that would affect other cases and lose general consistency. Qualitative ratherthan quantitative consistency is thus the over-all objective of this simulator. The selected cases are:- Runout of One Feedwater Pump (Figure 15.1-2)- Turbine Trip (All Bypass Valves Available, Figure 15.2-6)- Turbine Trip (Failure of All Bypass Valves, Figure 15.2-8)- Closure of All Main Steam Isolation Valves (Direct Scram, Figure 15.2-9)- Trip of all Reactor Internal Pumps (Figure 15.3-2)- Fast Runout of All Reactor Internal Pumps (Figure 15.4-3)- ATWS MSIV Closure with ARI (Figure 15E-2)- ATWS Loss of Feedwater, FMCRD Run-in (Figure 15E-10)- ATWS Turbine Trip with Bypass, SLCS Initiation (Figures 15E-19)

- Maximum Steam Line Break Inside Containment (Figures 6.3-12 to 6.3-20)- Feedwater Line Break (Figures 6.3-21 to 6.3-29)- High Pressure Core Flooder Line Break (Figures 6.3-44 to 6.3-52)- Maximum Steam Line Break Outside Containment (Figures 6.3-60 to 6.3-75)

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4.1 Runout of One Feedwater Pump

Section 15.1.2.3.2.1 (page 15.1-6) and Table 15.1-4 of GE Standard FSAR for ABWR describesthis case. At time zero the simulated runout of one feedwater pump flow is 75% of ratedfeedwater flow. This corresponds to 75/50 or 125% of the pumps normal flow. Therefore, wewill initialize this transient by right mouse click at one of the feedwater pump and enter 125%malfunction value. The feedwater controller starts to command the remaining feedwater pump toreduce its flow immediately. The vessel water level increases slightly and then settles back to itsnormal level. The vessel pressure and MCPR change very little for this transient.

Note, the feedwater system of Lungmen is significantly different from the standard ABWR. Asdescribed in Section 1.9, there are three turbine-driven pumps and one motor-driven pump. Dur-ing normal operation two turbine-driven pumps are running at 50% each with the third one instandby. The motor-driven pump has a capacity of about 25% of total capacity and is used forstart-up. In the event of one of the operating turbine-driven pump trip, the motor-driven pumpwill be started automatically. After the combined feedwater flow reaches approximately 75% andthe reactor power control system adjusts itself accordingly, the third turbine-driven pump will be

started and the motor-driven pump dropped.

Run Time = 20 seconds

Initial Condition(IC #)

Malfunction(#/%)

InteractiveControl(panel/%)

OutputVariable(Name)

Range(min/max/unit)

1 NA Right mouseclick at one ofthe MFW pumpand enter 125%

malfunctionfraction

NTFXPFCTASHXFWFL

VSTFVSPRTBPRRVFLBPFLTBFLLVINCOFLPMFPVDRKDPRK

RDRKTLRK

0/180(%)

-25/150(PSI)

0/180(INCH)0/180(%)0/180(%)-200/200(CENT)

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19

FSAR Figure 15.1.2 (Runout 1 FW Pump PCTRAN Results)

0

20

40

60

80

100

120

140

160

0 5 10 15 20 25 30

Time (sec)

%Rated

NTFX

PFCT

ASHX

FWFL

VSTF


-200

-150

-100

-50

0

50

100

150

200

0 5 10 15 20 25 30

Time (sec)

Reactivity(CENT)

VDRK

DPRK

SCRK

TLRK


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20


-25

0

25

50

75

100

125

0 5 10 15 20 25 30

Time (sec)

%o

fRatedFlow

-30

-10

10

30

50

70

90

110

PressureRise(PSI)

RVFL

BPFL

TBFL

VSPR

TBPR


0

25

50

75

100

125

150

0 5 10 15 20 25 30

Time (sec)

%o

fRatedFlow

0

20

40

60

80

100

120

140

160

Level(inches)

COFL

PMPF

LVIN


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4.2 Turbine Trip with Bypass

In Section 15.2.3.2.3 of Standard ABWR, a turbine trip with all bypass valves available issimulated as shown in Figure 15.2-6 and Table 15.2-6.

At full power, reactor scram is bypassed if no more than two of the ten bypass valves fail to openwithin 0.2 seconds. For lower operating power there is staggered trip bypass logic according tothe available bypass valves.

Following a turbine trip, the neutron flux increases by the stop valve closure. With all bypassvalves available, the combined 110% steam relief capacity may prevent a reactor trip. For theanalyzed case, the reactor is scrammed on high dome pressure. PCTRAN calculated results arevery comparable to the Standard ABWR FSAR.


Initial Condition

(IC #)

Malfunction

(#/%)

Interactive

Control(panel/%)

Output

Variable(Name)

Range

(min/max/unit)

1 9 None NTFXPFCTASHXFWFLVSTFVSPRTBPRRVFLBPFL

TBFLLVINCOFLPMFPVDRKDPRKRDRKTLRK

0/180(%)

-25/150(PSI)

0/180(INCH)0/180(%)0/180(%)-200/200(CENT)

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4.3 Turbine Trip (Failure of All Bypass Valves)

Section 15.2.3.2.3, a turbine trip with failure of the bypass system is simulated as shown in Figure15.2-8 and Table 15.2-8. Neutron flux increase by the stop valve scram and the trip of fourRIPs. With the bypass valve failure, the pressure relief system opens more valves and stays openlonger time to relieve the pressure. The maximum pressure is also higher than the previous case.PCTRAN/ABWR calculated results are very comparable to the PSAR.



Malfunction(#/%)



Range(min/max/unit)

1 10 None NTFXPFCTASHXFWFL

VSTFVSPRTBPRRVFLBPFLTBFLLVINCOFLPMFPVDRKDPRK

RDRKTLRK

0/180(%)

-25/150(PSI)

0/180(INCH)0/180(%)0/180(%)-200/200(CENT)

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27

FSAR Figure 15.2-8 (Turbine Trip w/o Bypass PCTRAN Results)

0

20

40

60

80

100

120

0 5 10 15 20 25 30 35

Time (sec)

%Rated

NTFX

PFCT

ASHX

FWFL

VSTF


-25

0

25

50

75

100

125

0 5 10 15 20 25 30 35

Time (sec)

%o

fRatedFlow

-30

-10

10

30

50

70

90

110

PressureRise(PSI)

RVFL

BPFL

TBFL

VSPR

TBPR


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28


0

25

50

75

100

125

150

0 5 10 15 20 25 30 35

Time (sec)

%o

fRatedFlow

0

20

40

60

80

100

120

140

160

Level(inches)

COFL

PMPF

LVIN


-200

-150

-100

-50

0

50

100

150

200

0 5 10 15 20 25 30 35

Time (sec)

Reactivity(CENT)

VDRK

DPRK

SCRK

TLRK


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4.4 Closure of All Main Steam Isolation Valves (Direct Scram)

Section 15.2.4.4.1, closure of all MSIVs initiates a direct scram. High reactor pressure signaltripped four RIPs. Safety/relief valves open due to high pressure. Vessel water level reaches L2set point and RCIC is initiated. The analysis is shown in Figure 15.2-9 and Table 15.2-9.PCTRAN/ABWR calculated results are very comparable to the PSAR.



Malfunction(#/%)



Range(min/max/unit)

1 6 None NTFXPFCTASHXFWFLVSTF

VSPRTBPRRVFLBPFLTBFLLVINCOFLPMFPVDRKDPRKRDRK

TLRK

0/180(%)

-25/150(PSI)

0/180(INCH)0/180(%)0/180(%)-200/200(CENT)

29


30/118

30


31/118

31


32/118

32


33/118

33


34/118

4.5 Trip of all Reactor Internal Pumps

Section 15.3.1.3.2.2 and Figure 15.3-2 demonstrates this event. This event is initiated by clickingat both sides RIPs. In order to get fine resolution for the short transient, plot interval is changedto 1 second in the Options. The reactor is tripped on high flux-flow. After the turbine tripfollowing the reactor trip, the bypass valves open to relieve up to 40% or nominal full powersteam as designed. Since we have more SRVs available than the PSAR has assumed, the reactorand turbine pressure rise is not as high as the PSARs result. PCTRAN-ABWR shows about 40 60 psi (0.3 0.5 Mpa) while PSAR shows over 0.8 Mpa increase. All other parameters are inclose agreement



Malfunction(#/%)



Range(min/max/unit)

1 6 None NTFXPFCTASHXFWFLVSTFVSPRTBPRRVFLBPFLTBFLLVIN

COFLPMFPVDRKDPRKRDRKTLRK

0/180(%)

-25/150(PSI)

0/180(INCH)

0/180(%)0/180(%)-200/200(CENT)

34


35/118

35


36/118

0

10

20

30

40

50

60

70

80

90

100

110

120

0 10 20 30

Time (sec)

ABWR Trip All Recircualtion Pumps Fig 15.3-2

PCTran3/27/01 2:34:31 PM

Flux Average Surface Heat (%)Flux Neutron (%)

0

10

20

30

40

50

60

70

80

90

100

110

120

0 10 20 30

Time (sec)


PCTran3/27/01 2:35:35 PM

Flow Feedwater (%)Flow Vessel Steam (%)

36


37/118

0

10

20

30

40

50

60

70

80

90

100

110

120

0 10 20 30

Time (sec)


PCTran3/27/01 2:36:18 PM

Pressue Turbine Press Rise (psi)Pressure Vessel Press Rise (psi)

0

10

20

30

40

50

60

70

80

90

100

110

120

0 10 20 30

Time (sec)


PCTran3/27/01 2:38:10 PM

Flow Bypass Valve (%)Flow Relief Valve (%)Flow Turbine Steam (%)

37


38/118

0

10

20

30

40

50

60

70

80

90

100

110

120

0 10 20 30

Time (sec)


PCTran3/27/01 2:39:14 PM

Flow Core Inlet (%)Flow Total Pump (%)

-4

-3

-2

-1

0

1

2

3

4

10 20 30

Time (sec)


PCTran3/27/01 2:39:55 PM

Reactivity Doppler ($)Reactivity Scram ($)Reactivity Total ($)Reactivity Void ($)

38


39/118

4.6 Fast Runout of All Reactor Internal PumpsSection 15.4.5.3.2.2 describes the fast runout of all RIPs with a maximum speed increase rate.This case must be initiated with the reactor operating at less than full power and flow condition.In the PCTRAN/ABWR programs Initial Condition List, IC9 for 60% power and 50% flow.MF7 is selected. Then select MF7 for the pump speed change with the pump runout fraction150% being entered. Table 14.4-5 shows that the reactor APRM high flux scram is initiatedfollowed by turbine trip. The reactor water level drops below L3 and four RIPs are tripped.PCTRAN/ABWR closely reproduces all these.



Malfunction(#/%)



Range(min/max/unit)

9 7/150% None NTFX

PFCTASHXFWFLVSTFVSPRTBPRRVFLBPFLTBFLLVINCOFL

PMFPVDRKDPRKRDRKTLRK

0/180(%)

-25/150(PSI)

0/180(INCH)0/180(%)

0/180(%)-200/200(CENT)

39


40/118

40


41/118

41

FSAR Figure 15.4-3 (Fast Runout of All RIPs PCTRAN Results)

0

10

20

30

40

50

60

70

80

90

0 10 20 30 40 50 60 70 80

Time (sec)

%Rated

NTFX

PFCT

ASHX

FWFL

VSTF


-25

0

25

50

75

100

125

0 10 20 30 40 50 60 70 80

Time (sec)

%o

fRatedFlow

-30

-10

10

30

50

70

90

110

PressureRise(PSI)

RVFL

BPFL

TBFL

VSPR

TBPR


42/118

42


0

25

50

75

100

125

150

0 10 20 30 40 50 60 70 80

Time (sec)

%o

fRatedFlow

0

20

40

60

80

100

120

140

160

Level(inches)

COFL

PMPF

LVIN


-2

-1.5

-1

-0.5

0

0.5

1

1.5

2

0 10 20 30 40 50 60 70 80

Time (sec)

Reactivity($)

VDRK

DPRK

SCRK

TLRK


43/118

4.7 HPCF Cold Water InjectionSection 15.5.1.3.1 describes HPCF system cold water at 4.4C injection. In PCTRAN/ABWR wehave to prepare an IC corresponding to this cold ECCS temperature. IC19 was duplicated fromIC1 with TECCS set to 4.4. After initiation of the run one HPCF pump is manually started andits isolation valve opened. The flow takes 36 seconds to start as assumed in the systemspecifications, although in this section it was assumed to start immediately. The flow rate at 70Kg/s is 3.2% of the full steam flow.

Addition of cold water causes a reduction in steam flow, which results in some depressurizationas the pressure regulator responds to the event. The flux level settles out slightly below operatinglevel. No significant variation is resulted. PCTRAN/ABWR closely reproduces all these.



Malfunction(#/%)

InteractiveControl

(panel/%)

OutputVariable

(Name)

Range(min/max/

unit)1 None Click on HPCF

pump

Click on valvefrom CST

NTFXPFCTASHXFWFLVSTFVSPRTBPRRVFLBPFLTBFL

LVINCOFLPMFPVDRKDPRKRDRKTLRKWMFWWHPF

0/180(%)

-25/150(PSI)

0/180(INCH)0/180(%)0/180(%)-200/200(CENT)

0/2400(KG/S)

43


44/118

44


45/118

45

FSAR Figure 15.5-1 (HPCF Cold Water Injection PCTRAN Results)

95

96

97

98

99

100

101

102

0 50 100 150 200 250 300 350

Time (sec)

%Rated

NTFX

PFCT

ASHX

FWFL

VSTF


-25

0

25

50

75

100

125

0 50 100 150 200 250 300 350

Time (sec)

%o

fRatedFlow

-30

-10

10

30

50

70

90

110

PressureRise(PSI)

RVFL

BPFL

TBFL

VSPR

TBPR


46/118

46


0

25

50

75

100

125

150

0 50 100 150 200 250 300 350

Time (sec)

%o

fRatedFlow

0

20

40

60

80

100

120

140

160

Level(inches)

COFL

PMPF

LVIN


-2

-1.5

-1

-0.5

0

0.5

1

1.5

2

0 50 100 150 200 250 300 350

Time (sec)

Reactivity($)

VDRK

DPRK

SCRK

TLRK


47/118

47


0

10

20

30

40

50

60

70

80

0 50 100 150 200 250 300 350

Time (sec)

Flow

(kg/sec)

WHPI


48/118

4.8 ATWS MSIV Closure with ARI

Section 15 of PSAR discussed Anticipated Transient without Scram (ATWS). The first case is anATWS caused by MSIV closure. Scram signal paths are assumed to fail include valve position, highneutron flux, high vessel pressure and all manual attempts. A short time after the MSIVs have closedcompletely, the ATWS high pressure setpoint is reached, which initiates four of the ten recirculationpumps to trip and the rest start to runback. The combined effect of the trip and runback reduces thecore flow and increases core voids, thereby reducing power generation that limits pressure increaseand steam discharge to the suppression pool. The ATWS high-pressure signal causes the actuationof the ARI and the electric insertion of the FMCRD. In this case only the ARI is assumed with theinsertion time equal to 25 seconds. Peak values of key parameters are shown in Table 15E-4 and thetransient results are shown in Figures 15E-2. The PCTRAN/ABWR calculated values are closelycomparable to the PSAR results. There are a few noticeable differences: PCTRANs maximumneutron flux around 150% is lower than the PSARs 450%, the dome pressure does not decreaseafter the reactor becomes subcritical. Also, the feedwater is not isolated. Not all the PSAR analysisassumptions are well documented. This may cause the discrepancies. However, the all-all conclusionin peak power, pressure, and fuel temperature not exceeding the design limits is consistent with the

PSAR.

48


49/118



Malfunction(#/%)



Range(min/max/unit)

1 8/0 (for ARI) Close MSIV NTFXPFCTASHXFWFLVSTFVSPRRVFLBPFLTBFLLVINCOFL

PMFPVDRKDPRKRDRKTLRKPTWWTPKT

0/180(%)

-25/150(PSI)

0/180(INCH)0/180(%)

0/180(%)-200/200(CENT)

0/90(Bar)

0/100(C)

0/900(C)

49


50/118

50


51/118

51


52/118

52


53/118

53


54/118

54

FSAR Figure 15E-2 (ATWS MSIV Closure with ARI PCTRAN Results)

0

20

40

60

80

100

120

140

160

0 50 100 150 200 250 300

Time (sec)

%Rated

NTFX

PFCT

ASHX

FWFL

VSTF


-25

0

25

50

75

100

125

0 50 100 150 200 250 300

Time (sec)

%o

fRatedFlow

-30

-10

10

30

50

70

90

110

PressureRise(PSI)

RVFL

BPFL

TBFL

VSPR

TBPR


55/118

55


0

25

50

75

100

125

150

0 50 100 150 200 250 300

Time (sec)

%o

fRatedFlow

0

20

40

60

80

100

120

140

160

Level(inches)

COFL

PMPF

LVIN


-20

-15

-10

-5

0

5

10

15

20

0 50 100 150 200 250 300

Time (sec)

Reactivity($)

VDRK

DPRK

SCRK

TLRK


56/118

56


73

74

75

76

77

78

79

80

81

82

83

0 50 100 150 200 250 300

Time (sec)

Pressure(kg/cm^2)

P


0

100

200

300

400

500

600

700

800

0 50 100 150 200 250 300

Time (sec)

Temperature(degC)

TFPK

TPCT


57/118

57


34

35

36

37

38

39

40

41

0 50 100 150 200 250 300

Time (sec)

Temperature(degC)

TWW


58/118

4.9 ATWS Loss of Feedwater, FMCRD Run-in

This event is simulated by selecting MF8 with flag=2 for disabling ARI and allowing FMCRD run-inby manual action. Then MF5 for loss of feedwater is entered (or the feedwater isolation valve in themimic could be manually isolated). The results are compared to the PSAR Table 15E-11 andFigures 15E-10 a to d.

After the loss of the feedwater has taken place, the pressure, water level and neutron flux begin tofall. Shortly the low water level L3 set point is reached. This trips four recirculation pumps. WhenL2 is reached, it trips remaining six pumps, activates ARI (bypassed for this case) and FMCRD run-in. Run-in takes 135 seconds and the reactor is brought to sub-critical. Both analyses are consistentto this point.

The rod reactivity calculated by PCTRAN/ABWR shows that all rods are driven in by FMCRD in135 seconds, consistent with the GE Standard FSAR Figure 15E-10c. Neutron flux becomes sub-critical. After L2 and ECCS initiation, the combined flow from RCIC and HPFC calculated byPCTRAN is less than the SRV, thus the reactor level keeps falling to L1. With a time delay of 480

seconds, the ADS start to depressurize the system. This is plausible but not shown in the GEanalysis.

Although there are differences between the two analyses, the over-all behavior of HPCF, RCICinitiation, recirculation pump trip, SRV cycling, FMCRD run-in and subcriticality are reproduced byPCTRAN/ABWR.

58


59/118



Malfunction(#/%)



Range(min/max/unit)

1 5 for LOFW

8/2 for ATWSno ARI

FMCRD click tostart run-in

NTFXPFCTASHXFWFLVSTFVSPRTBPRRVFLBPFLTBFLLVIN

COFLPMFPVDRKDPRKRDRKTLRKWMFWWHPFWRICWSRV

0/180(%)

-25/150(PSI)

0/180(INCH)

0/180(%)0/180(%)-200/200(CENT)

0/2400(KG/S)

59


60/118

60


61/118

61


62/118

62


63/118

63


64/118

64

FSAR Figure 15E-10 (ATWS LOFW FMCRD Runin PCTRAN Results)

0

20

40

60

80

100

120

0 200 400 600 800 1000 1200

Time (sec)

%Rated

NTFX

PFCT

ASHX

FWFL

VSTF


0

10

20

30

40

50

60

70

80

90

0 100 200 300 400 500 600 700 800

Time (sec)

Pressure(kg/cm^2)

P


65/118

65


0

500

1000

1500

2000

2500

0 200 400 600 800 1000 1200

Time (sec)

Flow

(kg/sec)

WHPI

WMFW

WRIC

WVP


0

100

200

300

400

500

600

0 200 400 600 800 1000 1200

Time (sec)

Level(cm)

LCOR

LEV


66/118

66


-40

-30

-20

-10

0

10

20

0 200 400 600 800 1000 1200

Time (sec)

Reactivity($)

VDRK

DPRK

SCRK

TLRK


67/118

4.10 ATWS Turbine Trip with Bypass, SLCS Initiation

Selecting MF8 with flag simulates this event=2 for disabling ARI and no FMCRD run-in by manualaction, followed by MF9 for a turbine trip with bypass. It is described in Section 15E.6.5 of theStandard FSAR. SRNM not downscale and the high-pressure signal will initiate the SLCS when thetimer runs up. According to 15E.6, automatic boron injection will be started with a 180-seconddelay, and boron will reach the core in 60 seconds. Since in our PCTRAN/ABWR model SLCS isnot automatic, we will do it manually at 240 seconds later by clicking on both SLC pumps. The flowrate is 100 gpm total (about 6 kg/s) of sodium pentaborate. The results are compared to Table15E-16 and Figures 15E-19 a to d.

After the turbine trip, steam is relieved from the bypass valves and SRVs. SLCS is started 240seconds afterwards. The negative reactivity brought by the borate solution turns the reactor intosub-critical within 15 minutes. This is consistent with the FSAR. In the GE analysis, sudden fluxsurges to greater than 200% was not explained. It was not reproduced by PCTRAN.

67


68/118



Malfunction(#/%)



Range(min/max/unit)

1 9 for TurbineTrip with bypass

8/2 for ATWSno ARI

Click at bothSLC pumps at240 secondsafter turbine trip

NTFXPFCTASHXFWFLVSTFVSPRTBPRRVFLBPFLTBFLLVIN

COFLPMFPVDRKDPRKRDRKBORKTLRKWMFWWHPFWRICWSRV

WSLC

0/180(%)

-25/150(PSI)

0/180(INCH)

0/180(%)0/180(%)-200/200(CENT)

0/2400(KG/S)

0/10(KG/S)

68


69/118

69


70/118

70


71/118

71


72/118

72


73/118

73

FSAR Figure 15E-19 (ATWS TB trip with SLCS PCTRAN Results)

0

20

40

60

80

100

120

0 200 400 600 800 1000 1200

Time (sec)

%Rated

NTFX

PFCT

ASHX

FWFL

VSTF


0

100

200

300

400

500

600

700

800

0 200 400 600 800 1000 1200

Time (sec)

Level(cm)

LCOR

LEV


74/118

74


-3

-2

-1

0

1

2

3

4

5

6

0 200 400 600 800 1000 1200

Time (sec)

Reactivity($)

VDRK

DPRK

SCRK

TLRK


0

1

2

3

4

5

6

0 200 400 600 800 1000 1200

Time (sec)

Flow

(kg/sec)

WSLC


75/118

75


0

20

40

60

80

100

120

140

0 200 400 600 800 1000 1200

Time (sec)

Temperature(degC)

TWW


73

74

75

76

77

78

79

80

0 200 400 600 800 1000 1200

Time (sec)

Pressure(kg/cm^2)

P


76/118

76


-25

0

25

50

75

100

125

0 200 400 600 800 1000 1200

Time (sec)

%o

fRatedFlow

-30

-10

10

30

50

70

90

110

PressureRise(PSI)

RVFL

BPFL

TBFL

VSPR

TBPR


77/118

4.11 Loss of Coolant Accidents

Section 6 of the ABWR Standard FSAR analyzed maximum steam line break inside and outsidecontainment, feedwater line breaks, and breaks of ECCS lines. The key parameters are listed in thefollowing table:Transient Variable PCTRAN/ABWR Parameter (unit)____________________________________ ______________________________1) Core flow COFL (%)2) Maximum critical power ratio DCPR (%) *3) Water level in the fuel channel LVCR (cm-TAF)4) Water level in the shroud LVCR (cm-TAF)5) Water level outside the shroud LEV (cm-TAF)6) Vessel pressure P (bar-abs)7) Flows out of the vessel WSTM, WSRV, WBK (kg/s)8) Flows into the vessel WFW, WHPF, WRIC, WLPF (kg/s)9) Peak cladding temperature TPCT (C)

10) Drywell and wetwell pressure PDW, PWW (bar)11) Drywell, Wetwell air and liquid temperature TDW, TWWG, TWW (C)

* CPR in GE figures is in dimensionless ratio, i.e. normal steady state value is about 1.13, whileDPCR is the delta increment in percent from its original value

Their corresponding parameters calculated by PCTRAN/ABWR are listed in the right column.

77


78/118

4.11.1 Maximum Steamline Break Inside Containment

According to Table 6.3-4, the break size is 985 cm2 for maximum double-ended break inside thedrywell. The single failure criteria limit the ECCS combination to 1 HPCF, RCIC, 2 RHR/LPFLand 8 ADS valves. The analyzed results are shown in the following figures. They are verycomparable to Figures 6.2-12 to 6.2-14 of ABWR FSAR for the containment pressure andtemperature, and Figures 6.3-12 to 6.3-20 of Lungmen PSAR for the other parameters. A fewnoticeable comments are described:- The reactor water level minimum calculated by PCTRAN is slightly above L1 ADS is never

initiated. This may be caused by the lack of two-phase discharge from the steamline break.More water inventory and lower pressure in the reactor vessel prevents ADS initiation.Since the vessel is already depressurized, LPFL has no problem to be injected. So theconsequence is the same.

- The normalized core flow calculated by PCTRAN shows step decreases following first four-pump trip and the remaining six-pump trip. PCTRAN uses a large time step size of 0.5seconds, thus in the short time span the curve is not smooth.

-The peak clad temperature increase in the early calculated by PCTRAN is less than that in theLungmen PSAR. This is probably caused by the heat transfer coefficient used for two-phaseswell is larger than GEs method.

- Change of the critical power ratio is comparable to PSAR.- Suppression pool air space temperature rise during the initial blowdown is less than PSAR.- Other parameters such as steam flow, feed flow, HPCF, RCIC, LPFL and dome pressure are

almost identical.



Malfunction(#/%)

InteractiveControl

(panel/%)

OutputVariable

(Name)

Range(min/max/

unit)

1 2/985% for 985cm2 steamlinebreak insidecontainment

5 for loss offeedwater

Right mouseclick at one ofthe RHR pumpto disable single failure

Disable oneHPCF pumpalso for singlefailure

COFLDCPRLVCRLEVTPCTPWFWWSTMWSRVWRIC

WHPFWLPF

0/150(%)-100/100 (%)-400/500 (CM)

0/900 (C)0/90 (bar-abs)0/2400 (kg/s)

78


79/118

79


80/118

80


81/118

81


82/118

82


83/118

83


84/118

84


85/118

85

Max Steam Line Break in Containment 1HP+RCIC+2LP (PCTRAN Results)

0

0.5

1

1.5

2

2.5

3

0 100 200 300 400 500 600

Time (sec)

SteamGeneratorLevel(%)

PDW

PWW


0

20

40

60

80

100

120

140

0 100 200 300 400 500 600

Time (sec)

Temperature(degC)

TWW

TWWG

TDW


86/118

86


0

10

20

30

40

50

60

70

80

0 100 200 300 400 500 600

Time (sec)

Pressure(kg/cm^2)

P


0

100

200

300

400

500

600

700

800

0 100 200 300 400 500 600

Time (sec)

Temperature(degC)

TFPK

TPCT


87/118

87


0

100

200

300

400

500

600

700

800

0 100 200 300 400 500 600

Time (sec)

Level(cm)

LCOR

LEV


0

500

1000

1500

2000

2500

0 0.5 1 1.5 2 2.5

Time (sec)

Pressure(kg/cm^2)

WSTM

WMFW


88/118

88


0

100

200

300

400

500

600

0 100 200 300 400 500 600

Time (sec)

Flow

(kg/sec)

WHPI

WLPI

WRIC

WVP


0

25

50

75

100

125

150

0 100 200 300 400 500 600

Time (sec)

%o

fRatedCoreInletFlow

COFL


89/118

89


-200

0

200

400

600

800

1000

1200

1400

0 100 200 300 400 500 600

Time (sec)

ChangeinCriticalPowerRatio(%)

DCPR


90/118

90


0

100

200

300

400

500

600

700

800

900

1000

0 100 200 300 400 500 600

Time (sec)

BreakFlow

(kg/sec)

WBK


91/118

4.11.2 Feedwater Line Break

According to Table 6.3-4, the break size is 839 cm2 for maximum double-ended break. No RCICsince it also enters the feedwater line. Single failure criterion limits the ECCS combination to 1HPCF, 2 RHR/LPFL and 8 ADS valves. The analyzed results are shown in the following figures.They are very comparable to Figures 6.2-6 and 6.2-7 of ABWR FSAR for the containment pressureand temperature, and Figures 6.3-21 to 6.3-29 of Lungmen PSAR for the other parameters. A fewdiscrepancies are commented below:- The normalized core flow calculated by PCTRAN shows step decreases following first four

pump trip and the remaining six pump trip. PCTRAN uses a large time step size of 0.5seconds, thus in the short time span the curve is not smooth.

- The peak clad temperature increase in the early calculated by PCTRAN is less than that in theLungmen PSAR. This is probably caused by the heat transfer coefficient used for two-phaseswell larger than GEs value.

- Change of the critical power ratio is comparable to PSAR.- Suppression pool air space temperature rise during the initial blowdown is less than PSAR.

-Other parameters such as steam flow, feed flow, ADS, HPCF, LPFL and dome pressure arealmost identical.



Malfunction(#/%)



Range(min/max/unit)

1 4/839% for 839cm2 feedwater

line break


Right mouseclick at one of

the RHR pumpto disable single failure


COFLDCPR

LVCRLEVTPCTPWFWWSTMWSRVWRICWHPFWLPF

0/150(%)-100/100 (%)

-400/500 (CM)

0/900 (C)0/90 (bar-abs)0/2400 (kg/s)

91


92/118

92


93/118

93


94/118

94


95/118

95


96/118

96


97/118

97


98/118

0

20

40

60

80

100

120

0 100 200 300 400Time (sec)

ABWR PSAR Fig 6.3-21 Feedwater Line Break

PCTran4/2/2001 2:55:33 PM

COFL (%)

0

100

200

300

400

500

600

700

800

900

1000

0 100 200 300 400

Time (sec)


PCTran4/2/2001 2:49:18 PM

WBK (kg/s)

98


99/118

0

100

200

300

400

500

600

0 100 200 300 400Time (sec)


PCTran4/2/2001 2:51:05 PM

WLPI (kg/s)WHPI (kg/s)

0

100

200

300

400

500

600

700

800

0 100 200 300 400

Time (sec)

ABWR PSAR Figs 6.3-24 and 6.3-25 Feedwater Line Break

PCTran4/2/2001 2:42:06 PM

LEV (cm)LCOR (cm)

99


100/118


101/118

0

20

40

60

80

100

120

140

160

180

0 100 200 300 400Time (sec)

ABWR PSAR Fig 6.2-7 Feedw ater Line Break

PCTran4/2/2001 2:54:42 PM

TWW (C)TDWU (C)TWWG (C)

0

10

20

30

40

50

60

70

80

90

100

0 100 200 300 400

Time (sec)


PCTran4/2/2001 2:43:31 PM

P (bar)

101


102/118

4.11.3 High Pressure Core Flooder Line Break

According to Table 6.3-4, the break size is 92 cm2 for a HPCF line break. The HPCF will not beavailable due to the break. The available ECCS combination is then RCIC, 2 RHR/LPFL and 8ADS valves. The analyzed results are shown in the following figures. They are comparable toFigures 6.3-44 to 6.3-52 of Lungmen PSAR. A few noticeable comments are described below:- The HPCF line break is pretty small (92 cm2). The reactor repressurized after its initial

pressure drop. Both PCTRAN and PSAR observe this analyses.- The water level drop is more significant for the PSAR and thus resulted earlier L1 and ADS

initiation (about 200 seconds after break). PCTRAN shows ADS at a later time.- Maximum PCT occurred after ADS. Both analyses resulted lower than 400C PCT.- LPCF started after ADS initiation.


Initial Condition

(IC #)

Malfunction

(#/%)

Interactive

Control(panel/%)

Output

Variable(Name)

Range

(min/max/unit)

1 1/92% for 92cm2 HPCF linebreak



Disable bothHPCF pumps orisolate the valveto prevent

HPCF entering

COFLDCPRLVCRLEVTPCTPWFWWSTMWSRVWRICWHPFWLPF

0/150(%)-100/100 (%)-400/500 (CM)

0/900 (C)0/90 (bar-abs)0/2400 (kg/s)

102


103/118

103


104/118

104


105/118

105


106/118

106


107/118

107


108/118


109/118

109


110/118

110


111/118

111


112/118

4.11 4 Maximum Steam line Break Outside Containment

According to Table 6.3-4, the break size is 3939 cm2 for maximum double-ended break outside thedrywell. It differs from the one inside containment by closure of the MSIV that isolates the break.The single failure criteria limit the ECCS combination to 1 HPCF, RCIC, 2 RHR/LPFL and 8 ADSvalve. The analyzed results are shown in the following figures. They are comparable to Figures 6.3-60 to 6.3-75 of Lungmen PSAR. A few noticeable comments are described below:- Closure of the broken steam lines isolation valve stops coolant loss. PCTRAN isolated at a

lower pressure than the Lungmens PSAR. It took longer time for the reactor dome pressureto recover to the SRV open set point.

- SRV cycling pattern after pressure recovery is similar between the two analyses.- The normalized core flow calculated by PCTRAN shows step decreases following first four-

pump trip and the remaining six-pump trip. This was discussed before.- The peak clad temperature increase in the early calculated by PCTRAN is less than that in

the Lungmen PSAR. This is probably caused by the heat transfer coefficient used for two-phase swell is larger than GEs method.

-Change of the critical power ratio is comparable to PSAR.

- There are no change in containment pressure and temperature because break is outside thecontainment.

- Other parameters such as steam flow, feed flow, HPCF, RCIC, LPFL and dome pressure arealmost identical.



Malfunction(#/%)



Range(min/max/unit)

1 3/3939% for 3939 cm2steamline breakoutsidecontainment




COFLDCPRLVCRLEVTPCTPWFWWSTMWSRVWRICWHPFWLPF

0/150(%)-100/100 (%)-400/500 (CM)

0/900 (C)0/90 (bar-abs)0/2400 (kg/s)

112


113/118

113


114/118

114


115/118

115


116/118

0

100

200

300

400

500

600

700

800

0 100 200 300 400 500 600Time (sec)

ABWR PSAR Figs 6.3-70 and 6.3-71 Main Steam Line Break Outside Containment

PCTran4/2/2001 2:16:56 PM

LEV (cm)LCOR (cm)

0

100

200

300

400

0 100 200 300 400 500 600

Time (sec)

ABWR PSAR Fig 6.3-75 Main Steam Line Break Outside Containment

PCTran4/2/2001 3:02:00 PM

TPCT (C)

116


117/118

0

500

1000

1500

2000

0 100 200 300 400 500 600Time (sec)


PCTran4/2/2001 3:01:14 PM

WRIC (kg/s)WMFW (kg/s)

0

1000

2000

3000

0 100 200 300 400 500 600

Time (sec)


PCTran4/2/2001 3:00:04 PM

WBK (kg/s)WVP (kg/s)

117


118/118

0

10

20

30

40

50

60

70

80

90

100

0 100 200 300 400 500 600Time (sec)


PCTran4/2/2001 2:59:00 PM

P (bar)

pctran abwr.doc

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