NUREG/IA-0464 

International 

 Agreement Report

RELAP5/MOD3.3 Model Assessment 

and Hypothetical Accident Analysis of 

Kuosheng Nuclear Power Plant with 

SNAP Interface 

Prepared by: 

Chunkuan Shih*, Hao-Chun Chang*, Jong-Rong Wang*, Shao-Wen Chen*, Hao-Tzu Lin, Show-Chyuan

Chiang**, Tzu-Yao Yu**

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.

1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

*Institute of Nuclear Engineering and Science, National Tsing Hua University; Nuclear and New Energy

Education and Research Foundation

101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

**Department of Nuclear Safety, Taiwan Power Company

242, Section 3, Roosevelt Rd., Zhongzheng District, Taipei, Taiwan

K. Tien, NRC Project Manager

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2015

Date Published: March 2016

Prepared as part of 

The technical exchange and cooperation between the U.S. Nuclear Regulatory Commission and the Task Force for

Safety Critical Software in the Field of Nuclear Safety Research

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0464 

International 

 Agreement Report

RELAP5/MOD3.3 Model Assessment 

and Hypothetical Accident Analysis of 

Kuosheng Nuclear Power Plant with 

SNAP Interface 

Prepared by: 

Chunkuan Shih*, Hao-Chun Chang*, Jong-Rong Wang*, Shao-Wen Chen*, Hao-Tzu Lin, Show-Chyuan

Chiang**, Tzu-Yao Yu**

Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.

1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, Taiwan

*Institute of Nuclear Engineering and Science, National Tsing Hua University; Nuclear and New Energy

Education and Research Foundation

101 Section 2, Kuang Fu Rd., HsinChu, Taiwan

**Department of Nuclear Safety, Taiwan Power Company

242, Section 3, Roosevelt Rd., Zhongzheng District, Taipei, Taiwan

K. Tien, NRC Project Manager

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2015

Date Published: March 2016

Prepared as part of 

The technical exchange and cooperation between the U.S. Nuclear Regulatory Commission and the Task Force for

Safety Critical Software in the Field of Nuclear Safety Research

Published by

U.S. Nuclear Regulatory Commission

iii 

ABSTRACT

After the measurement uncertainty recapture power uprates, Kuosheng nuclear power plant 

(NPP) has been uprated the power from 2894 MWt to 2943 MWt. For these power upgrade, 

several analysis codes were applied to assess the safety of Kuosheng Nuclear Power Plant. In 

our group, there were a lot of effort on thermal hydraulic code, TRACE, had been done before. 

However, to enhance the reliability and confidence of these transient analyses, thermal hydraulic 

code, RELAP5/MOD3.3 will be applied in the future. The main work of this research is to 

establish a RELAP5/MOD3.3 model of Kuosheng NPP with SNAP interface. Model 

establishment of RELAP5 code is referred to the Final Safety Analysis Report (FSAR), training 

documents, and TRACE model which has been developed and verified before. After completing 

the model establishment, three startup test scenarios would be applied to the RELAP5 model. 

With comparing the startup test data and TRACE model analysis results, the applicability of 

RELAP5 model would be assessed. 

Recently, Taiwan Power Company is concerned in stretch power uprated plan and uprates the 

power to 3030 MWt. Before the stretch power uprates, several transient analysis should be done 

for ensuring that the power plant could maintain stability in higher power operating conditions. In 

this research, three overpressurization transient scenario including main steam isolation valves 

closure, turbine trip with bypass failure and load rejection with bypass failure would be performed. 

These analysis results would be compared with the TRACE model results which had been 

performed before. 
iv 

 
v 

FOREWORD 

Taiwan and the United States have signed an agreement on CAMP (Code Applications and 

Maintenance Program). INER (Institute of Nuclear Energy Research, Atomic Energy Council, 

R.O.C.) is the organization in Taiwan responsible for the application of TRACE and RELAP in 

thermal hydraulic safety analysis, for recording user’s experiences of it, and providing 

suggestions for its development. To meet this responsibility, the RELAP5/MOD3.3 model of 

Kuosheng nuclear power plant has been built. The startup test data and TRACE model analysis 

results were used to compare with the results of RELAP5/MOD3.3 model. 
vi 

 

 
vii 

CONTENTS 

Page

ABSTRACT ............................................................................................................................... iii 

FOREWORD .............................................................................................................................. v 

CONTENTS .............................................................................................................................. vii 

FIGURES ................................................................................................................................... ix 

TABLES .................................................................................................................................... xi 

EXECUTIVE SUMMARY ......................................................................................................... xiii 

ABBREVIATIONS .................................................................................................................... xv 

1. INTRODUCTION ................................................................................................................1-1 

2. MODEL ESTABLISHMENT ...............................................................................................2-1 

3. START-UP TEST RESULTS .............................................................................................3-1 

3.1 100% Load Rejection with Bypass Valve Open (LRWB) ............................................3-1 

3.2 96% Main Steam Line Isolation Valves Closure (MSIVC) ..........................................3-8

3.3 94% Feedwater Pumps Trip (94% FWPT) ............................................................... 3-14 

4. HYPOTHETICAL TRANSIENT RESULTS .........................................................................4-1 

4.1 Main Steam Line Isolation Valves Closure with Bypass Failure .................................4-1 

4.2 Turbine Trip with Bypass Failure ...............................................................................4-7 

4.3 Load Rejection with Bypass Failure ......................................................................... 4-13 

5. CONCLUSIONS .................................................................................................................5-1 

6. REFERENCES ...................................................................................................................6-1 

 
viii 

 

 
ix 

FIGURES

Page

Figure 1 Overview of the Kuosheng NPP RELAP5/MOD3.3 Model .................................... 2-2

Figure 2 Parameters of Reactor Kinetics in RELAP5 Model ................................................ 2-4

Figure 3 Density Reactivity Feedback Table of Reactor Kinetics ........................................ 2-5

Figure 4 Doppler Reactivity Table of Reactor Kinetics ........................................................ 2-5

Figure 5 Scram Reactivity Feedback Table ........................................................................ 2-6

Figure 6 Properties of Main Steam Isolation Valves ............................................................ 2-8

Figure 7 Properties of Turbine Stop Valves ........................................................................ 2-9

Figure 8 Properties of Turbine Control Valves .................................................................. 2-10

Figure 9 Time Dependent Volume Properties of the Main Steam Pipe Lines .................... 2-11

Figure 10 Properties of Safety/Relief Valves ...................................................................... 2-12

Figure 11 Details of Feedwater Time Dependent Volume ................................................... 2-14

Figure 12 Details of Time Dependent Junction of Feedwater Pumps .................................. 2-15

Figure 13 Recirculation Pumps Properties .......................................................................... 2-16

Figure 14 Properties of Recirculation Controlling Valves .................................................... 2-17

Figure 15 Jet Pumps Properties ......................................................................................... 2-18

Figure 16 RELAP5/MOD3.3 Animation Model .................................................................... 2-19

Figure 17 Controlling System of Turbine Control Valves in 100% LRWB Transient .............. 3-3

Figure 18 Controlling System of Turbine Bypass Valve in 100% LRWB Transient ................ 3-3

Figure 19 Controlling System of Recirculation Pumps Trip ................................................... 3-4

Figure 20 Controlling System of Safety Valves in 100% LRWB Transient ............................ 3-4

Figure 21 Steam Flow Variation during the 100% LRWB Transient ...................................... 3-6

Figure 22 Core Power Variation during the 100% LRWB Transient ...................................... 3-6

Figure 23 Dome Pressure Variation during the 100% LRWB Transient ................................ 3-7

Figure 24 Controlling System of Reactor Scram in 96% MSIVC Transient ......................... 3-10

Figure 25 Controlling System of Safety/Relief Valves in 96% MSIVC Transient ................. 3-10

Figure 26 Steam Flow Variation during the 96% MSIVC Transient ..................................... 3-12

Figure 27 Core Power Variation during the 96% MSIVC Transient ..................................... 3-12

Figure 28 Dome Pressure Variation during the 96% MSIVC Transient ............................... 3-13

Figure 29 Feedwater Flow Variation during the 96% MSIVC Transient............................... 3-13

Figure 30 Controlling System of Feedwater Pump Trip in 96% FWPT Transient ................ 3-16

Figure 31 Controlling System of Flow Control Valves in 96% FWPT Transient ................... 3-16

Figure 32 Feedwater Flow Variation during the 94% FWPT Transient ................................ 3-18

Figure 33 Narrow Range Water Level Variation during the 94% FWPT Transient .............. 3-18

Figure 34 Core Flow Variation during the 94% FWPT Transient ......................................... 3-19

Figure 35 Core Power Variation during the 94% FWPT Transient ...................................... 3-19

Figure 36 Controlling System of Reactor Scram in MSIVC Hypothetical Accident ................ 4-2

Figure 37 Controlling System of Safety Valves in MSIVC Hypothetical Accident .................. 4-2

Figure 38 Steam Flow Variation during the MSIVC Hypothetical Accident ............................ 4-5

Figure 39 Dome Pressure Variation during the MSIVC Hypothetical Accident ...................... 4-5

Figure 40 Core Power Variation during the MSIVC Hypothetical Accident ............................ 4-6

Figure 41 Controlling System of Reactor Scram in TTBF Hypothetical Accident ................... 4-8

Figure 42 Controlling System of Relief Valves in TTBF Hypothetical Accident ...................... 4-8

Figure 43 Steam Flow Variation during the TTBF Hypothetical Accident ............................ 4-11

Figure 44 Dome Pressure Variation during the TTBF Hypothetical Accident ...................... 4-11

Figure 45 Core Power Variation during the TTBF Hypothetical Accident ............................ 4-12

Figure 46 Controlling System of Reactor Scram in LRNB Hypothetical Accident ................ 4-14

Figure 47 Controlling System of Safety/Relief Valves in LRNB Hypothetical Accident ........ 4-14

Figure 48 Steam Flow Variation during the LRNB Hypothetical Accident ............................ 4-17

Figure 49 Dome Pressure Variation during the LRNB Hypothetical Accident ...................... 4-17

Figure 50 Core Power Variation during the LRNB Hypothetical Accident ............................ 4-18

Figure 51 Core Flow Variation during the LRNB Hypothetical Accident .............................. 4-18
x 

 
xi 

TABLES 

Page

Table 1 Initial Conditions of RELAP5 and Startup Test Data in 100% LRWB Transient ...... 3-2

Table 2 Event Sequence of 100% LRWB Transient ............................................................ 3-2

Table 3 Initial Conditions of RELAP5 and Startup Test Data in 96% MSIVC Transient ....... 3-9

Table 4 Event Sequence of 96% MSIVC Transient ............................................................. 3-9

Table 5 Initial Conditions of RELAP5 and Startup Test Data in 94% FWPT Transient ...... 3-15

Table 6 Event Sequence of 94% FWPT Transient ............................................................ 3-15

Table 7 Initial Conditions of TRACE and RELAP5 Models .................................................. 4-1

Table 8 Events Sequence and Comparison of MSIVC Hypothetical Accident ..................... 4-3

Table 9 Events Sequence and Comparison of TTNB Hypothetical Accident ....................... 4-9

Table 10 Events Sequence and Comparison of LRNB Hypothetical Accident ..................... 4-15
xii 

 
xiii 

EXECUTIVE SUMMARY

RELAP5/MOD3.3Patch04 code, which was developed for light water reactor (LWR) transient 

analysis at Idaho National Engineering Laboratory (INEL) for U.S. NRC, is applied in this 

research. This code is often performed to support rulemaking, licensing audit calculations, 

evaluation of accident, mitigation strategies, evaluation of operator guidelines, and experiment 

planning analysis. Same as other thermal hydraulic analysis codes, RELAP5/MOD3.3 is based 

on nonhomogeneous and nonequilibrium model for the two-phase system. However, 

calculations in this code will be solved by a fast, partially implicit numerical scheme to permit 

economical calculation of system transients. It can produce accurate transient analysis results in 

relatively short time. 

Symbolic Nuclear Analysis Package (SNAP) is an interface of NPP analysis codes which 

developed by U.S. NRC and Applied Programming Technology, Inc. Different from the traditional 

input deck in ASCII files, the graphical control blocks and thermal hydraulic connections make 

researches comprehend the whole power plant and control system more easily. Due to these 

advantages, the RELAP5/MOD3.3 model of Kuosheng NPP was developed with SNAP 

interface.

Kuosheng NPP is located on the northern coast of Taiwan. Its nuclear steam supply system is a 

type of BWR/6 designed and built by General Electric on a twin unit concept. Each unit includes 

two loops of recirculation piping and four main steam lines, with the thermal rated power of 

2894MWt. After the project of MUR for Kuosheng NPP, the operating power is 2943 MWt. Unit 1 

started SPU from Cycle 24 and Unit 2 started SPU from Cycle 23. The operating power will be 

3030 MWt. In this research, a RELAP5/MOD3.3 model of Kuosheng NPP is developed. Further, 

the model in this research is built up with the SNAP interface. The startup test data and TRACE 

model analysis results were used to compare with the results of RELAP5/MOD3.3 model for 

overpressurization transients and feedwater pump trip transient. The predictions of 

RELAP5/MOD3.3 were consistent with the data of startup test and TRACE roughly. It indicates 

that there is a respectable accuracy in Kuosheng NPP RELAP5/MOD3.3 model 
xiv 

 

 

 
xv 

ABBREVIATIONS

BWR Boiling-Water Reactor 

BPV Bypass valve 

FWPT Feedwater pump(s) Trip 

INER Institute of Nuclear Energy Research Atomic Energy Council, R.O.C. 

kg kilogram(s) 

kW kilowatt(s) 

LRWB Load Rejection with Bypass 

LRNB Load Rejection with Bypass Failure 

MPa Megapascal(s) 

MSIVC Main Steam Isolation Valves Closure 

MUR Measurement Uncertainty Recapture 

NPP Nuclear Power Plant 

NRC Nuclear Regulatory Commission 

Pa Pascal(s) 

psi pounds per square inch 

SNAP Symbolic Nuclear Analysis Program 

SPU Stretch Power Uprated 

SRV Safety/Relief Valves 

TBV Turbine Bypass Valve 

TCVC Turbine Control Valves Closure 

TRACE TRAC/RELAP Advanced Computational Engine

TSVC Turbine Stop Valves Closure 

TTNB Turbine Trip with Bypass Failure 

W Watt(s) 

US United States 

 
xvi 

 

 
1-1 

1. INTRODUCTION 

Kuosheng NPP is located on the northern coast of Taiwan. Its nuclear steam supply system is a 

type of BWR/6 designed and built by General Electric on a twin unit concept. Each unit includes 

two loops of recirculation piping and four main steam lines, with the thermal rated power of 

2894MWt. After the project of MUR for Kuosheng NPP, the operating power is 2943 MWt [1, 2]. 

Unit 1 started SPU from Cycle 24 and Unit 2 started SPU from Cycle 23. The operating power 

will be 3030 MWt. To uprate the power, the assessments of power plant transient should be 

analyzed. In the past, a TRACE model of Kuosheng NPP has been developed. The analysis and 

simulation of the TRACE model for the startup tests and hypothetical transients has been done. 

In this research, a RELAP5/MOD3.3 model of Kuosheng NPP is developed referred to FSAR [1], 

training documents [2], RELAP5 3D and TRACE model which had been developed before [3]. 

Further, the model in this research is built up with the SNAP interface rather than the text files. 

The visible thermal hydraulic components and control blocks in the SNAP interface help users 

build the model more efficiently and concretely. 

RELAP5/MOD3.3Patch04 code, which was developed for light water reactor (LWR) transient 

analysis at Idaho National Engineering Laboratory (INEL) for U.S. NRC, is applied in this 

research. This code is often performed to support rulemaking, licensing audit calculations, 

evaluation of accident, mitigation strategies, evaluation of operator guidelines, and experiment 

planning analysis [4]. Same as other thermal hydraulic analysis codes, RELAP5/MOD3.3 is 

based on nonhomogeneous and nonequilibrium model for the two-phase system. However, 

calculations in this code will be solved by a fast, partially implicit numerical scheme to permit 

economical calculation of system transients. It can produce accurate transient analysis results in 

relatively short time, which means large amounts of sensitivity or uncertainty analysis might be 

possible. 

Symbolic Nuclear Analysis Package (SNAP) is an interface of NPP analysis codes which 

developed by U.S. NRC and Applied Programming Technology, Inc. Different from the traditional 

input deck in ASCII files, the graphical control blocks and thermal hydraulic connections make 

researches comprehend the whole power plant and control system more easily. Due to these 

advantages, the RELAP5/MOD3.3 model of Kuosheng NPP was developed with SNAP 

interface.

To ensure the applicability of this model, three startup tests including feedwater pumps trip 

(FWPT), load rejection with bypass (LRWB) and main isolation valves closure (MSIVC) would be 

analyzed first. With the comparison of RELAP5 results and startup tests data, it shows that the 

RELAP5/MOD 3.3 model of Kuosheng NPP is consistent with the startup tests data. Moreover, 

three hypothetical accidents which referred to the TRACE model [5,6] would be performed. The 

comparison of RELAP data results and TRACE data results show great consistency. 
1-2 

 

 
2-1 

2. MODEL ESTABLISHMENT 

Overview of the RELAP5/MOD 3.3 Model 

Different from typical thermal hydraulic model establishment, the RELAP model in this research 

was developed in the SNAP interface. With SNAP interface, the components and control blocks 

are visible as shown in Figure 1. As a result, the users could set up component parameters and 

nodding diagram at same time. In this RELAP model, situation of NSSS in transient was mainly 

concerned. Hence, the turbines and feedwater pumps of Kuosheng NPP were assumed to be 

boundaries which were simulated by components Time dependent volume (TMDVOL). Four 

main steam pipe lines, which were consistent with the configuration of Kuosheng NPP, were 

developed. On these pipe lines, three important valves including main steam line isolation valves, 

turbine stop valves and turbine control valves are developed. Further, there are totally 16 

safety/relief valves connected on the main steam pipe lines. All the opening and closing setpoints 

are also developed according to the arrangement of Kuosheng NPP. 

The reactor vessel is developed by several kinds of components including Branch, Pipe, Single 

junction and single volume. Four pipes which are developed to simulate fuel assemblies are 

connected to heat structures inside the reactor vessel. Source data of these heat structures are 

referred to the total reactor power. To simplified the model and save the computational time, 

Point Kinetics and Separable feedback types are chosen for the reactor kinetics. Two 

recirculation loops and recirculation pumps are set up according to the configuration of the power 

plant. Further, there are two control valves developed on two recirculation loops respectively to 

adjust the recirculation flow rate. 20 jet pumps in the NPP are merged into two jet pump 

components to save the computational time. More details of the components properties and the 

setting of control systems are described below. 
2-2 

Figure 1 Overview of the Kuosheng NPP RELAP5/MOD3.3 Model 
2-3 

Reactor Kinetics 

The Point kinetic and SEPARABL feedback types are developed in this transient. Further, 

GAMMA-AC fission product decay type is chosen to calculate the decay heat. The total reactor 

power was set according to different transient scenario respectively. Initial reactivity and value of 

Beta over Lambda were 0 and 125 respectively, which were referred to the manual of RELAP5. 

In addition, with some unit converting, density and Doppler reactivity feedback table were 

referred to the TRACE and RETRAN model which had been verified before. Details of the 

reactor kinetics are shown in Figure 2 to Figure 4. 

To simplify the control model and reduce the computer time, the scram control systems were 

developed according to each case respectively. However, as shown in Figure 5, the scram 

reactivity feedback was written into Table 900, which was also referred to the TRACE model 

verified before, to control the power after the reactor scram. As the table trip set in Table 900 was 

initiated by the power control system in each case, the negative reactivity feedback, which was 

increasing with time, would be concerned in the power calculation. 
2-4 

Figure 2 Parameters of Reactor Kinetics in RELAP5 Model 

 
2-5 

Figure 3 Density Reactivity Feedback Table of Reactor Kinetics 

Figure 4 Doppler Reactivity Table of Reactor Kinetics 
2-6 

Figure 5 Scram Reactivity Feedback Table 

 
2-7 

Main Steam Pipe Lines 

According to the configuration of Kuosheng NPP, there are four main steam pipe lines developed 

in this RELAP5 model. On the main steam pipe lines, there are one main steam line isolation 

valve, one turbine stop valve and one turbine control valve in each pipe lines respectively. 

Further, 16 safety/relief valves are also set up on the main steam pipe lines. At the ends of the 

pipe lines, Time Dependent Volume components are set to simulate the turbines, which are 

assumed to be the boundaries of this model. In addition, the turbine bypass pipe line and valve 

are also developed in this model as shown in Figure 1. All the geometry and thermal hydraulic 

conditions of these components mentioned above are referred to the FSAR, training materials 

and configuration of Kuosheng NPP. Details of the properties are shown in Figure 6 to Figure 10. 

 
2-8 

Figure 6 Properties of Main Steam Isolation Valves

 
2-9 

Figure 7 Properties of Turbine Stop Valves 

 
2-10 

Figure 8 Properties of Turbine Control Valves 

 
2-11 

Figure 9 Time Dependent Volume Properties of the Main Steam Pipe Lines 

 
2-12 

Figure 10 Properties of Safety/Relief Valves 

 
2-13 

Pumps 

Typically, three types of pumps including feedwater pumps, recirculation pumps and jet pumps 

are built to supply core flow. Because the feedwater inlet assumed to be the boundary in this 

transient, three feedwater pumps in Kuosheng NPP were developed with the component Time 

Dependent Volume. After these components, there are three Time Dependent Junction 

components which are developed to control the feedwater flow rate. Properties of the Time 

Dependent Volume and Time Dependent Junction are shown in Figure 11 and Figure 12. There 

are two recirculation loops in Kuosheng NPP. As a result, two recirculation loops and flow control 

valves are developed in this RELAP5 model. Generally, rotation speed of the recirculation 

pumps would not be changed. The recirculation flow rates are varied by controlling the area of 

the recirculation control valves. Details of the recirculation loop setting are shown in Figure 13 

and Figure 14. There are 20 jet pumps in the reactor vessel of Kuosheng NPP. However, to save 

the computational time, these jet pumps are merged into two jet pumps. Details of the jet pump 

properties are shown in Figure 15. 
2-14 

Figure 11 Details of Feedwater Time Dependent Volume 

 
2-15 

Figure 12 Details of Time Dependent Junction of Feedwater Pumps 
2-16 

Figure 13 Recirculation Pumps Properties 

 
2-17 

Figure 14 Properties of Recirculation Controlling Valves 

 
2-18 

Figure 15 Jet Pumps Properties 

 
2-19 

Animation Model 

Because of developing analysis model with SNAP interface, analysis data results could be 

transferred into animations which could illustrate the situation of NPP during transients more 

easily and clearly. Further, based on the TRACE animation model which had been developed by 

our group before, the RELAP5/MOD3.3 animation model just needs a few modifications in this 

research. As shown in Figure 16, the reactor vessel was developed with several “Plenum” 

animation components. Plenum at different height level would be connected to correspondent 

“Branch”. Feedwater pumps were developed with “Fill” components because they were the 

boundary conditions of the RELAP5/MOD3.3 thermal hydraulic model. Two “Pipe” components 

which would change color depth with mass flow rate were connected to the feedwater pumps. 

Recirculation pumps were also developed with “Single Pump” components, which would be 

green if the pumps are working and would be red if the pumps are tripped. 4 main steam pipe 

lines were developed with “Pipe” animation components. These pipes would indicate the mass 

flow rate by changing color depth. Besides, all the valves including MSIVs, TSVs, TCVs and 

SRVs were also developed with animation components “Gate Valve”. These valves would be 

green if they are open and would be red if they are closed. On the right side of the animation 

model, there are two bars which could immediately indicate the reactor power and dome 

pressure respectively during transients. With this animation model, the interactions of different 

components could be observed more directly and easily. It would also be more attractive and 

comprehensive on the presentation. Moreover, this RELAP5/MOD3.3 animation model was 

modified from the existing TRACE animation model. It would be convenient to do cross validation 

of TRACE and RELAP5 analysis results. 

Figure 16 RELAP5/MOD3.3 Animation Model 
2-20 

 
3-1 

3. START-UP TEST RESULTS 

The startup tests data results collected in this research came from report prepared by INER in 

1996. However, this report only recorded the data results before Measurement Uncertainty 

Recapture Power Uprate. To confirm the applicability of the RELAP5 model, the results of 

TRACE which had been performed for the startup tests transients were referenced. Hence, there 

are three curves (startup test, TRACE and RELAP5) compared in each case illustrated be low. 

Moreover, the steady state parameters of startup test will be a little lower than that of TRACE 

and RELAP5 computational results. 

3.1 100% Load Rejection with Bypass Valve Open (LRWB) 

Event Description 

The LRWB test was held in November 11th, 1981. The operator triggered the Breaker of the main 

generator as the reactor was operated in 100% power. Then, the turbine control valves closed 

and turbine bypass valve opened which caused the transient start and reactor scram. The 

objective of this test is to ensure whether the turbine control valves could closed immediately or 

not. In addition, the responsibility of the turbine bypass valve, safety/relief valves and the reactor

protection system should also be assessed. The acceptance criteria are shown as follows: 

 Turbine bypass valve should open in 0.1 second after the turbine control valves closure. 

Further, the bypass valve should reach to 80% open of the design capacity in 0.3 second. 

 The feedwater setting should avoid the main steam pipe lines being flooded. 

 The feedwater control system should maintain the water level which would not trigger the 

MSIVs closure 

 The dome pressure increasing should not be higher than 25 psi in 30 seconds and the heat 

flux ratio should not be over than 2%. 

To verify the RELAP5 model, the turbine control valves and bypass valves were developed 

according to the FSAR of the power plant [1]. Table 1 lists the comparisons of the initial values 

from the RELAP5 analysis and startup test data. Due to the stability, a 210-second steady state 

was performed for ensuring that all the parameters matched the operating conditions as shown 

in Table 2. However, to compare the results with startup test and TRACE model data more easily, 

the steady state will be eliminated in the figures. 

At 210.2 second, the TCVs started to close. Control system of the TCVs is shown in Figure 17. 

The control system of TBV opening is shown in Figure 18. After closure of TCVs for 0.011 

second (variable trip 409), the TBV start to close. Further, when the dome pressure was lower 

than 940 psia (6.48 MPa, variable trip 410), the TBV would start to close as shown in Figure 18. 

The recirculation pumps were tripped as the dome pressure reached to 1134.7 psia (7.82 MPa). 

The control logic of recirculation pumps trip was shown in Figure 19. Three groups of the SRVs 

setpoints were developed as shown in Figure 20. 
3-2 

Table 1 Initial Conditions of RELAP5 and Startup Test Data in 100% LRWB Transient 

Parameter Test data RELAP5 

Power,MW 2894.0 2894.0 

Steam Dome Pressure (MPa) 6.96 6.99 

Feedwater Flow(kg/sec) 1551.66 1574.11 

Steam Flow(kg/sec) 1510.91 1574.10 

Narrow Range Water Level,cm 90 89.9 

Test Date ’81/11/11 -

Table 2 Event Sequence of 100% LRWB Transient 

Event(sec) Test data RELAP5 

TCV Start to Close 0.2 210.2 

Reactor Scram 0.236 210.29 

BPV Fully Open 0.329 210.32 

TCV Fully Close 0.394 210.35 

Peak Dome Pressure Time 

(value) 3.9 (7.43MPa) 212.84 (7.55MPa) 

End of Analysis - 230 

 
3-3 

Figure 17 Controlling System of Turbine Control Valves in 100% LRWB Transient 

Figure 18 Controlling System of Turbine Bypass Valve in 100% LRWB Transient 
3-4 

Figure 19 Controlling System of Recirculation Pumps Trip 

Figure 20 Controlling System of Safety Valves in 100% LRWB Transient 
3-5 

Analysis Data Results 

Once the generator tripped, the turbine control valves started to close and fully close in 0.1 

second. As the TCVs started to close, the steam flow decreased immediately as shown in Figure 

21. Figure 22 shows great consistency between RELAP5 model and startup test result. As the 

TCVs started to close, the reactor scram signal was initiated. Due to the large negative reactivity 

feedback, the power decreased rapidly at first second as shown in Figure 22. Nonetheless, the 

reactor core still generated some decay heat by the fission product. Hence, the steam was still 

produced inside the reactor vessel. Even though the turbine bypass valve started to open at 0.1 

second according to the acceptance criteria mentioned above, the dome pressure still increased 

as shown in Figure 23. 

 
3-6 

Figure 21 Steam Flow Variation during the 100% LRWB Transient 

Figure 22 Core Power Variation during the 100% LRWB Transient 

0

200

400

600

800

1000

1200

1400

1600

1800

0 5 10 15 20 25

Flow Rate (kg/sec)

Time (Sec)

Main Steam Flow

RELAP5 TRACE Satartup Test Data

0

500

1000

1500

2000

2500

3000

3500

0 5 10 15 20 25

Power (MW)

Time (Sec)

Core Power

RELAP5 TRACE Startup Test Data
3-7 

Figure 23 Dome Pressure Variation during the 100% LRWB Transient 

 

6.2

6.4

6.6

6.8

7

7.2

7.4

7.6

7.8

0 5 10 15 20 25

Pressure (MPa)

Time (Sec)

Dome Pressure

RELAP5 TRACE Startup Test Data
3-8 

3.2 96% Main Steam Line Isolation Valves Closure (MSIVC) 

Event Description 

The Main steam line Isolation Valves Closure test was held in November 6th in 1981. As the 

reactor operated at 96% power, the operator triggered the circuits F6A and F6B to close all the 

MSIVs. The objective of this test including: 

(1) Whether the MSIVs work normally in such power and operating condition. 

(2) Obtained the transient variation as the MSIVs were closing and fully closed. 

(3) Ensure the fully closure time of the MSIVs. 

The acceptance criteria are shown as follows: 

(1) The closure time of each MSIV should not shorter than 3 seconds and should not longer than 

5 seconds. Further, the electric signal delayed time should shorter than 0.5 second. 

(2) After MSIVs closure, the dome pressure increment should not larger than 160 psi in 30 

seconds. The heat flux increment should not exceed 2% of the initial value. 

(3) The feedwater control system should avoid flooding of the main steam pipe lines. 

(4) After the overpressurization transient, the safety/relief valves could be re-closed. 

From the startup test results, the reactor scram as the MSIVs reached to 94% open. The MSIVs 

closure time was about 4.08 to 4.53 seconds. The dome pressure increment was about 118 psi 

during the transient. Because the dome pressure reached to the setpoints of the safety/relief 

valves, there were two groups of valves opened during the transient. Based on these data results, 

the RELAP5 model control system in the following description could be developed. 

To ensure the analysis results were consistent to the operating conditions, a 210-second steady 

state was performed. Comparison of the startup test data and RELAP5 initial condition results 

was listed in Table 3 and the events sequence of this transient was shown in Table 4. At 210 

second, the MSIVs started to close and fully closed in 4.9 second according to the startup test 

report. The reactor scram signal might come from three sources including neutron flux, reactor 

vessel pressure and main steam line isolation valves area as shown in Figure 24. In this transient, 

the reactor scram signal was initiated due to the MSIVs reached to 94% open with 0.94 second 

delayed time. Different from the startup test report, there are three groups of SRVs setpoints in 

this transient as shown in Figure 25 and Table 4, which lists the event sequence of the MSIVC 

transient. 

 
3-9 

Table 3 Initial Conditions of RELAP5 and Startup Test Data in 96% MSIVC Transient 

Parameters Test data RELAP5 

Power (MW) 2836.97 2836.72 

Steam Dome Pressure (MPa) 6.92 6.96 

Feedwater Flow (kg/sec) 1498.41 1519.29 

Steam Flow (kg/sec) 1456.64 1519.29 

Narrow Range Water Level,cm 90.0 89.9 

Test Date ’81/11/06 -- 

Table 4 Event Sequence of 96% MSIVC Transient 

Event (sec.) Test data RELAP5 

MSIVFI Initiated 0.3 210.3 

Reactor Scram 1 210.92 

MSIV Full Isolation 4.83 214.825 

Peak Dome Pressure Time 

(value) 18.4 (7.7MPa) 218.1 (7.44MPa) 

End of Analysis - 240 
3-10 

Figure 24 Controlling System of Reactor Scram in 96% MSIVC Transient 

Figure 25 Controlling System of Safety/Relief Valves in 96% MSIVC Transient 

 
3-11 

Analysis Data Results 

Figure 26 shows the steam flow comparison of startup test data, TRACE and RELAP5 results. 

As the MSIVs started to close, the steam flow decreased gradually. 5 seconds later, the MSIVs 

fully closed and the steam flow rate reached to 0 kg/sec. From Figure 27, it is obviously that after 

the reactor scram, the power variations of startup test and computational results were a little 

different. This difference might come from the decay heat model choice. In the RELAP5 model, 

only GAMMA-AC fission product decay model was chosen and the fission product type was not 

chosen. This model choice might be different from the real reactor core. 

Because of the MSIV closure, the steam would be trapped inside the reactor vessel. Hence, the 

dome pressure might increase as shown in Figure 28. However, from this figure, it can be 

noticed that before the dome pressure increasing, it would decrease first because the reactor 

had scrammed at about 1 second, as shown in Figure 28. The steam produced inside the reactor 

core would be reduced. Hence, the dome pressure decreased until the MSIVs fully closed. In 

addition, from this figure, it can be found that the increasing of dome pressure in RELAP5 model 

was faster than that in startup test or in TRACE model. The SRVs would function earlier in 

RELAP5 model. Hence, the peak dome pressure of RELAP5 model would be lower than that of 

other two data results. 

Figure 29 shows the comparison of feedwater flow among these three data results. For the 

computational results, they are consistent to each other; however, the feedwater flow of startup 

test data was much different. This difference might come from the setting of recirculation pumps. 

The curves of recirculation pumps of Kuosheng NPP were hard to be collected completely. Some 

of the pump curves in RELAP5 model were referred to the EPRI data. Although the pumps 

developed EPRI data would not be so different from the pumps installed in the NPP, however, 

the feedwater feedback system might be very sensitive to the core flow. Hence, the feedwater 

flow rate of computational result was much different from the startup test data. 
3-12 

Figure 26 Steam Flow Variation during the 96% MSIVC Transient 

Figure 27 Core Power Variation during the 96% MSIVC Transient 

 

-500

0

500

1000

1500

2000

0 5 10 15 20 25 30 35

Flow Rate (kg/sec)

Time (Sec)

Main Steam Flow

RELAP5 TRACE Startup Test Data

0

500

1000

1500

2000

2500

3000

3500

0 5 10 15 20 25 30 35

Power (MW)

Time (Sec)

Power

RELAP5 TRACE Startup Test Data
3-13 

Figure 28 Dome Pressure Variation during the 96% MSIVC Transient 

Figure 29 Feedwater Flow Variation during the 96% MSIVC Transient 

6.4

6.6

6.8

7

7.2

7.4

7.6

7.8

0 5 10 15 20 25 30 35

Pressure (MPa)

Time (Sec)

Dome Pressure

RELAP5 TRACE Startup Test Data

0

500

1000

1500

2000

2500

0 5 10 15 20 25 30 35

Flow Rate (kg/sec)

Time (Sec)

Feedwater Flow

RELAP5 TRACE Startup Test Data
3-14 

3.3 94% Feedwater Pumps Trip (94% FWPT) 

Event Description 

The 94% Feedwater Pumps Trip was held in November 6th, 1981. The transient started as the 

operator tripped one of three operating feedwater pumps at 94% power. The objective of this test 

is to examine whether the flow control valves (FCV) run back in the recirculation loops could 

lower the core power or not when a feedwater pump failed. If the core power could be lowered by 

the reduction of the recirculation flow, the water level would not be below L3, which was the 

setpoint of the reactor scram. 

Table 5 shows the initial conditions of RELAP5 analysis results and startup test data and Table 6 

shows the event sequence of this transient. Same as previous two cases, a 210-second steady 

state analysis was performed to ensure all the parameters matched the operating conditions. At 

214.9 second, one of the three feedwater pumps was tripped. Generally speaking, three 

feedwater inlet junctions were controlled by control block 994 (FW-24) as shown in Figure 30. To 

perform the FWPT transient, one of the feedwater junctions was controlled by control block 1 

(FW-25), which was functioned by feedwater trip table 902. Further, the L4 water level signal 

would initiate FCV run back to lower the core power which would maintain the water level. 

According to the configuration of KS NPP, the FCV area of recirculation loop A would decline 

from 80% to 50% and the FCV area of recirculation loop B would decline from 69% to 46%. 

Control system of the FCV is shown in Figure 31. 
3-15 

Table 5 Initial Conditions of RELAP5 and Startup Test Data in 94% FWPT Transient 

Parameters Test data RELAP5 

Power (MW) 2790.40 2790.80 

Steam Dome Pressure (MPa) 7.00 7.01 

Feedwater Flow (kg/sec) 1470.35 1494.77 

Steam Flow (kg/sec) -- 1494.76 

Narrow Range Water Level,cm 90.0 89.9 

Test Date ’81/11/06 -- 

Table 6 Event Sequence of 94% FWPT Transient 

Event (sec.) Test data RELAP5 

Feedwater pump Trip 4.9 214.9 

L4 Singal 15.1 217.01 

Minimum Power Value 18.5 (57.1%) 218.90 (54.2%) 

Minimum Core Flow 19.4 (79.3%) 218.31 (78.76%) 

END of Analysis - 240 

 
3-16 

Figure 30 Controlling System of Feedwater Pump Trip in 96% FWPT Transient 

Figure 31 Controlling System of Flow Control Valves in 96% FWPT Transient 
3-17 

Analysis Data Results 

Figure 32 shows the comparisons of feedwater flow among startup test, TRACE and RELAP 

models. At 4.9 seconds, one of three feedwater pumps was tripped. The feedwater flow 

decreased for about 5 seconds. About 17 seconds later, as shown in Figure 33, the NRWL 

reached to 0.848 m, which is L4 for Kuosheng NPP. The recirculation flow control valves ran 

back. Hence, core inlet flow rate decreased to 80 %, as shown in Figure 34. Decreasing of the 

core flow would increase the void fraction, which means that the density reactivity feedback 

would also decrease. The core power dropped down at about 17 seconds as shown in Figure 35. 

In addition, because the water level declination in TRACE model was a little slower than the 

other two cases, it can be noticed that the core power decreasing in TRACE model was later 

than that in the other two data results. 

Due to the declination of feedwater flow rate and water level, the feedwater control system 

adjusted the other two available feedwater junctions and increased the flow rate as shown in 

Figure 32. However, at the end of the transient, core power of computational data results was 

different from that of startup test since the water level might be controlled in startup test as 

shown in Figure 33. The density reactivity feedback would be different from the computational 

models. Nonetheless, due to lack of some other transient scenario or information, the water level 

followed the computational results rather than matched the startup test data in both TRACE and 

RELAP5 models. 

 
3-18 

Figure 32 Feedwater Flow Variation during the 94% FWPT Transient 

Figure 33 Narrow Range Water Level Variation during the 94% FWPT Transient 

1000

1100

1200

1300

1400

1500

1600

1700

0 5 10 15 20 25 30 35

Flow rate (kg/sec)

Time (Sec)

Feedwater Flow

RELAP5 TRACE Startup Test Data

0.7

0.75

0.8

0.85

0.9

0.95

1

0 5 10 15 20 25 30 35

Water Level (m)

Time (Sec)

Narrow range water level

RELAP5 TRACE Startup test
3-19 

Figure 34 Core Flow Variation during the 94% FWPT Transient 

Figure 35 Core Power Variation during the 94% FWPT Transient 

60

70

80

90

100

110

Flow rate (normalized %)

0 5 10 15 20 25 30 35

Time (Sec)

Core Flow

RELAP5 TRACE Startup Test Data

1450

1950

2450

2950

0 5 10 15 20 25 30 35

Power (MW)

Time (Sec)

Core Power

RELAP5 TRACE Startup Test Data
3-20 

 

 
4-1 

4. HYPOTHETICAL TRANSIENT RESULTS 

To ensure the safety of SPU, a series of overpressurization transient test of TRACE code had 

been performed before by our group [5, 6]. In this research, same hypothetical accidents have 

been performed again by RELAP5 code to verify the accuracy of TRACE model. As a result, the 

initial conditions such as power, feedwater flow, steam flow and dome pressure would be same 

as the prediction of TRACE code as shown in Table 7. 

Table 7 Initial Conditions of TRACE and RELAP5 Models 

Parameters Unit FSAR TRACE RELAP5 

Power (MWt) 3030 3030 3030 

Dome Pressure (MPa) 7.17 7.17 7.17 

Feedwater Flow (kg/sec) 1645.3 1641 1640.6 

Steam Flow (kg/sec) 1645.3 1641 1640.6 

Core inlet flow (kg/sec) 10645 10704 10674 

Recirculation flow (Single loop) (kg/sec) 1549.5 1538.6 1548.4 

Core exit pressure (MPa) 7.23 7.3 7.23 

NWRL (m) 0.934 0.934 0.933 

4.1 Main Steam Line Isolation Valves Closure with Bypass Failure 

Hypothetical Event Sequence 

In this hypothetical accident, a 210-second steady state was performed for ensuring that all the 

parameters matched the operating conditions. At time point 210 second, the main steam line 

isolation valves (MSIVs) closed in three seconds. In order to promote the difficulties of this 

transient for the plant, the closure time was assumed to be 3 seconds in this case, which was 

shorter than 5 seconds that described in FSAR. According to the FSAR description, reactor 

scram signal in this case might come from two ways including neutron flux exceeding 122% or 

dome pressure exceeding 7.66 MPa. The reactor scram control system is shown in Figure 36. 

Further, as the dome pressure reached to 7.82 MPa, recirculation pump trip signal would be sent 

out with delayed time 0.14 second. For the conservation reason, only 11 safety valves works 

even though that there are 16 safety/relief valves on the main steam pipelines. All these 11 

safety valves were still divided into three groups, as shown in Figure 37, with different set-points 

including 8.38 MPa for two valves, 8.48 MPa for five valves and 8.55 MPa for four valves. 

Besides, after the closure signal sent out, the safety valves would fully close in 0.15 second with 

an electronic delayed time 0.4 second. Table 8 lists the MSIVC events sequence and important 

set-points. From the data results, the reactor scrammed due to neutron flux exceeding 122% in 

this transient. Further, the safety valves will fully open in 0.15 second with delayed time 0.4 

second. Once the safety valves open, the dome pressure was under controlled and the NPP was 

back to the safe situation. 
4-2 

Figure 36 Controlling System of Reactor Scram in MSIVC Hypothetical Accident 

Figure 37 Controlling System of Safety Valves in MSIVC Hypothetical Accident 
4-3 

Table 8 Events Sequence and Comparison of MSIVC Hypothetical Accident

Events RELAP5 

(sec) 

TRACE 

(sec) Notes 

Steady state 0~210 0~210 

Power 3030 MW 

Feedwater flow rate 1641 kg/sec

Feedwater temperature 488.7K 

Core inlet flow rate 11177.3 

kg/sec 

Dome pressure 7.17MPa 

MSIVs closure 210 210 Fully closure time 3 seconds 

MSIVs fully closed 213.210 213.06 

Reactor scram signal 

initiated 213.211 213.14 Power reached to 122% 

Reactor scram 213.310 213.23 Delayed time 0.09 second 

Recirculation pumps 

tripped 213.315 213.46 Dome pressure 7.82MPa 

Safety valves group 1 

opened (8.38MPa,Group 

1) 

213.915 214.53 Delayed 0.4 second 

Safety valves group 2 

opened (8.48MPa,Group 

2) 

214.015 214.65 Delayed 0.4 second 

Safety valves group 3 

opened (8.55MPa,Group 

3) 

214.115 214.73 Delayed 0.4 second 

End of the analysis 220 220 

 
4-4 

Analysis Data Results 

Figure 38 shows the comparison of steam flow rate between TRACE and RELAP5 model. From 

this figure, it is obviously that the steam flow rate dropped rapidly once the MSIVs closed at 210 

second. At 213 second, the MSIVs fully closed and the dome pressure increased greatly as 

shown in Figure 39. Hence, the void fraction inside the reactor core decreased which increased 

the power greatly as shown in Figure 40. In this figure, the core power of RELAP5 model was 

much lower than that of TRACE model. This difference might came from the neutron kinetic 

feedback table setting. In the RELAP5 model, the neutron slowing effect would reference the 

table which includes density and reactivity. However, in the TRACE model, the reactor power 

iteration would reference the table containing void fraction and reactivity. As a result, the power 

calculation might have some differences during high-power transient. 

Despite the fact that the reactor had scrammed at 213.23 second, the decay heat still heated the 

reactor vessel and generated steam in both the RELAP5 and TRACE model. As a result, the 

dome pressure kept increasing. At 214.53 second, the dome pressure was higher enough to 

meet the setpoint of group 1 safety valves. However, 3 opened safety valves discharged the 

steam insufficiently. The dome pressure kept increasing until the group 2 and group 3 safety 

valves opened at 214.65 and 214.73 second respectively in TRACE model. Further, because the 

dome pressure varied near the setpoints, the safety valves opened and closed for three times. 

On the other hand, in the RELAP5 model, safety valves of group 1 opened at 214.31 second, 

which was not far from that in TRACE model. Safety valves of group 2 and group 3 also opened 

at 214.40 second and 214.51 second. However, the dome pressure prediction of RELAP5 model 

was higher than that of TRACE model. As a result, the dome pressure variation did not triggered 

the setpoinhts of safety valves closure in the RELAP5 model. Hence, as shown in Figure 39, the 

reactor vessel in the RELAP5 model maintain a higher pressure than that in the TRACE model. 

From both the RELAP5 TRACE data results, it is known that the core power could be controlled 

by the scram system. In addition, as long as the safety valves opened, the dome pressure would 

not exceed the ASME criteria 9.58 MPa. Kuosheng NPP was safe in the MSIVC hypothetical 

transient. 
4-5 

Figure 38 Steam Flow Variation during the MSIVC Hypothetical Accident 

Figure 39 Dome Pressure Variation during the MSIVC Hypothetical Accident 

-500

0

500

1000

1500

2000

210 212 214 216 218 220 222

Flow rate (kg/sec)

Time (Sec)

Steam Flow

Main steam flow rate_RELAP5

Main steam flow rate_TRACE

7

7.5

8

8.5

9

9.5

210 212 214 216 218 220 222

Pressure (MPa)

Time (Sec)

Dome Pressure (MPa)

Dome Pressure_RELAP5 Dome Pressure_TRACE
4-6 

Figure 40 Core Power Variation during the MSIVC Hypothetical Accident 

 

0

1000

2000

3000

4000

5000

6000

7000

8000

210 212 214 216 218 220 222

Power (MW)

Time (Sec)

Core Power (MW)

Power_RELAP5 Power_TRACE
4-7 

4.2 Turbine Trip with Bypass Failure 

Hypothetical Event Sequence 

In the hypothetical accident of TSVC, a 500 second steady state was performed to ensure all the 

parameters matched the operation conditions. Same as the MSIVC hypothetical accident, the 

reactor scram signal might come from neutron flux exceeding 122% or dome pressure exceeding 

7.66 MPa. However, in the TSVC case, the scram signal might also come from the TSVs 

reaching 90% open. In the RELAP5 model, all the possibilities were concerned as shown in 

Figure 41. 

At 500 second, the turbine tripped and the turbine stop valves started to close with closure time 

0.1 second. As the TSVs reached to 90% open, the reactor scram signal was initiated. According 

to the FSAR, there would be 0.08 second delayed time. After the TSVs closure, the dome 

pressure increased and the void fraction inside the reactor core decreased. Hence, a positive 

reactivity feedback functioned and the power increased. 

For conservative reason, on 6 safety relief valves opened with delayed time 0.4 second at the 

dome pressure 7.94 MPa and closed at dome pressure 7.62 MPa. The control system of SRVs 

was shown in Figure 42. Further, the safety relief valves would fully opened in 0.15 second. 

Once the safety relief valves opened, the dome pressure and the power plants would be under 

controlled. Table 9 listed the event sequences and important setpoints of the TRACE model 

during TSVC transient. 
4-8 

Figure 41 Controlling System of Reactor Scram in TTBF Hypothetical Accident 

Figure 42 Controlling System of Relief Valves in TTBF Hypothetical Accident 
4-9 

Table 9 Events Sequence and Comparison of TTNB Hypothetical Accident 

Events RELAP5 (sec) TRACE (sec) Notes 

Steady state 0~499 0~499 

Power 3030 MW 

Feedwater flow rate 1640 kg/sec 

Feedwater temperature 488.7K 

Core inlet flow rate 11177.3 

kg/sec 

Dome pressure 7.17MPa 

TSVs start to close 500 500 Fully closure time 0.1 second 

Reactor scram signal 

initiated 500.02 500.01 Signal initiated at TSVs 90% open 

Reactor scram 500.11 500.09 Delayed time 0.08 second 

TSVs fully closed 500.1 500.1 

Relief valves opened 501.34 501.77 

Open at dome pressure 7.94MPa 

and closed at dome pressure 

7.63MPa 

Peak dome pressure 

(MPa) 

503.81 

(8.53MPa) 

502.27 

(8.2MPa) 

End of the analysis 505 505 
4-10 

Analysis Data Results 

The TSVs started to close at 500 second and fully closed at 500.1 second. Figure 43 shows the 

comparison of steam flow rate in TRACE and RELAP5 model. At the beginning of the transient, 

the steam flow dropped rapidly in both models. As mentioned in previous case, because the 

friction data is hard to be determined in the RELAP5 model, the steam flow rate might decreased 

(or increased) faster than that in TRACE model. The steam inside the reactor vessel was trapped; 

hence, the dome pressure increased as shown in Figure 44. Due to the increasing dome 

pressure, the void fraction of the reactor core would decrease, which caused a positive reactivity 

feedback. As a result, the power increased as shown in Figure 45. Once the TSVs reached to 

90% open, reactor scram signal was sent out with a delayed time 0.08 second. That is, the 

negative scram feedback would function at 500.09 second. However, the scram reactivity 

feedback needed some time to dominate the core power. Hence, as shown in Figure 45, the core 

power kept increasing until the scram feedback dominated the power at 501.4 second. Despite 

that the reactor scram control system in RELAP5 model was same as that in TRACE model, the 

negative reactivity feedback dominated the core power with a timing difference. Although the 

reactor scrammed at 501.4 second, the dome pressure still increased as shown in Figure 44 

because the decay heat still produced steam inside the reactor core both in TRACE and 

RELAP5 model. 

As the dome pressure reached to 7.94 MPa, the relief valves open signal was sent out with 

delayed time 0.4 second. Hence, as shown in Figure 44, the relief valves really functioned until 

the dome pressure reached to 8.2 MPa. Once the relief valves opened, the dome pressure 

decreased in TRACE model. However, in the RELAP5 model, the dome pressure would maintain 

at 8.5 MPa, which was still under the ASME regulation 9.58MPa. 

Due to the opening of the relief valves, the steam flow rate increased again as shown in Figure 

43. However, only six relief valves functioned in this hypothetical accident, the steam flow was 

only 0.3 times of the normal operating flow rates. From the analysis results of RELAP5 and 

TRACE models, it is noticed that the reactor scram system can successfully inhibit positive 

reactivity which comes from the decreasing void fraction. In addition, even though the bypass 

valve is failed and only six relief valves are functioned, the dome pressure can still be controlled 

sufficiently. 
4-11 

Figure 43 Steam Flow Variation during the TTBF Hypothetical Accident 

Figure 44 Dome Pressure Variation during the TTBF Hypothetical Accident 

-1000

-500

0

500

1000

1500

2000

500 501 502 503 504 505 506

Flow Rate (kg/sec)

Time (Sec)

Steam flow

Main steam flow_RELAP5 Main steam flow_TRACE

7

7.2

7.4

7.6

7.8

8

8.2

8.4

8.6

500 501 502 503 504 505 506

Pressure (MPa)

Time (Sec)

Dome pressure

Dome pressure_RELAP5 Dome P_TRACE
4-12 

Figure 45 Core Power Variation during the TTBF Hypothetical Accident 

0

500

1000

1500

2000

2500

3000

3500

4000

4500

500 501 502 503 504 505 506

Power (MW)

Time (Sec)

Core Power (MW)

Power_RELAP5 Power_TRACE
4-13 

4.3 Load Rejection with Bypass Failure 

Hypothetical Event Sequence 

In the TCVC hypothetical accident, a 210-second steady state simulation was performed to 

ensure all the parameters reached to the operating conditions. At 210 second, the load rejection 

happened and the TCVs closed immediately. Once the turbine control valves closed, the main 

steam line flow decreased immediately which increased the dome pressure. The positive void 

fraction reactivity feedback functioned and the power increased. Different from the previous two 

cases, the recirculation pumps trip signal and the reactor scram signal were sent out immediately 

as the TCVs closed. However, due to the signal delayed time, the reactor would scram 0.07 

second later as shown in Figure 46. Further, the recirculation pumps would be out of service 0.14 

second later. For conservative reason, there were only 7 safety valves and 6 relief valves 

functioned in this case with signal delayed time 0.4 second. The control system of the SRVs is 

shown in Figure 47. Same as previous two cases, once the safety/relief valves open, the dome 

pressure decreased and the NPP was under controlled. Table 10 lists the important setpoints 

and transient events of the TCVC hypothetical accident. 
4-14 

Figure 46 Controlling System of Reactor Scram in LRNB Hypothetical Accident 

Figure 47 Controlling System of Safety/Relief Valves in LRNB Hypothetical Accident 

 
4-15 

Table 10 Events Sequence and Comparison of LRNB Hypothetical Accident 

Events RELAP5 (sec) TRACE (sec) Notes 

Steady state 0~209 0~209 Power 3030 MW 

Feedwater flow rate 1641 kg/sec 

Feedwater temperature 488.7K 

Core inlet flow rate 11177.3 kg/sec 

Dome pressure 7.17MPa 

Turbine control 

valves closed 

210 210 Start of the transient; valve closed in 0.15 

second 

Reactor scram 210.07 210.07 Signal initiated at TCVs close with delayed 

0.07 second 

Recirculation 

pumps trip 

211.4 211.4 Signal initiated at TCVs close with delayed 

for 0.14 second 

Safety/relief 

valves open 

211.36 211.76 7 safety valves and 6 relief valves 

functioned in this model with delayed time 

0.4 second 

Peak dome 

pressure (MPa) 

213.75 

(8.5 MPa) 

211.83 

(8.2 MPa) 

End of analysis 215 215 

 
4-16 

Analysis Data Results 

Figure 48 shows the comparison of steam flow rate in TRACE and RELAP5 models. Once the 

turbine control valves closed, the steam flow rate dropped in both models. However, the steam 

flow in RELAP5 model dropped more rapidly than that in TRACE model even with same TCVs 

closure time. Due to the declination of the steam flow, the dome pressure increased immediately 

as shown in Figure 49. As a result, the void fraction inside the reactor core decreased which 

caused a positive reactivity feedback. The power increased as shown in Figure 50. Different from 

previous two cases, the reactor scram signal was generated at once as the turbine control valves 

closed. The power would decrease at the beginning of the transient both in the TRACE and 

RELAP5 models. After 210.5 second, the power increased because the positive void fraction 

reactivity feedback dominated the power. However, at 211.1 second for RELAP5 model and 

211.2 second for TRACE model, the negative scram reactivity feedback dominated the power 

again. The power dropped rapidly. After 211.76 second, the safety/relief valves opened and 

released the steam inside the reactor vessel. Same as previous two cases, the dome pressure of 

RELAP5 model would maintain at a higher pressure and the dome pressure of TRACE model 

would decrease a little after the SRVs opening. Nonetheless, both the dome pressure of TRACE 

and RELAP5 model were under controlled and the steam flow rate was back to a stable value. 

On the other hand, the recirculation pump and feedwater pump trip signal would be sent out to 

avoid the void fraction increased after reactor scram. As a result, Figure 51 shows that the core 

flow rate decreased. 

From those data results mentioned above, the dome pressure did not exceed the criteria 

9.58Mpa which was regulated by ASME. Further, all the inside and outside flow rates became 

stable. Kuosheng NPP was under control in the TCVC transient. 

 
4-17 

Figure 48 Steam Flow Variation during the LRNB Hypothetical Accident 

Figure 49 Dome Pressure Variation during the LRNB Hypothetical Accident 

-1000

-500

0

500

1000

1500

2000

2500

Flow Rate (kg/sec)

210 211 212 213 214 215 216

Time (Sec)

Steam flow

Main steam flow_RELAP5 Main steam flow_TRACE

7

7.2

7.4

7.6

7.8

8

8.2

8.4

8.6

210 211 212 213 214 215 216

Pressure (MPa)

Time (Sec)

Dome Pressure

Dome pressure_RELAP5 Dome pressure_TRACE
4-18 

Figure 50 Core Power Variation during the LRNB Hypothetical Accident 

Figure 51 Core Flow Variation during the LRNB Hypothetical Accident 

 

0

1000

2000

3000

4000

5000

210 211 212 213 214 215 216

Power (MW)

Time (Sec)

Core Power

Power_RELAP5 Power_TRACE

6000

7000

8000

9000

10000

11000

12000

13000

210 211 212 213 214 215 216

Flow rate (kg/sec)

Time (Sec)

Core Flow

Core flow_RELAP5 Core flow_TRACE
5-1 

5. CONCLUSIONS 

In this research, the RELAP5/MOD 3.3 model of Kuosheng NPP was successfully developed. 

The startup test data and TRACE model analysis results were used to compare with the results 

of RELAP5/MOD3.3 model for FWPT, LRWB and MSIVC transient. The predictions of 

RELAP5/MOD3.3 were consistent with the data of startup test and TRACE roughly. It indicates 

that there is a respectable accuracy in Kuosheng NPP RELAP5/MOD3.3 model. Additionally, the 

analysis of the above RELAP5/MOD3.3 model for the overpressurization hypothetical accidents 

was performed. The predictions of RELAP5/MOD3.3 were also consistent with the results of 

TRACE roughly. However, there is a difference between the prediction values of 

RELAP5/MOD3.3 and TRACE for the above transients. Therefore, we will study this difference 

and modify the models in the future. 
5-2 

 

 
6-1 

6. REFERENCES 

1. Taiwan Power Company, Final Safety Analysis Report for Kuosheng Nuclear Power Station 

Units 1&2 (FSAR), Taiwan Power Company, Republic of China (Taiwan), 2001. 

2. Taiwan Power Company, BWR-6 MARK-III Containment of Kuosheng NPS, Taiwan Power 

Company, Republic of China (Taiwan), 1995. 

3. Lin, K.Y., et al., “Verification of the Kuosheng BWR/6 TRACE Model with Load Rejection 

Startup Test”, ASME verification and validation, Las Vegas, Neveda, 2012. 

4. Information Systems Laboratories, Inc., RELAP5/MOD3.3 Code Manual Volume I: Code 

Structure, System Models, and Solution Methods, Information Systems Laboratories, Inc., 

Rockville Maryland, 2010 

5. Chang, H.C., et al., “The Turbine Trip without Bypass Analysis of Kuosheng BWR/6 Using 

TRACE/FRAPCON/FRPATRAN”, American Nuclear Society Annual Meeting, Reno Nevada, 

2014. 

6. Chen, S.W., et al., “Fuel Rod Uncertainty Analysis of Kuosheng NPP with 

TRACE/FRAPTRAN/DAKOTA codes in SNAP Interface”, ICAPP 2015, Nice France. 







NUREG/IA-0464 RELAP5/MOD3.3 Model Assessment and Hypothetical Accident Analysis of 

Kuosheng Nuclear Power Plant with SNAP Interface