NUREG/IA-0445 

International 

 Agreement Report

The Establishment and Assessment of 

Chinshan (BWR/4) Nuclear Power Plant 

TRACE/SNAP Model

 

Prepared by:

Jong-Rong Wang, Hao-Tzu Lin, Hsiung-Chih Chen*, Chunkuan Shih* 

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 

101 Section 2, Kuang Fu Rd., HsinChu, 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: April 2014

Date Published: July 2014

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0445 

International 

 Agreement Report

The Establishment and Assessment of 

Chinshan (BWR/4) Nuclear Power Plant 

TRACE/SNAP Model

 

Prepared by:

Jong-Rong Wang, Hao-Tzu Lin, Hsiung-Chih Chen*, Chunkuan Shih* 

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 

101 Section 2, Kuang Fu Rd., HsinChu, 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: April 2014

Date Published: July 2014

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

iii

ABSTRACT

Chinshan Nuclear Power Plant (NPP) is the first NPP in Taiwan which is of the BWR/4 type plant. 

This research focuses on the development of the Chinshan NPP TRACE/SNAP model. In order 

to check the system response of the Chinshan NPP TRACE/SNAP model, this study uses the 

analysis results of Final Safety Analysis Report (FSAR) to assess the Chinshan NPP 

TRACE/SNAP model. The increase in reactor pressure transients including turbine trip and main 

steam isolation valves closure were selected to validate the Chinshan NPP TRACE/SNAP

model. The trends of TRACE analysis results were consistent with the FSAR data. It indicates 

that there is a credible fidelity in the Chinshan NPP TRACE/SNAP model. In addition, this 

research also investigates the application of the Chinshan NPP TRACE/SNAP model for the 

core shroud leakage. The core shroud leakage is one of current issues of concern of the U.S. 

NRC and other BWR/4 NPP owners. This research utilizes the Chinshan NPP TRACE/SNAP

model to perform the core shroud leakage transients for Chinshan NPP safety analysis. The 

TRACE analysis results show that the simple core shroud leakage transient does not influence 

the Chinshan NPP fuel temperature to cause any core damage. However, the core shroud 

leakage combined with station blackout (SBO) or steamline break (loss of coolant accident, 

LOCA) transient would cause the cladding temperature to reach higher than 1088 K and affected

the Chinshan NPP safety.
iv

 
v

FOREWORD

The US NRC (United States Nuclear Regulatory Commission) is developing an advanced 

thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development 

of TRACE is based on TRAC, integrating RELAP5 and other programs. NRC has determined 

that in the future, TRACE will be the main code used in thermal hydraulic safety analysis, and no 

further development of other thermal hydraulic codes such as RELAP5 and TRAC will be 

continued. A graphic user interface program, SNAP (Symbolic Nuclear Analysis Program) which 

processes inputs and outputs for TRACE is also under development. One of the features of 

TRACE is its capacity to model the reactor vessel with 3-D geometry. It can support a more 

accurate and detailed safety analysis of nuclear power plants. TRACE has a greater simulation 

capability than the other legacy codes, especially for events like LOCA. 

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

Maintenance Program) which includes the development and maintenance of TRACE. INER 

(Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) is the organization in 

Taiwan responsible for the application of TRACE in thermal hydraulic safety analysis, for 

recording user’s experiences of it, and providing suggestions for its development. To meet this 

responsibility, the TRACE/SNAP model of Chinshan NPP has been built. In this report, this study 

uses the analysis results of FSAR to assess the Chinshan NPP TRACE/SNAP 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 OF CHINSHAN (BWR/4) NPP .................................................................................... 2-1

3. RESULTS................................................................................................................................ 3-1

3.1 Turbine Trip without Bypass Valve Analysis ..................................................................3-2

3.2 Main Steam Isolation Valve Closure Analysis ................................................................3-8

3.3 Core Shroud Leakage Analysis .................................................................................... 3-14

4. CONCLUSIONS ...................................................................................................................... 4-1

5. REFERENCES ........................................................................................................................ 5-1
viii

 
ix

FIGURES

Page

Figure 1: The flow chart of establishing and verifying the TRACE/SNAP model of 

Chinshan NPP................................................................................................................2-2

Figure 2: The TRACE/SNAP model of Chinshan NPP...................................................................2-3

Figure 3: The animation model of Chinshan NPP...................................................................2-4

Figure 4: The 3D vessel condition of Chinshan NPP..............................................................2-5

Figure 5: The channel condition of Chinshan NPP.................................................................2-6

Figure 6: The separator condition of Chinshan NPP..............................................................2-7

Figure 7: The recirculation pump condition of Chinshan NPP..............................................2-8

Figure 8: The jetpump condition of Chinshan NPP ................................................................2-9

Figure 9: The SRV condition of Chinshan NPP..................................................................... 2-10

Figure 10: The power condition of Chinshan NPP.................................................................2-11

Figure 11: The comparison of power between FSAR and TRACE for turbine trip 

without bypass valve...............................................................................................3-3

Figure 12: The comparison of steam dome pressure between FSAR and TRACE 

for turbine trip without bypass valve ....................................................................3-4

Figure 13: The comparison of core inlet flow between FSAR and TRACE for turbine 

trip without bypass valve........................................................................................3-5

Figure 14: The 3-D picture of steam dome pressure and core inlet pressure for 

turbine trip without bypass valve ..........................................................................3-6

Figure 15: The animation model of Chinshan NPP for turbine trip without 

bypass valve.............................................................................................................3-7

Figure 16: The comparison of power between FSAR and TRACE for MSIV closure ..........3-9

Figure 17: The comparison of steam dome pressure between FSAR and TRACE for 

MSIV closure ..........................................................................................................3-10

Figure 18: The comparison of core inlet flow between FSAR and TRACE for MSIV 

closure .....................................................................................................................3-11

Figure 19: The 3-D picture of steam dome pressure and core inlet pressure for MSIV

closure .................................................................................................................... 3-12

Figure 20: The animation model of Chinshan NPP for MSIV closure ................................. 3-13

Figure 21: The TRACE results: power for core shroud leakage transient ......................... 3-18
x

Figure 22: The TRACE results: Doppler reactivity for core shroud leakage transient.....3-19

Figure 23: The TRACE results: NRWL for core shroud leakage transient)........................3-20

Figure 24: The TRACE results: vessel water level for core shroud leakage transient..... 3-21

Figure 25: The TRACE results: cladding temperature for core shroud leakage 

transient..................................................................................................................3-22

Figure 26: The TRACE results: cladding oxidation for core shroud leakage transient....3-23

Figure 27: The animation model of Chinshan NPP for core shroud leakage transient....3-24
xi

TABLES

Page

Table 1: The comparison of initial conditions between FSAR and TRACE data for 

turbine trip without bypass valve ..................................................................................... 3-1

Table 2: The comparison of initial conditions between FSAR and TRACE for MSIV 

closure............................................................................................................................. 3-1

Table 3: The comparison of sequences between FSAR and TRACE data for turbine 

trip without bypass valve .................................................................................................3-2

Table 4: The comparison of sequences between FSAR and TRACE for MSIV closure .................3-8

Table 5: The sequences of case 1 for core shroud leakage transient ............................... 3-15

Table 6: The sequences of case 2 for core shroud leakage transient ............................... 3-15

Table 7: The sequences of case 3 for core shroud leakage transient ...............................3-16

Table 8: The sequences of case 4 for core shroud leakage transient ............................... 3-17

 
xii

 

 
xiii

EXECUTIVE SUMMARY

An agreement in 2004 which includes the development and maintenance of TRACE has been signed 

between Taiwan and USA on CAMP. INER is the organization in Taiwan responsible for applying TRACE 

to thermal hydraulic safety analysis in order to provide users’ experiences and development suggestions. 

To fulfill this responsibility, the TRACE/SNAP model of Chinshan NPP is developed by INER. 

According to the user manual, TRACE is the product of a long term effort to combine the capabilities of the 

NRC’s four main systems codes (TRAC-P, TRAC-B, RELAP5 and RAMONA) into one modernized 

computational tool. NRC has ensured that TRACE will be the main code used in thermal hydraulic safety 

analysis in the future without further development of other thermal hydraulic codes, such as RELAP5 and 

TRAC. Besides, the 3-D geometry model of reactor vessel, which is one of the representative features of 

TRACE, can support a more accurate and detailed safety analysis of NPPs. On the whole TRACE 

provides greater simulation capability than the previous codes, especially for events like LOCA.

Chinshan NPP was building in 1970. It is the first NPP in Taiwan which is the BWR/4 plant and the

original rated power for each unit is 1775 MWt. After the project of MUR (Measurement Uncertainty 

Recovery) for Chinshan NPP, Unit 2 started MURPU (Measurement Uncertainty Recovery Power Uprate) 

from April 6, 2008 for Cycle 23 and Unit 1 started MURPU from November 8, 2008 for Cycle 24. The 

operating power is 101.7% of the original designed rated power, which is 1805 MWt now. This research 

focuses on the development of the Chinshan NPP TRACE/SNAP model. In order to check the system 

response of the Chinshan NPP TRACE/SNAP model, this study uses the analysis results of FSAR to 

assess the Chinshan NPP TRACE/SNAP model. The increase in reactor pressure transients including 

turbine trip and main steam isolation valves closure were selected to validate the Chinshan NPP 

TRACE/SNAP model. The trends of TRACE analysis results were consistent with the FSAR data. It 

indicates that there is a respectable accuracy in the Chinshan NPP TRACE/SNAP model. Besides, this 

research also focuses on the application of the Chinshan NPP TRACE/SNAP model in the core shroud 

leakage. The core shroud leakage is one of issues of concern by the U.S. NRC and NPPs. This research 

is using the Chinshan NPP TRACE/SNAP model to perform the core shroud leakage transients for 

Chinshan NPP safety analysis. The TRACE analysis results show that the pure core shroud leakage

transient wasn’t influence the Chinshan NPP safety. However, the core shroud leakage + SBO or

steamline LOCA transient caused the cladding temperature larger than 1088 K and affected the 

Chinshan NPP safety.

xv

ABBREVIATIONS

BPV Bypass Valve

BWR Boiling Water Reactor

CAMP Code Applications and Maintenance Program

FSAR Final Safety Analysis Report

IGSCC Intergranular Stress Corrosion Cracking

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

LOCA Loss of Coolant Accident

MSIV Main Steamline Isolation Valve

MUR Measurement Uncertainty Recovery

MURPU Measurement Uncertainty Recovery Power Uprate

NPP Nuclear Power Plant

NRC Nuclear Regulatory Commission

RPV Reactor Pressure Vessel

SBO Station Blackout

SNAP Symbolic Nuclear Analysis Package

SRV Safety Relief Valve

TCV Turbine Control Valve

TPC Taiwan Power Company

TRACE TRAC/RELAP Advanced Computational Engine

TSV Turbine Stop Valve
xvi

 
1-1

1. INTRODUCTION

Chinshan NPP was building in 1970. It is the first NPP in Taiwan which is the BWR/4 plant and the

original rated power for each unit is 1775 MWt. After the project of MUR for Chinshan NPP, Unit 2 started 

MURPU from April 6, 2008 for Cycle 23 and Unit 1 started MURPU from November 8, 2008 for Cycle 

24. The operating power is 101.7% of the original designed rated power, which is 1805 MWt now. The 

safety analysis of the NPP is very important work in the NPP safety. The importance of NPP safety 

analysis has been raised and there is more concern for the safety of the NPPs in the world after the 

Fukushima NPP event occurred. The advanced thermal hydraulic code named TRACE has been

developed by U.S. NRC for NPP safety analysis. According to the user manual [1], TRACE is the product 

of a long term effort to combine the capabilities of the NRC’s four main systems codes (TRAC-P, 

TRAC-B, RELAP5 and RAMONA) into one modernized computational tool. The development of TRACE 

is based on TRAC, combining with the capabilities of RELAP5 and other programs. SNAP, a graphic user 

interface program that processes the inputs and outputs of TRACE, has also been developing. One of the 

features of TRACE is its capacity to model the reactor vessel with 3-D geometry. It could support a more 

accurate and detailed safety analysis for nuclear power plants. TRACE offers the greater simulation 

capability than other old codes, especially for events such as LOCA.

This research focuses on the development of the Chinshan NPP TRACE/SNAP model. In our previous 

study [2], the preliminary TRACE/SNAP model of Chinshan NPP was established and assessed by the 

startup tests data. The Chinshan NPP TRACE/SNAP model includes one 3-D vessel, six channels which 

are used to simulate 408 fuel bundles, four steamlines, and 10 SRVs components, etc.. In order to check 

the system response of the Chinshan NPP TRACE/SNAP model, this study uses the analysis results of 

FSAR to assess the Chinshan NPP TRACE/SNAP model. The increase in reactor pressure transients 

including turbine trip and main steam isolation valves closure were selected to validate the Chinshan NPP 

TRACE/SNAP model. 

Besides, this research also focuses on the application of the Chinshan NPP TRACE/SNAP model in the 

core shroud leakage. Intergranular stress corrosion cracking (IGSCC) of BWR internal components has 

been identified as a technical issue of concern by the U.S. NRC and NPPs. The core shroud is among the 

list of internals susceptible to IGSCC. So IGSCC may cause core shroud leakage generated. Therefore, 

the core shroud leakage safety analysis of Chinshan NPP was performed by the Chinshan NPP 

TRACE/SNAP model in this study. Besides, refer to the conditions of Fukushima NPP event, the core 

shroud leakage + SBO or LOCA transients were also analyzed by TRACE for Chinshan NPP.
 
2-1

2. MODEL OF CHINSHAN (BWR/4) NPP

The code versions adopted in this research are SNAP v 2.2.0 and TRACE v 5.0p3. The process

of Chinshan NPP TRACE/SNAP model development is as follows (shown in Fig. 1): First, the 

system and operating data for the cases of FSAR of Chinshan NPP are collected [1]-[7]. Second, 

several important control systems such as recirculation flow control system, pressure control 

system and feedwater flow control system etc. are established by SNAP and TRACE. Next, other 

necessary components (e.g., RPV (Reactor pressure vessel) and main steam piping) are added 

into the TRACE/SNAP model to complete the TRACE/SNAP model for Chinshan NPP. Finally, 

the Chinshan TRACE/SNAP model is verified with the cases of FSAR. The TRACE/SNAP model 

of Chinshan NPP is presented in Fig. 2. SNAP also can use the TRACE results data to make an

animation for transient, such as Fig. 3.

The reactor vessel is divided into 9 axial levels, 4 radial rings, and 2 azimuthal sectors (shown in 

Fig. 4). Six channels are used to modeling 408 fuel rods and connect to the vessel’s level 3~4. 

There were 27 axial nodes and 6 radial nodes in one channel. The water rods and partial length 

rods were also simulated in the channels (shown in Fig. 5). Besides, six separators connect to 

the level 6~8 of the vessel (see Fig. 6). Two recirculation loops are set outside the vessel, with a 

recirculation pump in each loop (see Fig. 7). In the TRACE model, 10 groups of injection pumps

are merged into an equal injection pump (see Fig. 8). We use valve components to simulate the 

SRV (safety relief valve), MSIV (main steamline isolation valve), TSV (turbine stop valve), TCV 

(turbine control valve) and BPV (bypass valve). The critical flow models for the MSIV, SRV, TCV, 

TSV, and BPV have been considered in our analysis (see Fig. 9). The reactor vessel connects 

with four steamlines. Besides, every steamline has one MSIV, one TCV/TSV, and 1~4 SRVs. 

The steam goes through the top of the reactor and into the main steamlines. Finally, the steam 

passes through the TCVs and enters the turbines (boundary conditions). We also build bypass 

pipelines and the turbine bypass valve. Moreover, the break component at the end of bypass 

valve is used to simulate the condenser.

In Chinshan NPP TRACE/SNAP model, “point kinetic” parameters such as delay neutron 

fraction, Doppler reactivity coefficient, and void reactivity coefficient are provided as TRACE 

input for power calculations. These data are set in the power component (shown in Fig. 10).

In the Chinshan NPP TRACE/SNAP analysis model, there are three simulation control systems

included (1) feedwater flow control system, (2) pressure control system and (3)recirculation flow 

control system. Currently, these three control systems have been built by the signal variables, 

control blocks, trips and components of TRACE/SNAP.
2-2

Figure 1 The flow chart of establishing and verifying the TRACE/SNAP model of 

Chinshan NPP

Gathering the design data, control system and FSAR data of

Chinshan NPP

Building the TRACE/SNAP model with control 

systems by TRACE/SNAP.

The TRACE/SNAP model for Chinshan NPP

Verifying the TRACE/SNAP model for Chinshan NPP

Recirculation flow control system Pressure control system Feedwater flow control system

Related data 

of FSAR 

chapter 15

Startup tests 

data

Testing the convergence of steady state and comparing 

the TRACE results with FSAR and startup tests data

Analyzing the transients and comparing the TRACE 

results with FSAR and startup tests data

Yes

Deficiency
2-3

Figure 2 The TRACE/SNAP model of Chinshan NPP
2-4

Figure 3 The animation model of Chinshan NPP
2-5

Figure 4 The 3D vessel condition of Chinshan NPP
2-6

 

 

Figure 5 The channel condition of Chinshan NPP

1. Full length fuel rod

2. Partial length fuel rod

3. Water rod
2-7

Figure 6 The separator condition of Chinshan NPP
2-8

Figure 7 The recirculation pump condition of Chinshan NPP
2-9

Figure 8 The jetpump condition of Chinshan NPP
2-10

Figure 9 The SRV condition of Chinshan NPP
2-11

Figure 10 The power condition of Chinshan NPP
2-12

 
3-1

3. RESULTS

Before the transient calculation of Chinshan TRACE/SNAP model begins, it is necessary to carry 

out the steady state calculation and make sure that the system parameters (such as the 

feedwater flow, steam flow, dome pressure, and core flow, etc.) are in agreement with FSAR 

data under the steady state condition. The results of analysis of TRACE are clearly consistent 

with FSAR data under the steady state condition (See Table 1 and 2).

Table 1 The comparison of initial conditions between FSAR and TRACE data for turbine 

trip without bypass valve

Parameters FSAR TRACE Difference (%)

Power (MWt) 1864 1864 0

Dome pressure (MPa) 7.13 7.07 -0.84

Core Inlet flow (kg/sec) 7240 7217 -0.32

Feedwater flow (kg/sec) 1009 1006 -0.3

Steam flow (kg/sec) 1009 1006 -0.3

NRWL (m) 1.10 1.11 0.91

Table 2 The comparison of initial conditions between FSAR and TRACE for MSIV 

closure

Parameters FSAR TRACE Difference (%)

Power (MWt) 1864 1864 0

Dome pressure (MPa) 7.27 7.17 -1.38

Core Inlet flow (kg/sec) 7240 7217 -0.32

Feedwater flow (kg/sec) 1009 1006 -0.3

Steam flow (kg/sec) 1009 1006 -0.3

NRWL (m) 1.10 1.11 0.91
3-2

3.1 Turbine Trip without Bypass Valve Analysis

The reactor is initially operating at 105% of NBR power with vessel dome pressure of 1034.7 psi 

(7.13 MPa). Table 3 compares the turbine trip transient’s sequences of FSAR data with TRACE. 

Their sequences are very similar. In turbine trip transient, the reactor scram and recirculation 

pump trip is initiated by position switches on the TSV when the valves are less than 90% open. 

The turbine bypass valve system is failed in this transient.

Fig.11~14 shows the results of TRACE. Fig. 11 depicts the power curves of FSAR data and

TRACE. The trends of their curves are similar. The increase of the power was caused by the 

TSVs closing. The TSVs closing decreased the reactor’s void fraction which generated the 

positive reactivity. Then, the scram initiated and the power dropped. Fig. 12 compares the steam 

dome pressure of FSAR and TRACE. The trends of the curves are approximately in agreement. 

The TSV closing caused the dome pressure to rise. Then, SRVs opened and led to the decline of

dome pressure. Fig. 13 shows the comparison of core inlet flow between FSAR and TRACE. The 

curve of TRACE is consistent with the FSAR data. Due to the dome pressure increase, it resulted 

in the core inlet flow rising during 0.5~1.5 sec. Then, recirculation pumps trip caused the 

decrease of core inlet flow. In summary, the trends of TRACE prediction are consistent with 

FSAR data but there are a few differences in the values of the prediction. Because we cannot 

find the detailed FSAR analysis data, we don’t know what the reasons cause the differences of 

TRACE results and FSAR data. Finally, by using TRACE results and SNAP’s animation function, 

Fig. 14 shows the dome pressure and core inlet pressure of 3-D animation pictures. The higher 

dome pressure and core inlet pressure show in 2 sec and it can be observed from Fig. 14.

Finally, Fig. 15 depicts the animation model of Chinshan NPP for this transient.

Table 3 The comparison of sequences between FSAR and TRACE data for turbine trip 

without bypass valve

Sequences

Time (s)

FSAR TRACE

Transient started 0 0

TSV closure less than 90 percent open 0.015 0.015

TSV fully closure 0.1 0.1

Recirculation pump tripped 0.6 0.6

SRVs opened 0.95 1.04

Reactor scrammed 1 1

Reactor pressure rose to peak 2.4 2.45

End of the time 6 6
3-3

Figure 11 The comparison of power between FSAR and TRACE for turbine trip without 

bypass valve
3-4

Figure 12 The comparison of steam dome pressure between FSAR and TRACE for 

turbine trip without bypass valve
3-5

Figure 13 The comparison of core inlet flow between FSAR and TRACE for turbine trip 

without bypass valve
3-6

Figure 14 The 3-D picture of steam dome pressure and core inlet pressure for turbine

trip without bypass valve
3-7

 

 

Figure 15 The animation model of Chinshan NPP for turbine trip without bypass valve

3.2 Main Steam Isolation Valve Closure Analysis

As main steam isolation valves close, position switches on the valves initiate a reactor scram 

when the valves in three or more main steamlines are less than 90 percent open. However, in

the FSAR case, it assumes multiple failures of the position switch on each MSIV to initiate a 

reactor scram and reliance upon a neutron flux initiated scram. The reactor is initially operating at 

105 percent of NB rated power with a vessel dome pressure of 1054.7 psi (7.27 MPa). Table 4 

compares the MSIV closure transient’s sequences of FSAR data with TRACE results. Their 

sequences are very similar. In MSIV closure transient, the reactor scram is initiated when the 

neutron flux rises to 122%. Then, the recirculation pumps trip and SRVs open.
3-8

Fig. 16~19 shows the comparisons of FSAR and TRACE. Fig. 16 depicts the power curves of

FSAR and TRACE. The result of TRACE is consistent with FSAR data. The MSIVs closing 

resulted in the increase of the power. The MSIVs closing decreased the reactor’s void fraction 

which generated the positive reactivity. Then, the scram was initiated by the neutron flux signal 

(122%) and the power started to decrease after 2 sec. However, at this time, the dome pressure 

still rose and the core inlet flow increased, which caused the void fraction still to decrease. It 

offered the positive reactivity which led to the power increase during 2.5~3 sec. Finally, the 

control rods fully inserted which resulted in the power dropped after 3 sec. Fig. 17 shows the 

steam dome pressure comparison of FSAR and TRACE. The trends of the curves are 

approximately in agreement. The steam dome pressure went up after the MSIV closure. When 

the dome pressure rose to the critical value, SRVs opened and led to the decline of dome 

pressure. Fig. 18 shows the core inlet flow results of FSAR and TRACE, and their curves are 

similar. The closure of MSIV caused the increase of the pressure. The increase of the pressure 

caused the core inlet flows rising during 1.5~3 sec. Then, the recirculation pump tripped which 

led to the decrease of core inlet flow. In summary, the TRACE results are consistent with FSAR 

data but there are a few differences in the values of the prediction. Due to the detailed FSAR 

analysis data cannot be found, we don’t know what the reasons cause the differences of TRACE 

results and FSAR data. Finally, Fig. 19 shows the dome pressure and core inlet pressure of 3-D 

animation pictures by using TRACE results and SNAP’s animation function. The higher 

pressures shows in 3.5 sec and it can be observed from Fig. 19. Finally, Fig. 20 depicts the 

animation model of Chinshan NPP for this transient.

Table 4 The comparison of sequences between FSAR and TRACE for MSIV closure

Sequences

Time (s)

FSAR TRACE

MSIV initiated closure 0.0 0

Control rod ready to insertion

(When Neutron flux rise to 122%)

1.5 1.37

SRVs opened 2.9 2.99

MSIV closure 3 3

Reactor pressure rose to peak 3.2 3.43

End of the time 7.5 7.5
3-9

Figure 16 The comparison of power between FSAR and TRACE for MSIV closure
3-10

Figure 17 The comparison of steam dome pressure between FSAR and TRACE for MSIV

closure
3-11

Figure 18 The comparison of core inlet flow between FSAR and TRACE for MSIV closure
3-12

Figure 19 The 3-D picture of steam dome pressure and core inlet pressure for MSIV

closure
3-13

 

 

Figure 20 The animation model of Chinshan NPP for MSIV closure
3-14

3.3 Core Shroud Leakage Analysis

There are four assumed cases in the Chinshan NPP TRACE analysis, as follows:

Case 1: the core shroud leakage occurred (only in 1 azimuthal sector of the vessel, a small hole), 

the length of break is 0.0254 m, the area of break is 0.07 m2

.

Case 2: the core shroud leakage occurred (in 2 azimuthal sectors of the vessel, 360 degree 

larger break), the length of break is 0.0254 m, the area of break is 11.1327 m2

.

Case 3: the case 2 condition + SBO + scram failed + RCIC failed.

Case 4: the case 2 condition + LOCA (one steamline break) + scram failed + RCIC failed.

The initial conditions of cases are 100% rated power/100% rated core flow and the sequences of 

case 1~4 are shown in Table 5~8. First, the simulation of steady state is performed during 0~200 

sec for all cases. Second, the core shroud leakage started to occur at 200 sec for the cases. 

Besides, the SBO transient (ex: recirculation pump, feedwater, turbine tripped) was performed 

for case 3 and the LOCA transient (ex: one steamline break, recirculation pump, feedwater 

tripped) was performed for case 4. Fig. 21~26 show the results of TRACE for all cases. 

Fig. 21 and 22 show the power and Doppler reactivity results of TRACE. In all cases, when core 

shroud leakage occurred, the power increased rapidly (shown in Fig. 21). Then the power 

dropped sharply due to the void fraction increase. After core shroud leakage generated fully, the 

powers of case 1 and 2 reached the new steady state. However, case 3 and case 4 had more 

negative Doppler reactivity (shown in Fig. 22) due to the fuel temperature raise (shown in Fig. 25). 

So the power of case 3 and case 4 was lower than the power case 1 and case 2 after core 

shroud leakage generated fully. 

Fig. 23 shows the water level results of Chinshan NPP core shroud leakage cases. In all cases, 

when core shroud leakage occurred, the water level rose rapidly and was larger than level 8. It 

also indicated that the level 8 signal was tripped. Therefore, if the Chinshan NPP finds the level 8 

signal tripped, it may be the core shroud leakage happened. The oscillation happened for the 

case 1 and case 2 after the core shroud leakage occurred. Besides, the feedwater tripped and 

RCIC failed resulted in no water injection to the vessel. Due to no water supply, the water level of 

the case 3 and case 4 decreased continuously and were lower than the TAF (see Fig. 24). Due 

to one steamline break, the larger steam rushed out for case 4 that caused the water level lower 

than the BAF. The above results caused the fuel temperature increase for the case 3 and case 4 

(shown in Fig. 25). The fuel temperature of the case 3 did not reach 1088K during 0~700 sec 

(when the temperature is larger than 1088 K, the zirconium-water reaction may generate). 

However, if the time of the transient is enough long, we think that the fuel temperature of the 

case 3 may be larger than 1088 K. Besides, because the fuel temperature of the case 4 was 

larger than 1088 K, the zirconium-water reaction happened (shown in Fig. 26). The above results

indicated that the fuels might be damaged after the zirconium-water reaction happened. Finally, 

Fig. 27 depicts the animation model of Chinshan NPP for this transient.
3-15

Table 5 The sequences of case 1 for core shroud leakage transient

Time (sec)

Transient started 0

The core shroud leakage started to occur 200

NRWL reached L8 209

The core shroud leakage generated fully 210

Transient ended 700

Table 6 The sequences of case 2 for core shroud leakage transient

Time (sec)

Transient started 0

The core shroud leakage started to occur 200

The core shroud leakage generated fully 201

NRWL reached L8 202

Transient ended 700
3-16

Table 7 The sequences of case 3 for core shroud leakage transient

Time (sec)

Transient started 0

The core shroud leakage started to occur 200

SBO happened 200.1

Recirculation pump tripped 200.1

Turbine tripped 200.1

Feedwater tripped 200.1

The core shroud leakage generated fully 201

SRVs opened (first time) 202.02

NRWL reached L8 203.02

Core water level reached TAF 250.1

Transient ended 700
3-17

Table 8 The sequences of case 4 for core shroud leakage transient

Time (sec)

Transient started 0

The core shroud leakage started to occur 200

One steamline broke 200.1

Feedwater tripped 200.1

The core shroud leakage generated fully 201

NRWL reached L8 201

Core water level reached TAF 204.02

Recirculation pump tripped 212.053

MSIV started to close 213.5

Core water level reached BAF 367.04

The temperature of cladding reached 1088K 377.04

Transient ended 700
3-18

Figure 21 The TRACE results: power for core shroud leakage transient
3-19

Figure 22 The TRACE results: Doppler reactivity for core shroud leakage transient
3-20

 

Figure 23 The TRACE results: NRWL for core shroud leakage transient
3-21

Figure 24 The TRACE results: vessel water level for core shroud leakage transient
3-22

Figure 25 The TRACE results: cladding temperature for core shroud leakage transient
3-23

Figure 26 The TRACE results: cladding oxidation for core shroud leakage transient
3-24

 

 

Figure 27 The animation model of Chinshan NPP for core shroud leakage transient
4-1

4. CONCLUSIONS

This research focuses on the development of the Chinshan NPP TRACE/SNAP model. The 

increase in reactor pressure transients including turbine trip and main steamline isolation valves 

closure were selected to validate the Chinshan NPP TRACE model. The results and sequences 

of TRACE are similar to the FSAR data for the turbine trip and main steamline isolation valves 

closure transients. Especially in the main steamline isolation valves closure transient, it shows 

the highest steam dome pressure happened (about rise 8.8 MPa) and the systems responses of 

TRACE are nearly the same with the FSAR data. By the above compared results, it indicates that 

there is a respectable accuracy in the Chinshan NPP TRACE/SNAP model and it also shows 

that the Chinshan NPP TRACE/SNAP model is satisfying for the purpose of Chinshan NPP 

safety analyses with confidence. 

This study also developed the TRACE/SNAP core shroud leakage models of Chinshan NPP. 

There are four assumed cases in this research and the results of TRACE show the accurate 

response of system for these cases. Besides, the results of TRACE indicated that the Chinshan 

NPP may find the core shroud leakage happened by the level 8 signal tripped. The TRACE 

analysis results also show that the pure core shroud leakage transient (case 1 and 2) wasn’t 

influence the Chinshan NPP safety. However, the core shroud leakage + station blackout (SBO) 

or one steamline break (LOCA) transient (case 3 and 4) may cause the cladding temperature 

larger than 1088 K and influence the Chinshan NPP safety.
4-2

 
5-1

5. REFERENCES

1. U.S. NRC, 2010. TRACE V5.0 User’s Manual.

2. Lee, J.H., Wang, J.R., Lin, H.T., Yang, S.M., Shih, C., 2009. TRACE modeling of Chinshan 

load rejection startup test, American Nuclear Society: 2009 Annual Meeting.

3. Taiwan Power Company, 2003. Chinshan Nuclear Power Station training manual, Revision 

6.

4. Taiwan Power Company, 1999. Final Safety Analysis Report for Chinshan Nuclear Power 

Station Units 1&2.

5. Hsu, W.S., Tang, J.R., 1997. The calculation notes of the startup test benchmark analysis for 

the Chinshan RETRAN-02 model, INER report.

6. Chen, W.S., Shih, C., Wang, J.R., Lin, H.T., 2011. TRACE Analysis of Inadvertent Startup of 

HPCI Transient in Chinshan BWR/4, ICAPP 2011.

7. Chang, W.J., Shih, C., Wang, J.R., Lin, H.T., 2012. TRACE Analysis of Station Blackout 

Transient in Chinshan BWR/4, NUTHOS-9.




UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA

-0445 The Establishment and Assessment of Chinshan (BWR/4) 

Nuclear Power Plant TRACE/SNAP Model