TRACE/SNAP Model Establishment of 

Chinshan Nuclear Power Plant for 

Ultimate Response Guideline

Prepared by:

Jong-Rong Wang*, Chunkuan Shih*, Jung-Hua Yang*, Hsiung-Chih Chen*, Shao-Wen Chen*, ShowChyuan Chiang**, Tzu-Yao Yu**

*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: February 2016

Date Published: May 2017

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

NUREG/IA-0478

TRACE/SNAP Model Establishment of 

Chinshan Nuclear Power Plant for 

Ultimate Response Guideline

Prepared by:

Jong-Rong Wang*, Chunkuan Shih*, Jung-Hua Yang*, Hsiung-Chih Chen*, Shao-Wen Chen*, ShowChyuan Chiang**, Tzu-Yao Yu**

*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: February 2016

Date Published: May 2017

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

NUREG/IA-0478

iii

ABSTRACT

In this research, the TRACE/SNAP model of Chinshan BWR/4 nuclear power plant (NPP) was 

established for the simulation and analysis of ultimate response guideline (URG). The main actions 

of URG are the depressurization and low pressure water injection of the reactor and containment 

venting. This research focuses to assess the URG utility of Chinshan NPP under Fukushima-like 

conditions. This study consists of three steps. The first step is the establishment of Chinshan NPP 

TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, startup 

tests and FSAR data were used to compare with the results of TRACE. The second step is the URG 

simulation and analysis under Fukushima-like conditions by using Chinshan NPP TRACE/SNAP

model. In this step, the no URG case was also performed in order to evaluate the URG effectiveness

of Chinshan NPP. In addition, the sensitivity study of URG and the required raw water injection were 

also performed. According to TRACE analysis results, the URG can keep the peak cladding 

temperature (PCT) below the criteria 1088.7 K under Fukushima-like conditions. It indicates that 

Chinshan NPP can be controlled in a safe situation. If Chinshan NPP does not perform the URG 

under Fukushima-like conditions, the water level may drop lower than TAF (top of active fuel) which 

means a safety issue about the fuel rods may be generated. In order to confirm the mechanical 

property and integrity of fuel rods, the final step is the analysis of FRAPTRAN. 

v 

FOREWORD

The U.S. 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 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 usually used in the nuclear power plants analysis. 

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

Maintenance Program) which includes the development and maintenance of TRACE. To meet this 

responsibility, the TRACE/SNAP model of Chinshan NPP was established. In this report, the 

TRACE/SNAP model of Chinshan NPP was used to evaluate the utility of URG. 

vii

TABLE OF CONTENTS

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

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

LIST OF FIGURES...............................................................................................................................ix

LIST OF TABLES................................................................................................................................. xi

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

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

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

2 ULTIMATE RESPONSE GUIDELINE (URG) ................................................................................. 3

3 CHINSHAN NPP TRACE/SNAP MODEL....................................................................................... 5

4 RESULTS......................................................................................................................................... 7

4.1 Load Rejection Transient........................................................................................................ 7

4.2 Turbine Trip without Bypass Valve Transient.......................................................................11

4.3 URG Simulation and Analysis ..............................................................................................14

5 CONCLUSIONS.............................................................................................................................47

6 REFERENCES...............................................................................................................................49

ix

LIST OF FIGURES

Figure 1 Correspondent Relation between NPP Operating States and Operating Procedures ..2

Figure 2 URG Flowchart..............................................................................................................4

Figure 3 Chinshan NPP TRACE Model.......................................................................................6

Figure 4 The Power of Startup Test and TRACE for Load Rejection Transient..........................8

Figure 5 The Steam Flow of Startup Test and TRACE for Load Rejection Transient .................9

Figure 6 The Dome Pressure of Startup Test and TRACE for Load Rejection Transient .........10

Figure 7 The NRWL of Startup Test and TRACE for Load Rejection Transient........................11

Figure 8 The Power of FSAR and TRACE for Turbine Trip without Bypass Valve Transient ...12

Figure 9 The Dome Pressure of FSAR and TRACE for Turbine Trip without Bypass 

Valve Transient ...........................................................................................................13

Figure 10 The Core Flow of FSAR and TRACE for Turbine Trip without Bypass

Valve Transient ...........................................................................................................14

Figure 11 The Dome Pressure Comparison of Case 1 and 2 .....................................................18

Figure 12 The Water Level Comparison of Case 1 and 2 ...........................................................19

Figure 13 The PCT Comparison of Case 1 and 2 .......................................................................20

Figure 14 The DRYWELL PRESSURE Comparison of Case 1 and 2........................................21

Figure 15 The Wetwell Pressure Comparison of Case 1 and 2 ..................................................22

Figure 16 The Drywell Temperature Comparison of Case 1 and 2.............................................23

Figure 17 The Suppression Pool Temperature Comparison of Case 1 and 2.............................24

Figure 18 The Dome Pressure Comparison of Case 1, 3 and 4 .................................................25

Figure 19 The Water Level Comparison of Case 1, 3 and 4 .......................................................26

Figure 20 The PCT Comparison of Case 1, 3 and 4 ...................................................................27

Figure 21 The Drywell Pressure Comparison of Case 1, 3 and 4 ...............................................28

Figure 22 The Wetwell Pressure Comparison of Case 1, 3 and 4 ..............................................29

Figure 23 The Drywell Temperature Comparison of Case 1, 3 and 4.........................................30

Figure 24 The Suppression Pool Temperature Comparison of Case 1, 3 and 4.........................31

Figure 25 The Dome Pressure Comparison of Case 1, 5 and 6 .................................................32

Figure 26 The Water Level Comparison of Case 1, 5 and 6 .......................................................33

Figure 27 The PCT Comparison of Case 1, 5 and 6 ...................................................................34

Figure 28 The Drywell Pressure Comparison of Case 1, 5 and 6 ...............................................35

Figure 29 The Drywell Temperature Comparison of Case 1, 5 and 6.........................................36

Figure 30 The Wetwell Pressure Comparison of Case 1, 5 and 6 ..............................................37

Figure 31 The Suppression Pool Temperature Comparison of Case 1, 5 and 6.........................38

Figure 32 The Water Level Results for The Sensitivity Study .....................................................39

Figure 33 The Water Metal Reaction Energy Result of FRAPTRAN for Case 2.........................40

Figure 34 The Oxide Thickness Result of FRAPTRAN for Case 2 .............................................41

Figure 35 The STRUCTURAL RADIAL GAP RESult of FRAPTRAN for Case 2........................42

Figure 36 The Gap Gas Pressure Result of FRAPTRAN for Case 2 ..........................................43

Figure 37 The Cladding Hoop Strain Result of FRAPTRAN for Case 2......................................44

Figure 38 The Cladding Hoop Stress Result of FRAPTRAN for Case 2.....................................45

Figure 39 The Fuel Enthalpy Result of FRAPTRAN for Case 2..................................................46

xi

LIST OF TABLES

Table 1 The Comparison of Initial Conditions of FSAR and TRACE ............................................5

Table 2 The Sequences of Startup Test and TRACE Data for Load Rejection Transient ............7

Table 3 The Sequences of FSAR and TRACE Data for Turbine Trip without

Bypass Valve Transient .................................................................................................12

Table 4 The Sequences of URG Case (Case 1).........................................................................17

Table 5 The Sensitivity Study of URG ........................................................................................17

xiii

EXECUTIVE SUMMARY

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. 

FRAPTRAN is a Fortran language computer code that calculates the transient performance of lightwater reactor fuel rods during reactor transients and hypothetical accidents such as loss-of-coolant 

accidents, anticipated transients without scram, and reactivity-initiated accidents. 

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 TRACE/SNAP and 

FRAPTRAN/SNAP models of Chinshan NPP was developed in this research.

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. In addition, Chinshan NPP finished the project of 

SPU (stretch power uprated) and the operating power is 1840 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 Taiwan after the Fukushima NPP 

disaster occurred. Therefore, with regard to this fact, Taiwan Power Company developed an 

additional URG to prevent BWR, PWR and ABWR from encountering core damage for events 

beyond design basis. 

The aim of this study is to use multiple computer codes to evaluate the URG effectiveness for 

Chinshan NPP. This study consists of three steps. The first step is the establishment of Chinshan

NPP TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, 

startup tests and FSAR data were used to compare with the results of TRACE. The second step is 

the URG simulation and analysis under Fukushima-like conditions by using Chinshan NPP 

TRACE/SNAP model. In this step, the no URG case was also performed in order to evaluate the 

URG effectiveness of Chinshan NPP. In addition, the sensitivity study of URG and the required raw 

water injection were also performed. According to TRACE analysis results, the URG can keep the 

PCT below the criteria 1088.7 K under Fukushima-like conditions. It indicates that Chinshan NPP 

can be controlled in a safe situation. If Chinshan NPP does not perform the URG under Fukushimalike conditions, the water level may drop lower than TAF which means a safety issue about the fuel 

rods may be generated. In order to confirm the mechanical property and integrity of fuel rods, the 

final step is the analysis of FRAPTRAN. 

xv

ABBREVIATIONS

AOPs Abnormal Operating Procedures

BPV Bypass valve

BWR Boiling-Water Reactor

CAMP Code Applications and Maintenance Program

EOPs Emergency Operating Procedures

MSIV Main Steamline Isolation Valve

NPP Nuclear Power Plant

OPs Operating Procedures

PCT Peak Cladding Temperature 

RPV Reactor pressure vessel

SAMPs Severe Accident Management Procedures

SNAP Symbolic Nuclear Analysis Program

SPU Stretch Power Uprated

SRV Safety Relief Valves

TAF Top of Active Fuel

TBV Turbine Bypass Valve

TCV Turbine Control Valve

TRACE TRAC/RELAP Advanced Computational Engine

TSV Turbine Stop Valve 

URG Ultimate Response Guideline

U.S. NRC United States Nuclear Regulatory Commission

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. In addition, Chinshan NPP finished the project of 

SPU and the operating power is 1840 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 Taiwan after the Fukushima NPP disaster occurred. 

In general, there are four categories for the NPP operating state, which involve normal operation, 

abnormal events/transients, accidents and severe accidents. For each operating state, there are

corresponding procedures to follow to secure NPPs safety and integrity. Figure 1 shows the 

correspondent relationship between NPP operating states and procedures. The first level is

operating procedures (OPs) which focus on the NPP operation within an acceptable range. The 

second level is abnormal operating procedures (AOPs) which aim at restoring the function of NPP 

systems that could impact the NPP operating margins. The third level is emergency operating 

procedures (EOPs) which focus on bringing the NPP to a safe and stable state by following a reactor 

trip or safety injection signal. The forth level is severe accident management procedures (SAMPs).

Uncertainties may exist in both NPP status and in the outcome of actions for severe accidents.

Therefore, SAMPS propose a range of possible actions and should allow for additional evaluation 

and alternative actions. However, EOP or SAMP is generally the symptom-based procedures to 

mitigate transients/accidents consequence and restore the NPP, depending on the real-time 

operational parameters of the NPP. For the compound severe accidents, such as Fukushima NPP 

disaster, its impact to NPP is relatively broad, rather than focus on one system or one area influence. 

Therefore, with regard to this fact, Taiwan Power Company developed an additional URG to prevent 

BWR, PWR and ABWR from encountering core damage for events beyond design basis [1]-[3]. 

The aim of this study is to use multiple computer codes to evaluate the URG effectiveness for 

Chinshan NPP. The advanced thermal hydraulic code named TRACE has been developed by U.S. 

NRC for NPP safety analysis. According to the user manual [4], 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. According to the 

user and assessment manual [4]-[5], TRACE also provides greater simulation capability than the 

previous codes (TRAC-P, TRAC-B, RELAP5 and RAMONA), especially for events like LOCA. 

Accordingly, an increasing number of researchers are using TRACE code to analyze test facilities 

and nuclear power plants. F. Mascari et al. [6] established a TRACE model for OSU-MASLWR test 

facility and performed a preliminary analysis, aiming at the evaluation of the code capability in 

predicting natural circulation phenomena and heat exchange from primary to secondary side by 

helical SG in superheated condition. J. Freixa and A. Manera [7] used TRACE to develop a model

of Olkiluoto-3 NPP. They performed the SBLOCA simulation and analysis by this model. In addition, 

they also used the ROSA large scale test facility experimental data to test the nodalization for 

transient analyses. Konstantin Nikitin and Annalisa Manera [8] developed a TRACE model of 

BWR/6 NPP and performed the analysis of an ADS spurious opening event. Their paper depicted 

that TRACE results were in good agreement with the plant data. Anthony Pollman et al. [9] used 

TRACE to establish a model of MANOTEA facility. The analysis of rapid-condensation transient

was performed by this TRACE model. They used the MANOTEA facility experimental data to 

compare the results of TRACE. Ovidiu-Adrian Berar et al. [10] performed the RELAP5 and TRACE 
2 

assessment of Achilles natural reflood experiment. Their paper indicated that both TRACE and 

RELAP5 are capable of reproducing the reflood phenomena at a satisfactory level. But some 

discrepancies between the predicted variables and the experimental data suggested further 

investigation of the TRACE reflood model. Davide Papini et al. [11] used the tests data carried out 

on a passive containment condenser unit in the PANDA large-scale thermal-hydraulic facility to 

assess the predictions of GOTHIC and TRACE. Their paper illustrated that both the GOTHIC and 

TRACE were able to reasonably predict the heat transfer capability of the PCC as well as the 

influence of non-condensable gas on the system. However, a slight underestimation of the 

condenser performance was obtained with both codes. Blaˇz Mikuˇz et al. [12] established a 

TRACE model for TOPFLOW facility. The main component of TOPFLOW facility was about 8 m 

long vertical tube with an inner diameter of 195.3 mm. The evaporation of liquid water to steam 

caused by depressurization was simulated using two different procedures: from stagnant water and 

during circulating of water in tubes. They also performed the simulation and analysis of the

decompression experiments by using this TRACE model. The TRACE simulations of the first 

procedure (stagnant water at beginning) were in a good agreement with experimental data. The 

TRACE simulations of the second procedure (circulating water in a loop) correctly predicted the

pressure and temperature decrease, but underpredicted the void fraction.

There were three main steps in this study. First, Chinshan NPP TRACE/SNAP model was 

established in this research. In order to assess the system response of TRACE/SNAP model, the 

startup tests and FSAR data were used to compare with the results of TRACE [13]-[14]. Second, 

by using the above TRACE/SNAP model, the URG simulation and analysis under Fukushima-like 

conditions was performed. In this step, the no URG case was also performed. Subsequently, we 

compared the results of these two cases in order to evaluate the URG effectiveness for Chinshan 

NPP. Third, in order to confirm the mechanical property and integrity of fuel rods, the analysis of 

FRAPTRAN is performed. 

Figure 1 Correspondent Relation between NPP Operating States and Operating 

Procedures
3 

2 ULTIMATE RESPONSE GUIDELINE (URG)

According to URG, the core concept is treating compound disaster beyond design basis (blue blocks 

in Figure 1). When Chinshan NPP meets a Fukushima-like accident (the accident with loss of all 

AC power or reactor cooling conditions), EOP/SAMP and URG will be initiated at the same time. 

EOP and SAMP focus on maintaining the reactor core cooling, preventing the release of radioactive 

material, and protecting the property of NPP. The main difference among EOP, SAMP and URG is 

that when the NPP status does not recovery in time, URG must be executed without most 

information of NPP. URG may result in the permanent damage on the reactor of NPP and is an 

irreversible choice, so it needs the senior manager such as the vice president or the plant manager 

to make this decision. The following are the main objectives of URG:

• Maintain the reactor core cooling. 

• Maintain the monitoring functions of the control room. 

• Prevent the release of radioactive material.

• Remove the amount of cumulated hydrogen in building.

• Maintain the spent fuel pool cooling and water level. 

When NPP encounters the Fukushima-like accident, URG will be activated to prevent reactor core 

from being damaged. Once entering the procedure of URG, the NPP reactor will be depressurized 

first, and if the electrical power cannot be recovered before passive reactor core isolation cooling

(RCIC) becomes inoperable, any water available will be injected into the reactor vessel. Figure 2 

depicts the URG procedure. The URG procedure includes several measures to be performed: 

• Perform the controlled depressurization for the reactor to bring down the dome

pressure to 35 kg/cm2 by opening one ADS (automatic depressurization system) when 

the NPP meets the situation “3” (RCIC is available).

• Perform the controlled depressurization for the reactor to bring down the dome

pressure to 15 kg/cm2 by opening one ADS when the NPP meets the situation “3 + 

1“ or “3 + 2” (RCIC is available).

• Prepare alternative water supply which might include service fresh water, reservoir 

gravity injection, fire engine creek, or sea water within the first hour. 

• If NPP status cannot be restored before RCIC is unavailable, perform the fully 

depressurization to 3 kg/cm2 by opening all ADS. The set-up of alternative water supply

must be finished before the fully depressurization is performed.

• Inject the low pressure water into the reactor vessel after the system fully 

depressurizes. 

Perform the containment venting if the containment pressure is beyond design to maintain 

containment integrity.
4 

Figure 2 URG Flowchart
5 

3 CHINSHAN NPP TRACE/SNAP MODEL

SNAP v 2.2.9 and TRACE v 5.0 patch4 are used in this research. The process of Chinshan NPP 

TRACE/SNAP model development is as follows: First, the system and operating data for the startup 

tests and FSAR of Chinshan NPP were collected. Second, several important control systems such 

as recirculation flow control system, pressure control system and feedwater flow control system etc. 

were established by SNAP and TRACE. Next, other necessary components (e.g., RPV (Reactor 

pressure vessel) and main steam piping) were added into the model to complete the TRACE/SNAP

model of Chinshan NPP. Finally, Chinshan TRACE/SNAP model was verified with the cases of 

startup tests and FSAR. The TRACE/SNAP model of Chinshan NPP is presented in Figure 3.

The reactor vessel is divided into 9 axial levels, 4 radial rings, and 2 azimuthal sectors. Six channels 

are used to modeling 408 fuel rods and connect to the vessel’s level 3~4. There are 27 axial nodes 

and 6 radial nodes in one channel. The water rods and partial length rods are also simulated in the 

channels. Six separators connect to the level 6~8 of the vessel. Two recirculation loops are set 

outside the vessel, with a recirculation pump in each loop. In the TRACE model, 10 injection pumps

are merged into an equal injection pump. 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. The reactor vessel connects with four steamlines. Every 

steamline has MSIV, TCV/TSV, and SRVs. The break components are used to simulate the turbines 

(boundary conditions). We also build bypass pipeline. Moreover, the break component at the end 

of bypass valve is used to simulate the condenser. In addition, the containment is composed of 

drywell, wetwell (included suppression pool) and vent header (shown in Figure 3). 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. In the Chinshan NPP TRACE/SNAP 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 components of TRACE.

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 time step of this model was 0.01 sec and the steady state convergence 

was 1×10-4. The results of analysis of TRACE are clearly consistent with FSAR data under the 

steady state condition (See Table 1). 

Table 1 The Comparison of Initial Conditions of FSAR and TRACE

Parameter FSAR TRACE Difference (%)

Power (MWt) 1840 1840 0

Dome pressure (MPa) 6.96 6.96 0

Narrow range water level (m) 0.9 0.9 0

Steam flow (kg/sec) 989.92 989.7 -0.02

Feedwater flow (kg/sec) 987.02 990.6 0.36

Core flow (kg/sec) 6676.74 6665.8 -0.16
6 

Figure 3 Chinshan NPP TRACE Model
7 

4 RESULTS

4.1 Load Rejection Transient

The purpose of this test is to verify the response of the reactor system and the capability of the 

protection system under the transient. The load rejection test was initiated by manually opening 

both generator breakers in April 11, 1978, while the plant was steadily operated at 1775 MWt. Such 

loss of load caused the fast closure of TCV within 0.4 sec. At the same time, the bypass valves 

were activated to open. Closure of the TCV to 90% opening initiated reactor scram with 0.1 second

delay. In response to fast closure of TCV, the system pressure quickly increased to open one of the 

SRVs. No recirculation pump trip occurred as designed and no RCIC/HPCI initiation was observed.

Table 2 compares the load rejection transient’s sequences of NPP test data with TRACE. Their 

sequences are similar. Figure 4~7 show the comparison of results of startup test and TRACE. 

Figure 4 depicts the power curves of startup test and TRACE. The trends of the curves are 

approximately in agreement. As a result, the power calculated from TRACE includes decay heat, 

and stays at a level higher than the neutron flux-determined power from NPP after 3 sec. Figure 5 

shows the steam flow results of startup test and TRACE. Because the TCV closed and the reactor 

scrammed, the steam flow decreased after 1.5 sec. Figure 6 illustrates the dome pressure results 

of startup test and TRACE. The pressure increased because the TCV closed. When the pressure 

was larger than 7.3 MPa, the SRVs opened until the pressure decreased to 7.1 MPa. Figure 7 

compares the NRWL of startup test and TRACE. The result of TRACE is similar to startup test data. 

In summary, the trends of TRACE prediction are consistent with startup test data for this transient

Table 2 The Sequences of Startup Test and TRACE Data for Load Rejection Transient

Action Time (sec)

Startup test TRACE

TCV initiated closure 1.5 1.5

Bypass valve open 1.5 1.5

Reactor scram 1.6 1.6

TCV totally closure 1.9 1.9

Bypass valve totally open 2.1 2.1

SRV open 3.1 3.4

SRV closure 9.9 10.8

End 16 16
8 

0 4 8 12 16

Time (sec)

0

400

800

1200

1600

2000

Power (MWt)

TRACE

Startup test

Figure 4 The Power of Startup Test and TRACE for Load Rejection Transient
9 

0 4 8 12 16

Time (sec)

0

200

400

600

800

1000

Steam flow (kg/sec)

TRACE

Startup test

Figure 5 The Steam Flow of Startup Test and TRACE for Load Rejection Transient
10

0 4 8 12 16

Time (sec)

4

5

6

7

8

Dome pressure (MPa)

TRACE

Startup test

Figure 6 The Dome Pressure of Startup Test and TRACE for Load Rejection Transient
11

0 4 8 12 16

Time (sec)

-0.4

0

0.4

0.8

1.2

Narrow range water level (m)

TRACE

Startup test

Figure 7 The NRWL of Startup Test and TRACE for Load Rejection Transient

4.2 Turbine Trip without Bypass Valve Transient

According to FSAR data, the reactor is initially operating at 1864 MWt. Table 3 shows the turbine 

trip transient’s sequences of FSAR and TRACE. Their sequences are the same. In this 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 assumed failure in this transient.

Figure 8~10 show the results of FSAR and TRACE. Figure 8 illustrates the power curves of FSAR and

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

The TSV closing decreased the reactor’s void fraction which generated the positive reactivity. 

Subsequently, the scram initiated and the power dropped. Figure 9 compares the dome pressure of 

FSAR and TRACE. The trends of the curves are approximately in agreement. The TSV closing caused 

the dome pressure to rise. Subsequently, SRVs opened and the dome pressure declined. Figure 10 

shows the comparison of core flow between FSAR and TRACE. The curve of TRACE is similar to FSAR 

data. Due to the dome pressure increase, it resulted in the core flow rising during 0.5~1.5 sec. The 

decrease of core inlet flow was caused by recirculation pumps trip after 1.5 sec.
12

Table 3 The Sequences of FSAR and TRACE Data for Turbine Trip without Bypass Valve 

Transient

Action Time (sec)

FSAR TRACE

TSV initiated closure 0.1 0.1

Recirculation pump initiated trip 0.6 0.6 

Reactor scram 1.0 1.0

End 6 6

0 2 4 6

Time (sec)

0

1000

2000

3000

4000

5000

Power (MWt)

TRACE

FSAR

Figure 8 The Power of FSAR and TRACE for Turbine Trip without Bypass Valve Transient
13

0 2 4 6

Time (sec)

4

6

8

10

Dome pressure (MPa)

TRACE

FSAR

 

Figure 9 The Dome Pressure of FSAR and TRACE for Turbine Trip without Bypass Valve 

Transient
14

0 2 4 6

Time (sec)

2000

4000

6000

8000

Core flow (kg/sec)

TRACE

FSAR

 

Figure 10 The Core Flow of FSAR and TRACE for Turbine Trip without Bypass Valve 

Transient

4.3 URG Simulation and Analysis

In order to estimate the URG efficiency under Fukushima-like conditions and perform the sensitivity 

study of depressurization and low pressure water injection, there are six cases in this paper. Table 

4 lists the sequence of URG simulation of TRACE under Fukushima-like conditions (base case, 

Case 1). In addition, there are some assumptions have been made in this analysis: (1) the SRVs 

activate in this transient; (2) the decay heat model ANS-73 is used in this transient simulation; (3) 

the low pressure water injection is 28.35 kg/sec. The main actions of URG are the depressurization 

and low pressure water injection of the reactor and containment venting. However, we focus on the 

sensitivity study of depressurization and low pressure water injection, shown in Table 5. Additionally, 

the depressurization of reactor consists of controlled and fully depressurization.

For Case 1 (base case), the simulation of steady state was first performed during 0~300 sec. 

Because the earthquake occurred and tsunami alarm announced (assumed conditions), the reactor 

scrammed at 300 sec. Controlled depressurization of reactor, MSIV closure, recirculation pumps 
15

trip and feedwater trip were also performed at 300 sec. In this controlled depressurization step, 1 

ADS opened which caused that the dome pressure dropped to 35 kg/cm2 and kept at this level 

(shown in Figure 11). RCIC was available at this step which controlled the water level between level 

2 and level 8 during 300~3905 sec (see Figure 12). Subsequently, station blackout (SBO) was 

assumed to occur at 2100 sec. At this step, controlled depressurization of reactor was performed 

again. Therefore, 1 ADS opened and the dome pressure dropped to 15 kg/cm2 (see Figure 11). We 

assumed RCIC failure at 3905 sec. Because RCIC failed, fully depressurization of reactor was 

performed in order to start low pressure water injection. At this step, 5 ADS opened and the dome 

pressure dropped to 3 kg/cm2 which caused the water level decrease (see Figure 11 and 12). 

Subsequently, low pressure water injected to the reactor at 5000 sec. The sources of low pressure 

water were assumed from service fresh water, reservoir gravity injection, fire engine creek or 

seawater (see Figure 2). The water level went up and PCT decreased after the low pressure water 

injection (see Figure 12 and 13). Finally, the transient finished at 8000 sec. The above results show 

that the URG can keep the PCT below the criteria 1088.7 K under Fukushima-like conditions (shown 

in Figure 13). Figure 14~17 show the containment parameters results of TRACE. According to the 

FSAR, the pressure and temperature of containment should be lower than 0.487 MPa and 444.26 

K. In this case, the results of TRACE didn’t reach these criteria. The pressure of drywell and wetwell 

went up due to the large amount of steam was released into suppression pool (see Figure 14 and 

15). The large amount of steam also caused that the temperature of containment rose (see Figure

16 and 17). In addition, due to containment venting was performed at 3905 sec, the pressure and 

temperature of drywell and wetwell dropped at this time (shown in Figure 14 ~ 16).

No URG case is Case 2. The dome pressure of Case 2 was kept at about 6.8~7.6 MPa and there 

was an oscillation in Case 2 due to SRVs activated (see Figure 11). After RCIC failure, the water 

level started to decrease and was lower than TAF at about 7500 sec (shown in Figure 12). The PCT 

started to increase after the water level lower than TAF and reached 1088.7 K at 16225 sec. It 

indicated that the zirconium-water reaction was able to generate. The above results depicted that 

Chinshan NPP was not at the safe situation. Furthermore, the pressure and temperature of 

containment were lower than 0.487 MPa and 444.26 K in this case (see Figure 14~17). It indicates 

that the safety issue may occur in the reactor core and not in the containment before 18000 sec. In 

the comparison of Case 1 and 2, it shows clearly that URG can keep Chinshan NPP at the safe 

situation under Fukushima-like conditions. 

Figure 18~24 depict the comparison of Case 1, 3 and 4. Because there was no controlled 

depressurization for Case 3 and 4, the SRVs were activated during 300~3905 sec (shown in Figure

18). If there was only the fully depressurization of reactor in URG (no controlled depressurization), 

TRACE results depicted that the water level dropping was larger (see Figure 19) than the controlled 

depressurization cases. It indicates that PCT increases faster. Furthermore, if no water injection 

was also in the above cases, the PCT was larger than 1088.7 K (see Case 4, Figure 20) which 

indicated that Chinshan NPP was not at the safety situation. Figure 21~24 illustrate the containment 

parameters results of TRACE. In Case 3 and 4, the results of TRACE didn’t reach the criteria (0.487 

MPa and 444.26 K). The pressure of drywell and wetwell went up due to the large amount of steam 

was released into suppression pool (see Figure 21 and 22). Subsequently, the large amount of 

steam caused that the temperature of containment rose (see Figure 23 and 24). Additionally, due 

to containment venting was performed at 3905 sec, the pressure and temperature of drywell and 

wetwell dropped at this time (shown in Figure 21 ~ 23). In the comparison of Case 1, 3 and 4, due 

to no controlled depressurization in Case 3 and 4, the pressure and temperature of containment for 

Case 1 were larger than Case 3 and 4 during 300~3905 sec.

Figure 25~31 illustrate the comparison of Case 1, 5 and 6. Comparing Case 1 and 5, if there was 

only the controlled depressurization of reactor in URG (no fully depressurization and low pressure 
16

water injection), TRACE results presented that the water level was lower than TAF at 7700 sec and 

the PCT was larger than 1088.7 K at 14030 sec. Comparing Case 1 and 6, if there was no water 

injection in URG, TRACE results indicated that PCT went up after the depressurization of reactor 

(see Case 6, Figure 27). It presented that the zirconium-water reaction generated and Chinshan 

NPP was not at the safe situation. The above results also indicated that if the NPP want to be at the 

safe situation, the low pressure injection must be performed after the depressurization of reactor 

finishes. Figure 28~31 show the containment parameters results of TRACE. In Case 5 and 6, the 

results of TRACE didn’t reach the criteria (0.487 MPa and 444.26 K). The pressure of drywell and 

wetwell went up due to the large amount of steam was released into suppression pool (see Figure

28 and 29). Subsequently, the large amount of steam caused that the temperature of containment 

increased (see Figure 30 and 31). In addition, due to containment venting for Case 6 was performed 

at 3905 sec, the pressure and temperature of drywell and wetwell dropped at this time (shown in 

Figure 28 ~ 30). In the comparison of Case 1, 5 and 6, due to no fully depressurization in Case 5, 

the temperature of suppression pool for Case 1 and 6 were larger than Case 5 after 3905 sec (see 

Figure 31).

In summary, Table 5 shows the time point of Case 1~6 for the water level lower than TAF and the 

PCT larger than 1088.7 K. It indicates clearly that Case 4 is the most severe case.

Additionally, the sensitivity study of the required raw water injection for URG was performed in this 

research. The low pressure water injection is 28.35 kg/sec for Case 1. Therefore, we assumed that 

the low pressure water injection is 14.18 kg/sec for Case 1-1; 10.63 kg/sec for Case 1-2; 7.09 kg/sec 

for Case 1-3. Figure 32 depicts the water level of Case 1, 1-1, 1-2, and 1-3. It shows that the water 

level of Case 1-3 was always lower than TAF after the water injection. The water level of Case 1-2 

reached the TAF after 8000 sec. But its water level was lower than TAF during 4620~8000 sec. The 

water level of Case 1-1 was kept at TAF after the water injection and was larger than TAF after 

6500 sec. Therefore, we think that the required raw water injection must be larger than 14.18 kg/sec.

In order to study the mechanical property and integrity of fuel rods, FRAPTRAN analysis was 

performed in this study. In this step, we only analyzed the no URG case. The URG case is not 

considered in this step because its PCT is lower than 1088.7 K. Figure 33~39 depict the results of 

FRAPTRAN. FRAPTRAN output file depicted that the max PCT was in the axial node 15. Hence, 

Figure 33~39 only present the results of the axial node 15. The zirconium-water reaction occurred 

at about 16000 sec (shown in Figure 33). At this time, the core water level was lower than bottom 

of active fuel (BAF) and PCT reached 1088.7 K. Because zirconium-water reaction occurred, the 

oxide thickness of cladding went up. According to 10 CFR 50.46 rule [15], the increasing 

oxide thickness of cladding should be less than 17%, Figure 34 shows the cladding oxide thickness 

is over the critical value (0.103 mm). Figure 35 and 36 present the structural radial gap and gap gas 

pressure results of FRAPTRAN. The structural radial gap became zero at about 14500 sec. It 

indicated that PCMI (Pellet Cladding Mechanical Interaction) generated. The gap gas pressure 

increases when structural radial gap decreases. Figure 37 and 38 illustrates the cladding hoop 

strain and stress results of FRAPTRAN. When the water level was lower than TAF, there was the 

variation in the cladding hoop strain and stress. As PCT increased, cladding hoop strain and stress 

also went up initially. Furthermore, when PCMI occurred, cladding hoop strain and stress had the 

more variation. According to NUREG-0800 Standard Review Plan [16], the fuel enthalpy should be 

lower than 7.11 x 105 J/kg. The result of FRAPTRAN did not reach this criterion (see Figure 39) but 

was close to this value.
17

Table 4 The Sequences of URG Case (Case 1)

Action Time (sec)

Start 0

Reactor scrams (because earthquake occurs and tsunami alarm 

announces),

MSIV closes,

Recirculation pumps trip,

Feedwater trip,

Controlled depressurization of reactor (open 1 ADS, the pressure 

drops to 35 kg/cm2

) 

(RCIC is available)

300

SBO (loss of all AC power),

Controlled depressurization of reactor (open 1 ADS, the pressure 

drops to 15 kg/cm2)

2100

RCIC is not available,

Fully depressurization of reactor (open 5 ADS, the pressure drops to 3 

kg/cm2

),

Containment venting

3905

Low pressure water injection 5000

Table 5 The Sensitivity Study of URG

Case 1 Case 2 Case 3 Case 4 Case 5 Case 6

Controlled 

depressurization of 

the reactor

○ x x x ○ ○

Fully 

depressurization of

the reactor

○ x ○ ○ x ○

Low pressure water 

injection ○ x ○ x x x 

The water level lower 

than TAF

4620~

5060 

sec

7500~

18000 

sec

4030~

8000 

sec

4030~

6880 

sec

7700~

15000 

sec

4680~

12100

sec

The PCT larger than 

1088.7 K ---

16225~

18000

sec

5620~

6740 

sec

5700~

6880 

sec

14030~

15000 

sec

9865~

12100 

sec

Transient end 8000

sec

18000

sec

8000

sec

6880 

sec

15000 

sec

12100 

sec

○: Perform this action

x: Don’t perform this action
18

0 3000 6000 9000 12000 15000 18000

Time (sec)

0

2

4

6

8

Dome pressure (MPa)

Case 1

Case 2

Figure 11 The Dome Pressure Comparison of Case 1 and 2
19

0 3000 6000 9000 12000 15000 18000

Time (sec)

0

4

8

12

16

Water level (m)

Case 1

Case 2

BAF

TAF

L 2

L 8

 

Figure 12 The Water Level Comparison of Case 1 and 2
20

0 3000 6000 9000 12000 15000 18000

Time (sec)

200

400

600

800

1000

1200

PCT (K)

Case 1

Case 2

1088.7 K

 

Figure 13 The PCT Comparison of Case 1 and 2
21

0 3000 6000 9000 12000 15000 18000

Time (sec)

0.1

0.12

0.14

0.16

0.18

Drywell pressure (MPa)

Case 1

Case 2

 

Figure 14 The DRYWELL PRESSURE Comparison of Case 1 and 2
22

0 3000 6000 9000 12000 15000 18000

Time (sec)

0.08

0.12

0.16

0.2

0.24

0.28

Wetwell pressure (MPa)

Case 1

Case 2

 

Figure 15 The Wetwell Pressure Comparison of Case 1 and 2
23

0 3000 6000 9000 12000 15000 18000

Time (sec)

320

340

360

380

Drywell temperature (K)

Case 1

Case 2

 

Figure 16 The Drywell Temperature Comparison of Case 1 and 2
24

0 3000 6000 9000 12000 15000 18000

Time (sec)

300

320

340

360

380

Suppression pool temperature (K)

Case 1

Case 2

 

Figure 17 The Suppression Pool Temperature Comparison of Case 1 and 2
25

0 3000 6000 9000

Time (sec)

0

2

4

6

8

Dome pressure (MPa)

Case 1

Case 3

Case 4

 

Figure 18 The Dome Pressure Comparison of Case 1, 3 and 4 
26

0 3000 6000 9000

Time (sec)

4

8

12

16

Water level (m)

Case 1

Case 3

Case 4

BAF

TAF

L 2

L 8

 

Figure 19 The Water Level Comparison of Case 1, 3 and 4 
27

0 3000 6000 9000

Time (sec)

0

400

800

1200

1600

2000

PCT (K)

Case 1

Case 3

Case 4

1088.7 K

 

Figure 20 The PCT Comparison of Case 1, 3 and 4 
28

0 3000 6000 9000

Time (sec)

0.1

0.12

0.14

0.16

Drywell pressure (MPa)

Case 1

Case 3

Case 4

 

Figure 21 The Drywell Pressure Comparison of Case 1, 3 and 4 
29

0 3000 6000 9000

Time (sec)

0.08

0.12

0.16

0.2

0.24

0.28

Wetwell pressure (MPa)

Case 1

Case 3

Case 4

 

Figure 22 The Wetwell Pressure Comparison of Case 1, 3 and 4 
30

0 3000 6000 9000

Time (sec)

320

340

360

380

Drywell temperature (K)

Case 1

Case 3

Case 4

 

Figure 23 The Drywell Temperature Comparison of Case 1, 3 and 4 
31

0 3000 6000 9000

Time (sec)

300

320

340

360

380

Suppression pool temperature (K)

Case 1

Case 3

Case 4

 

Figure 24 The Suppression Pool Temperature Comparison of Case 1, 3 and 4 
32

0 3000 6000 9000 12000 15000

Time (sec)

0

2

4

6

8

Dome pressure (MPa)

Case 1

Case 5

Case 6

 

Figure 25 The Dome Pressure Comparison of Case 1, 5 and 6 
33

0 3000 6000 9000 12000 15000

Time (sec)

4

6

8

10

12

14

16

Water level (m)

Case 1

Case 5

Case 6

BAF

TAF

L 2

L 8

 

Figure 26 The Water Level Comparison of Case 1, 5 and 6 
34

0 3000 6000 9000 12000 15000

Time (sec)

0

400

800

1200

1600

2000

PCT (K)

Case 1

Case 5

Case 6

1088.7 K

 

Figure 27 The PCT Comparison of Case 1, 5 and 6 
35

0 3000 6000 9000 12000 15000

Time (sec)

0.1

0.12

0.14

0.16

Drywell pressure (MPa)

Case 1

Case 5

Case 6

 

Figure 28 The Drywell Pressure Comparison of Case 1, 5 and 6 
36

0 3000 6000 9000 12000 15000

Time (sec)

320

340

360

380

Drywell temperature (K)

Case 1

Case 5

Case 6

 

Figure 29 The Drywell Temperature Comparison of Case 1, 5 and 6 
37

0 3000 6000 9000 12000 15000

Time (sec)

0.08

0.12

0.16

0.2

0.24

0.28

Wetwell pressure (MPa)

Case 1

Case 5

Case 6

 

Figure 30 The Wetwell Pressure Comparison of Case 1, 5 and 6 
38

0 3000 6000 9000 12000 15000

Time (sec)

300

320

340

360

380

Suppression pool temperature (K)

Case 1

Case 5

Case 6

 

Figure 31 The Suppression Pool Temperature Comparison of Case 1, 5 and 6 
39

0 2000 4000 6000 8000

Time (sec)

0

4

8

12

16

Water level (m)

Case 1

Case 1-1

Case 1-2

Case 1-3

BAF

TAF

L 2

L 8

 

Figure 32 The Water Level Results for The Sensitivity Study 
40

0 4000 8000 12000 16000 20000 24000

Time (sec)

0

0.2

0.4

0.6

0.8

Water metal reaction energy (kW/m)

 

Figure 33 The Water Metal Reaction Energy Result of FRAPTRAN for Case 2
41

0 4000 8000 12000 16000 20000 24000

Time (sec)

0

0.1

0.2

0.3

0.4

0.5

Oxide thickness (mm)

Figure 34 The Oxide Thickness Result of FRAPTRAN for Case 2
42

0 4000 8000 12000 16000 20000 24000

Time (sec)

0

0.02

0.04

0.06

Structural radial gap (mm)

Figure 35 The Structural Radial Gap Result of Fraptran for Case 2
43

0 4000 8000 12000 16000 20000 24000

Time (sec)

1

1.5

2

2.5

3

3.5

4

Gap gas pressure (MPa)

 

Figure 36 The Gap Gas Pressure Result of FRAPTRAN for Case 2
44

0 4000 8000 12000 16000 20000 24000

Time (sec)

-0.008

-0.006

-0.004

-0.002

0

0.002

0.004

Cladding hoop strain

 

Figure 37 The Cladding Hoop Strain Result of FRAPTRAN for Case 2 
45

0 4000 8000 12000 16000 20000 24000

Time (sec)

-60

-40

-20

0

20

Cladding hoop stress (MPa)

 

Figure 38 The Cladding Hoop Stress Result of FRAPTRAN for Case 2
46

0 4000 8000 12000 16000 20000 24000

Time (sec)

0.0E+000

1.0E+005

2.0E+005

3.0E+005

4.0E+005

5.0E+005

Fuel enthalpy (J/kg)

Figure 39 The Fuel Enthalpy Result of FRAPTRAN for Case 2
47

5 CONCLUSIONS 

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

test (load rejection transient) and FSAR (turbine trip without bypass valve transient) data were used 

to compare with the results of TRACE. The compared results indicate that the trends of parameters 

are similar. By the above compared results, it depicts that there is a respectable accuracy in 

Chinshan NPP TRACE/SNAP model and it also shows that Chinshan NPP TRACE/SNAP model is 

satisfying for the purpose of Chinshan NPP safety analyses with confidence.

This study used the above model to estimate the effectiveness of URG for Chinshan NPP. TRACE 

analysis results show that the URG can keep the PCT below the criteria 1088.7 K under Fukushimalike conditions. It indicates that Chinshan NPP is at the safe situation. On the sensitivity study of 

URG, the summary is as follows:

• In no water injection cases, if the controlled and fully depressurization of reactor are

both performed, its PCT will increase faster than the no depressurization case.

• The set-up of raw water must be finished as fast as possible. If RCIC is unavailable,

Chinshan NPP can perform the fully depressurization of the reactor immediately and

inject the low pressure water to the reactor.

• The controlled depressurization should be performed before the fully depressurization.

It can mitigate the dropping of the water level when Chinshan NPP runs the fully

depressurization.

• The required raw water injection must be larger than 14.18 kg/sec.

• In all cases, the results of TRACE did not reach the criteria (0.487 MPa and 444.26 K)

of containment. It indicates that the safety issue may occur in the reactor core earlier

than in the containment.

Additionally, the results of FRAPTRAN also show that the integrity of fuel was not kept for Case 2.

49

6 REFERENCES 

[1] Taiwan Power Company, 2012. Chinshan nuclear power plant ultimate response 

guideline, No. 1451. 

[2] Liang, K.S., Chiang, S.C., Hsu, Y.F., Young, H.J., Pei, B.S., Wang, L.C., 2012. The 

ultimate emergency measures to secure a NPP under an accidental condition with no 

designed power or water supply. Nucl. Eng. Des. 253, 259-268.

[3] Liu, K.H., Hwang, S.L., 2012. Human performance evaluation: The procedures of 

ultimate response guideline for nuclear power plants. Nucl. Eng. Des. 253, 259-268. 

[4] U.S. NRC, 2014. TRACE V5.840 user’s manual.

[5] U.S. NRC, 2010. TRACE V5.0 assessment manual.

[6] Mascari, F., Vella, G., Woods, B.G., Welter, K., Pottorf, J., Young, E., Adorni, M., D’

auria, F., 2011. Sensitivity analysis of the MASLWR helical coil steam generator using 

TRACE. Nucl. Eng. Des. 241, 1137-1144. 

[7] Freixa, J., Manera, A., 2011. Verification of a TRACE EPRTM model on the basis of a 

scaling calculation of an SBLOCA ROSA test. Nucl. Eng. Des. 241, 888-896. 

[8] Nikitin, K., Manera, A., 2011. Analysis of an ADS spurious opening event at a BWR/6 

by means of the TRACE code. Nucl. Eng. Des. 241, 2240-2247. 

[9] Pollman, A., Vierow, K., di Marzo, M., 2012. Analysis of rapid-condensation transient 

using TRACE. Nucl. Eng. Des. 250, 512-519. 

[10] Berar, O.-A., Proˇsek, A., Mavko, B., 2013. RELAP5 and TRACE assessment of the 

Achilles natural reflood experiment. Nucl. Eng. Des. 261, 306-316. 

[11] Papini, D., Adamsson, C., Andreani, M., Prasser, H.-M., 2014. Assessment of GOTHIC 

and TRACE codes against selected PANDA experiments on a passive containment 

condenser. Nucl. Eng. Des. 278, 542-557.

[12] Mikuˇz, B., Tiselj, I., Beyer, M., Lucas, D., 2015. Simulations of flashing experiments in 

TOPFLOW facility with TRACE code. Nucl. Eng. Des. 283, 60-70.

[13] Taiwan Power Company, 1999. Final safety analysis report for Chinshan nuclear power 

station units 1&2. 

[14] Wang, J.R., Jang, S.C., Chen, B.S., Wang S.F., 1989. Chinshan nuclear power plabt 

startup test documentation. INER report, INER-T1322.

[15] U.S. NRC, 2011. 10CFR50.46c, ML110970044. 

[16] U.S. NRC, 2007. Standard review plan, NUREG-0800. 

TRACE/SNAP Model Establishment of Chinshan Nuclear Power Plant for 

Ultimate Response Guideline

NUREG/IA-0478

Jong-Rong Wang*, Chunkuan Shih*, Jung-Hua Yang*, Hsiung-Chih Chen*, ShaoWen Chen*, Show-Chyuan Chiang**, Tzu-Yao Yu**

*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

K. Tien, NRC Project Manager

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

In this research, the TRACE/SNAP model of Chinshan BWR/4 nuclear power plant (NPP) was established for the simulation 

and analysis of ultimate response guideline (URG). The main actions of URG are the depressurization and low pressure 

water injection of the reactor and containment venting. This research focuses to assess the URG utility of Chinshan NPP 

under Fukushima-like conditions. This study consists of three steps. The first step is the establishment of Chinshan NPP 

TRACE/SNAP model. In order to evaluate the system response of TRACE/SNAP model, startup tests and FSAR data were 

used to compare with the results of TRACE. The second step is the URG simulation and analysis under Fukushima-like 

conditions by using Chinshan NPP TRACE/SNAP model. In this step, the no URG case was also performed in order to 

evaluate the URG effectiveness of Chinshan NPP. In addition, the sensitivity study of URG and the required raw water 

injection were also performed. According to TRACE analysis results, the URG can keep the peak cladding temperature 

(PCT) below the criteria 1088.7 K under Fukushima-like conditions. It indicates that Chinshan NPP can be controlled in a 

safe situation. If Chinshan NPP does not perform the URG under Fukushima-like conditions, the water level may drop lower 

than TAF (top of active fuel) which means a safety issue about the fuel rods may be generated. In order to confirm the 

mechanical property and integrity of fuel rods, the final step is the analysis of FRAPTRAN.

Ultimate Response Guideline (URG)

Fukushima-like Events 

TRACE/SNAP/FRAPTRAN

Peak Cladding Temperature (PCT)

Top of Active Fuel (TAF)

Chinshan Nuclear Power Plant (NPP)

May 2017

Technical

**Department of Nuclear Safety, Taiwan Power 

Company

242, Section 3, Roosevelt Rd., Zhongzheng 

District, Taipei, Taiwan



NUREG/IA-0478 TRACE/SNAP Model Establishment of Chinshan Nuclear Power Plant for Ultimate Response 

Guideline

May 2017