NUREG/IA-0450 

International 

 Agreement Report

The Development and Application 

of Kuosheng (BWR/6) Nuclear 

Power Plant TRACE/SNAP Model 

 

Prepared by: 

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

Chia-Chuan Liu**

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

**Department of Nuclear Safety, Taiwan Power Company 

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

K. Tien, NRC Project Manager 

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2014

Date Published: December 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
NRC Reference Material 

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

NRC’s Public Electronic Reading Room at 

http://www.nrc.gov/reading-rm.html. Publicly released 

records include, to name a few, NUREG-series 

publications; Federal Register notices; applicant, 

licensee, and vendor documents and correspondence; 

NRC correspondence and internal memoranda; bulletins 

and information notices; inspection and investigative 

reports; licensee event reports; and Commission papers 

and their attachments. 

NRC publications in the NUREG series, NRC 

regulations, and Title 10, “Energy,” in the Code of

Federal Regulations may also be purchased from one 

of these two sources. 

1. The Superintendent of Documents

 U.S. Government Printing Office

Mail Stop SSOP

Washington, DC 20402–0001

Internet: bookstore.gpo.gov

Telephone: 202-512-1800

Fax: 202-512-2250 

2. The National Technical Information Service 

Springfield, VA 22161–0002

www.ntis.gov

1–800–553–6847 or, locally, 703–605–6000 

A single copy of each NRC draft report for comment is 

available free, to the extent of supply, upon written 

request as follows:

Address: U.S. Nuclear Regulatory Commission

 Office of Administration 

Publications Branch

 Washington, DC 20555-0001

E-mail: DISTRIBUTION.RESOURCE@NRC.GOV

Facsimile: 301–415–2289 

Some publications in the NUREG series that are 

posted at NRC’s Web site address 

http://www.nrc.gov/reading-rm/doc-collections/nuregs 

are updated periodically and may differ from the last 

printed version. Although references to material found on 

a Web site bear the date the material was accessed, the 

material available on the date cited may subsequently be 

removed from the site.

Non-NRC Reference Material 

Documents available from public and special technical 

libraries include all open literature items, such as books, 

journal articles, transactions, Federal Register notices, 

Federal and State legislation, and congressional reports. 

Such documents as theses, dissertations, foreign reports 

and translations, and non-NRC conference proceedings 

may be purchased from their sponsoring organization. 

Copies of industry codes and standards used in a 

substantive manner in the NRC regulatory process are 

maintained at—

The NRC Technical Library

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852–2738 

These standards are available in the library for reference 

use by the public. Codes and standards are usually 

copyrighted and may be purchased from the originating 

organization or, if they are American National Standards, 

from—

American National Standards Institute

11 West 42nd Street

New York, NY 10036–8002

www.ansi.org

212–642–4900 

AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

Legally binding regulatory requirements are stated only 

in laws; NRC regulations; licenses, including technical 

specifications; or orders, not in NUREG-series 

publications. The views expressed in contractorprepared publications in this series are not necessarily 

those of the NRC.

The NUREG series comprises (1) technical and 

administrative reports and books prepared by the staff 

(NUREG–XXXX) or agency contractors (NUREG/CR–

XXXX), (2) proceedings of conferences (NUREG/CP–

XXXX), (3) reports resulting from international 

agreements (NUREG/IA–XXXX), (4) brochures 

(NUREG/BR–XXXX), and (5) compilations of legal 

decisions and orders of the Commission and Atomic and 

Safety Licensing Boards and of Directors’ decisions

under Section 2.206 of NRC’s regulations (NUREG–

0750).

DISCLAIMER: This report was prepared as an account 

of work sponsored by an agency of the U.S. 

Government. Neither the U.S. Government nor any 

agency thereof, nor any employee, makes any warranty, 

expressed or implied, or assumes any legal liability or 

responsibility for any third party’s use, or the results of 

such use, of any information, apparatus, product, or 

process disclosed in this publication, or represents that 

its use by such third party would not infringe privately 

owned rights.
NUREG/IA-0450

International 

 Agreement Report

The Development and Application of Kuosheng (BWR/6) 

Nuclear Power Plant TRACE/SNAP Model

Prepared by:

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

Chia-Chuan Liu**

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

**Department of Nuclear Safety, Taiwan Power Company

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

K. Tien, NRC Project Manager

 Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2014

Date Published: December 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

Kuosheng Nuclear Power Plant (NPP) is the second NPP in Taiwan which is the BWR/6 plant. 

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

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

analysis results of Final Safety Analysis Report (FSAR) and startup tests data to assess the 

Kuosheng NPP TRACE/SNAP model. The load rejection and a feedwater pump trip transients

were selected to validate the Kuosheng NPP TRACE/SNAP model. The trends of TRACE 

analysis results were consistent with the FSAR and startup tests data. It indicates that there is a 

respectable accuracy in the Kuosheng NPP TRACE/SNAP model. Besides, this research also 

focuses on the application of the Kuosheng 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 Kuosheng NPP TRACE/SNAP model to perform the core shroud leakage 

transients for Kuosheng NPP safety analysis. The TRACE analysis results show that the pure

core shroud leakage transient wasn’t influence the Kuosheng NPP safety. However, the core 

shroud leakage + station blackout (SBO) / steamline break (loss of coolant accident, LOCA) 

transient caused the cladding temperature larger than 1088 K and affected the Kuosheng 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 old 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 Kuosheng NPP has been built. In this report, this 

study uses FSAR and startup tests data to assess the Kuosheng 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 KUOSHENG (BWR/6) NPP............................................................................2-1

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

3.1 Startup Test : Generator Load Rejection with Bypass Valves .............................3-2

3.2 Startup Test : One Feedwater Pump Trip.............................................................3-10

3.3 FSAR Case : Generator Load Rejection with Bypass Valves .............................3-16

3.4 Core Shroud Leakage Analysis ............................................................................3-22

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

Kuosheng NPP..................................................................................................2-2

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

Figure 3 The animation model of Kuosheng NPP..........................................................2-4

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

Figure 5 The channel condition of Kuosheng NPP .......................................................2-6

Figure 6 The recirculation pump condition of Kuosheng NPP.....................................2-7

Figure 7 The jetpump condition of Kuosheng NPP.......................................................2-8

Figure 8 The SRV condition of Kuosheng NPP..............................................................2-9

Figure 9 The turbine condition of Kuosheng NPP.......................................................2-10

Figure 10 The upper pool condition of Kuosheng NPP ................................................2-11

Figure 11 The reactor building condition of Kuosheng NPP........................................2-12

Figure 12 The wetwell and suppression pool condition of Kuosheng NPP ................2-13

Figure 13 The drywell condition of Kuosheng NPP ......................................................2-14

Figure 14 The power condition of Kuosheng NPP ........................................................2-15

Figure 15 The power data of TRACE and startup test.....................................................3-4

Figure 16 The dome pressure data of TRACE and startup test......................................3-5

Figure 17 The steam flow data of TRACE and startup test.............................................3-6

Figure 18 The feedwater flow data of TRACE and startup test.......................................3-7

Figure 19 The narrow range water level (NRWL) data of TRACE and startup test........3-8

Figure 20 The core inlet flow data of TRACE and startup test .......................................3-9

Figure 21 The core inlet flow data of TRACE and startup test .....................................3-12

Figure 22 The power data of TRACE and startup test...................................................3-13

Figure 23 The feedwater flow data of TRACE and startup test.....................................3-14

Figure 24 The NRWL data of TRACE and startup test...................................................3-15

Figure 25 The power data of TRACE and FSAR ............................................................3-18

Figure 26 The dome pressure data of TRACE and FSAR .............................................3-19

Figure 27 The steam flow data of TRACE and FSAR.....................................................3-20

Figure 28 The core inlet flow data of TRACE and FSAR ...............................................3-21

Figure 29 The power results ...........................................................................................3-24

Figure 30 The NRWL results ...........................................................................................3-25

Figure 31 The vessel water level results........................................................................3-26

Figure 32 The fuel rod temperature results ...................................................................3-27

Figure 33 The dome pressure results ............................................................................3-28

Figure 34 The animation model of Kuosheng NPP for core shroud leakage 

transient ..........................................................................................................3-30
x

 
xi

TABLES

Page

Table 1 The comparison of initial conditions between FSAR and TRACE data ............3-1

Table 2 The comparison of sequences between startup test and TRACE data ............3-3

Table 3 The comparison of sequences between startup test and TRACE...................3-11

Table 4 The comparison of sequences between FSAR and TRACE ............................3-17

Table 5 The sequences of case 1~6 for core shroud leakage transient.......................3-23
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 Kuosheng

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. TRACE also can provide greater simulation 

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

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

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

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

2894MWt. After the project of MURPU (Measurement Uncertainty Recovery Power Uprate) for 

Kuosheng NPP, the operating power is 101.7% of the original designed rated power, which is 

2943 MWt now. Unit 1 will start SPU (Stretch Power Uprate) from Cycle 24 and Unit 2 will start 

SPU from Cycle 23. The operating power will be 104.7% of the OLTP (Original Licensed Thermal 

Power), which will be 3030 MWt. In order to check the system response of the Kuosheng NPP 

TRACE/SNAP model, this study uses the analysis results of FSAR and startup tests data to 

assess the Kuosheng NPP TRACE/SNAP model. The load rejection and a feedwater pump trip 

transients were selected to validate the Kuosheng NPP TRACE/SNAP model. The trends of 

TRACE analysis results were consistent with the FSAR and startup tests data. It indicates that 

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

research also focuses on the application of the Kuosheng 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 Kuosheng NPP TRACE/SNAP model to perform the core 

shroud leakage transients for Kuosheng NPP safety analysis. The TRACE analysis results show 

that the pure core shroud leakage transient wasn’t influence the Kuosheng NPP safety. 

However, the core shroud leakage + SBO/LOCA transient caused the cladding temperature 

larger than 1088 K and affected the Kuosheng NPP safety.
xiv 

 
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

MURPU Measurement Uncertainty Recovery Power Uprate

NPP Nuclear Power Plant 

NRC Nuclear Regulatory Commission

NRWL Narrow Range Water Level

RPV Reactor Pressure Vessel

OLTP Original Licensed Thermal Power

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

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

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

two loops of recirculation piping and four main steam lines, with the thermal rated power of 2894

MWt. After the project of MURPU for Kuosheng NPP, the operating power is 101.7% of the 

original designed rated power, which is 2943 MWt now. Besides, Unit 1 will start SPU from Cycle 

24 and Unit 2 will start SPU from Cycle 23. The operating power will be 104.7% of the OLTP, 

which will be 3030 MWt. 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 Kuosheng NPP TRACE/SNAP model. In our 

previous study [2], the preliminary TRACE/SNAP model of Kuosheng NPP was established. The 

Kuosheng NPP TRACE/SNAP model includes one 3-D vessel, six channels which are used to 

simulate 624 fuel bundles, four steamlines, and 16 SRVs components, etc.. In order to check the 

system response of the Kuosheng NPP TRACE/SNAP model, this study uses the FSAR and 

startup tests data to assess the Kuosheng NPP TRACE/SNAP model. The load rejection and a 

feedwater pump trip transients were selected to validate the Kuosheng NPP TRACE/SNAP 

model.

Besides, this research also focuses on the application of the Kuosheng 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 

Kuosheng NPP was performed by the Kuosheng NPP TRACE/SNAP model in this study. 

Besides, refer to the conditions of Fukushima NPP event, the core shroud leakage + SBO/LOCA 

transients were also analyzed by TRACE for Kuosheng NPP.
1-2

 
2-1

2. MODEL OF KUOSHENG (BWR/6) NPP

The code versions adopted in this research are SNAP v 2.2.7 and TRACE v 5.0p3. The process

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

system and operating data for the cases of FSAR and startup tests of Kuosheng NPP are 

collected [1]-[5]. 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 Kuosheng NPP. Finally, the Kuosheng TRACE/SNAP model is verified with the cases 

of FSAR and startup tests. The TRACE/SNAP model of Kuosheng NPP is presented in Fig. 2.

SNAP also can use the TRACE results data to make an animation for transient, such as Fig. 3.

Kuosheng NPP TRACE/SNAP model has been built according to the FSAR, startup tests data,

design documents, and TRACE manuals [1]-[5]. Kuosheng NPP reactor, divided into two

azimuthal sectors, four radial rings, eleven axial levels, altogether eighty-eight cells, are 

simulated by the 3-D vessel component (Fig. 4). Six channels, which are one dimensional 

component, are used for simulating 624 fuel bundles (Fig. 5). Full length fuel rods, partial length 

fuel rods and water rods are also simulated in the channel components. Two recirculation loops 

are set outside the reactor, with a recirculation pump in each loop (Fig. 6). 10 groups of jet

pumps are merged into an equal jet pump (Fig. 7). Four steam lines connected with the vessel 

and each steam line had one MSIV (main steamline isolation valve), several SRVs (safety relief 

valves), one TCV (turbine control valve), and one TSV (turbine stop valve). The bypass valve 

(BPV) was also simulated in this mode. We use valve components to simulate the MSIV, SRVs, 

TCV, TSV and BPV (Fig. 8). The critical flow models for the MSIVs, SRVs, TCVs, TSVs, and 

BPV have been considered in our analysis. The break components were used to simulate the 

boundary conditions of turbine and condenser (Fig. 9). The containment of Kuosheng NPP was 

also simulated in the TRACE/SNAP model. The containment was composed of drywell, wetwell, 

suppression pool, vent annulus, horizontal vent, upper pool, and reactor building which were 

shown in Fig. 2 and 10~13. In the Kuosheng NPP TRACE/SNAP model, there are three 

simulation control systems included (1) feedwater flow control system, (2) steam bypass and 

pressure control system and (3) recirculation flow control system. Besides, in Kuosheng 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 (Fig. 14). The initial conditions of Kuosheng NPP were also presented in Fig. 3. 
2-2

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

Kuosheng NPP

Gathering the design data, control system, startup tests and FSAR 

data of Kuosheng NPP

Building the TRACE/SNAP model with control 

systems by TRACE/SNAP.

The TRACE/SNAP model for Kuosheng NPP

Verifying the TRACE/SNAP model for Kuosheng 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 Kuosheng NPP
2-4

Figure 3 The animation model of Kuosheng NPP
2-5

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

 

Figure 5 The channel condition of Kuosheng NPP
2-7

Figure 6 The recirculation pump condition of Kuosheng NPP
2-8

Figure 7 The jetpump condition of Kuosheng NPP
2-9

Figure 8 The SRV condition of Kuosheng NPP
2-10

Figure 9 The turbine condition of Kuosheng NPP
2-11

Figure 10 The upper pool condition of Kuosheng NPP
2-12

Figure 11 The reactor building condition of Kuosheng NPP
2-13

Figure 12 The wetwell and suppression pool condition of Kuosheng NPP
2-14

 

Figure 13 The drywell condition of Kuosheng NPP
2-15

Figure 14 The power condition of Kuosheng NPP
2-16

 
3-1

3. RESULTS 

Before the transient calculation of Kuosheng 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).

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

Parameter

FSAR TRACE

Point kinetics Difference (%)

Power

(MWt) 2894 2894 0

Dome Pressure

(MPa) 7.3 7.3 0

Feedwater Flow

(kg/sec) 1647 1652 -0.3

Steam Flow

(kg/sec) 1647 1652 -0.3

Core inlet flow

(kg/sec) 10647 10521 1.2
3-2

3.1 Startup Test : Generator Load Rejection with Bypass Valves

This test was performed in November 11, 1981 and the initial power was 2894 MWt. The 

purpose of the test was to confirm the functions of TCVs, BPV, SRVs and the response of 

system. Table 2 compares the generator load rejection with bypass valves transient’s sequences 

of startup test data with TRACE. Their sequences are very similar. In this transient, when load 

rejection occurred, the TCV closed quickly. Then the BPV opened and reactor scrammed. When 

the water level reached level 3, the recirculation pumps were tripped. Finally, the BPV was reset 

at 6.48 MPa.

Fig.15~20 shows the results of startup test and TRACE. Fig. 15 depicts the power curves of

startup test data and TRACE. The trends of their curves are similar. The TCV fast closure tripped 

the reactor scram. Therefore, the power dropped after 0.3 sec. Fig. 16 compares the steam 

dome pressures of startup test and TRACE. The trends of the curves are approximately in 

agreement. The TCV closing caused the dome pressure to rise. Then, BPV and SRVs opened 

and led to the decline of dome pressure. Fig. 17 shows the steam flow data of startup test and 

TRACE. The TRACE result was consistent with the startup test data. The TCV closing caused 

the steam flow to drop sharply after 0.4 sec. Besides, the feedwater flow and water level 

decreased after TCV closure (shown in Fig. 18 and 19). Due to the dome pressure increase, it 

resulted in the core inlet flow rising during 0.5~2 sec (see Fig. 20). Then, recirculation pumps trip 

caused the decrease of core inlet flow. In summary, the trends of TRACE prediction were 

consistent with startup test data but there were a few differences in the values of the prediction. 

Because we cannot find the detailed startup test data, we don’t know what the reasons cause the 

differences of TRACE results and startup test data. 
3-3

Table 2 The comparison of sequences between startup test and TRACE data 

Event(sec) Startup test TRACE

Transient started 0 0

TCV started to close 0.2 0.2

BPV started to open 0.21 0.21

Reactor scrammed 0.236 0.236

BPV fully opened 0.329 0.329

TCV fully closed 0.394 0.394

Water level reached level 3 2.2 3.1

Steam dome pressure peak 3.9 (7.43 MPa) 2.5 (7.36 MPa)

BPV reset at 6.48 MPa 16.3 18.4

End of analysis - 20
3-4

0 4 8 12 16 20

Time (sec)

0

500

1000

1500

2000

2500

3000

P

o

wer (M

Wt)

TRACE

Startup test

Figure 15 The power data of TRACE and startup test
3-5

0 4 8 12 16 20

Time (sec)

6

6.4

6.8

7.2

7.6

D

o

m

e

pre

s

s

ure (M

P

a)

TRACE

Startup test

Figure 16 The dome pressure data of TRACE and startup test
3-6

0 4 8 12 16 20

Time (sec)

0

500

1000

1500

2000

2500

Ste

a

m flo

w (k

g/s

e

c)

TRACE

Startup test

Figure 17 The steam flow data of TRACE and startup test
3-7

0 4 8 12 16 20

Time (sec)

0

500

1000

1500

2000

2500

F

e

e

d

w

ater flo

w (k

g/s

e

c)

TRACE

Startup test

Figure 18 The feedwater flow data of TRACE and startup test
3-8

0 4 8 12 16 20

Time (sec)

-1

-0.5

0

0.5

1

1.5

2

N

R

W

L (m)

TRACE

Startup test

Figure 19 The narrow range water level (NRWL) data of TRACE and startup test
3-9

0 4 8 12 16 20

Time (sec)

0

2000

4000

6000

8000

10000

12000

C

ore inlet flo

w (k

g/s

e

c)

TRACE

Startup test

Figure 20 The core inlet flow data of TRACE and startup test
3-10

3.2 Startup Test : One Feedwater Pump Trip

This test was performed in November 6, 1981 and the initial power was 2778 MWt. The purpose 

of the test was to confirm the function of FCV (flow control valve) when one feedwater pump 

tripped. Table 3 compares the one feedwater pump trip transient’s sequences of startup test data 

with TRACE. Their sequences are nearly the same. In this transient, after one feedwater pump 

tripped, water level decreased. When the water level reached level 4, the FCV runback was 

started. Then, the power and core flow decreased.

Fig.21~24 shows the results of startup test and TRACE. Fig. 21 depicts the core inlet flow curves 

of startup test data and TRACE. The trends of their curves are similar. One feedwater pump trip 

caused the water level decrease. The FCV runback was tripped when the water level reached 

level 4. Therefore, the core inlet flow decreased due to the FCV runback. Fig. 22 compares the 

powers of startup test and TRACE. The trends of the curves are approximately in agreement. 

After FCV runback, the power decreased. Fig. 23 shows the feedwater flow data of startup test 

and TRACE. The TRACE result was consistent with the startup test data. One feedwater pump 

trip caused the feedwater flow to decrease after 4.9 sec. Besides, the NRWL result of TRACE 

was similar to startup test data (shown in Fig. 24). In summary, the results of TRACE prediction 

were similar to startup test data but there were a few differences in the values of parameters. 
3-11

Table 3 The comparison of sequences between startup test and TRACE

Action (sec) Startup test TRACE

Transient Started 0 0

One feedwater pump tripped 4.9 4.9

Water level dropped to level 4 15.1 15.4

Minimum power value 18.5 (57%) 18.5 (57%)

Minimum core flow 19.4 (79.3%) 18.7 (77%)

End of analysis - 30
3-12

0 10 20 30

Time (sec)

0

40

80

120

C

ore inlet flo

w (%)

TRACE

Startup test

Figure 21 The core inlet flow data of TRACE and startup test
3-13

0 10 20 30

Time (sec)

1000

1500

2000

2500

3000

P

o

wer (M

Wt)

TRACE

Startup test

Figure 22 The power data of TRACE and startup test
3-14

0 10 20 30

Time (sec)

600

800

1000

1200

1400

1600

F

e

e

d

w

ater flo

w (k

g/s

e

c)

TRACE

Startup test

Figure 23 The feedwater flow data of TRACE and startup test
3-15

0 10 20 30

Time (sec)

0

0.2

0.4

0.6

0.8

1

N

R

W

L (m)

TRACE

Startup test

Figure 24 The NRWL data of TRACE and startup test
3-16

3.3 FSAR Case : Generator Load Rejection with Bypass Valves

This transient is similar to the startup test: generator load rejection with bypass valves. The initial 

power was 2894 MWt in this transient. Table 4 compares the generator load rejection with 

bypass valves transient’s sequences of FSAR data with TRACE. Their sequences are very 

similar. In this transient, when load rejection occurred, the TCV closed quickly. Then the 

recirculation pump tripped, reactor scrammed, BPV opened, and loss of feedwater generated. 

Fig. 25~28 shows the results of FSAR and TRACE. Fig. 25 depicts the power curves of FSAR

and TRACE. The trend of TRACE was consistent with FSAR data. The TCV fast closure caused 

the dome pressure increase which made the core void fraction decrease and power increase. 

Then, TCV fast closure tripped the reactor scram. Therefore, the power dropped after 0.7 sec. 

Fig. 26 compares the steam dome pressures of FSAR and TRACE. The trends of their curves 

were similar. The TCV closing caused the dome pressure to rise. Then, BPV and SRVs opened 

and led to the decline of dome pressure. Fig. 27 shows the steam flow data of FSAR and 

TRACE. The TRACE result was consistent with the FSAR data. The TCV closing caused the 

steam flow to drop sharply. The BPV and SRVs opened subsequently which resulted in the 

steam flow increase. Due to the dome pressure increase, it resulted in the core inlet flow rising 

during 0~0.25 sec (see Fig. 28). Then, recirculation pumps trip caused the decrease of core inlet 

flow. In summary, TRACE prediction results were consistent with FSAR data but there were a 

few differences in the values of the prediction. Because we cannot find the detailed FSAR data, 

we don’t know what the reasons cause the differences of TRACE results and FSAR data.
3-17

Table 4 The comparison of sequences between FSAR and TRACE

Action (sec) FSAR TRACE

Transient started 0 0

TCV started to close 0 0

Recirculation pump tripped 0.25 0.25

Reactor scrammed 0.6 0.69

Bypass valve opened 1.7 2.1

Loss of feedwater 3.6 3.5

End of analysis 20 20
3-18

0 4 8 12 16 20

Time (sec)

0

500

1000

1500

2000

2500

3000

3500

4000

P

o

wer (M

Wt)

TRACE

FSAR

Figure 25 The power data of TRACE and FSAR
3-19

0 4 8 12 16 20

Time (sec)

6

7

8

9

D

o

m

e

pre

s

s

ure (M

P

a)

TRACE

FSAR

Figure 26 The dome pressure data of TRACE and FSAR
3-20

0 4 8 12 16 20

Time (sec)

-500

0

500

1000

1500

2000

Ste

a

m flo

w (k

g/s

e

c)

TRACE

FSAR

Figure 27 The steam flow data of TRACE and FSAR
3-21

0 4 8 12 16 20

Time (sec)

0

2000

4000

6000

8000

10000

12000

C

ore inlet flo

w (k

g/s

e

c)

TRACE

FSAR

Figure 28 The core inlet flow data of TRACE and FSAR
3-22

3.4 Core Shroud Leakage Analysis

There are six assumed cases in the Kuosheng 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 0.4 m, the area of break is 0.004 m2

.

Case 2: the core shroud leakage occurred (only in 1 azimuthal sector of the vessel, a small hole), 

the length of break is 0.4 m, the area of break is 0.014 m2

.

Case 3: the core shroud leakage occurred (in 2 azimuthal sectors of the vessel, 360 degree 

larger break), the area of break is 10.248 m2

.

Case 4: the case 3 condition + SBO + scram failed + RCIC failed.

Case 5: the case 3 condition + SBO + LOCA (one steamline break) + scram failed + RCIC failed.

Case 6: the case 3 condition + SBO + LOCA (one steamline break).

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

case 1~6 are shown in Table 5. 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 transients (ex: recirculation pump, feedwater, turbine tripped and MSIV 

closure) were performed for case 4~6 and the LOCA transients (one steamline break) were 

performed for case 5~6. Fig. 29~33 show the results of TRACE for all cases. 

Fig. 29 show the power results of TRACE. In all cases, when core shroud leakage occurred, the 

power increased rapidly. Then the power dropped sharply due to the void fraction increase. After 

core shroud leakage generated fully, the powers of case 1~3 reached the new steady state. 

However, case 4 and 5 had more negative Doppler reactivity due to the fuel temperature raise 

(shown in Fig. 32). So the power of case 4 and case 5 was lower than the power case 1~3 after 

core shroud leakage generated fully. Because the reactor scram occurred, the power of case 6 

dropped sharply.

Fig. 30 shows the water level results of Kuosheng NPP core shroud leakage cases. In case 2~6, 

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

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

level 8 signal tripped, it may be the core shroud leakage happened. Besides, the feedwater 

tripped and RCIC failed resulted in no water injection to the vessel for case 4 and 5. Due to no 

water supply, the water level of the case 4 and case 5 decreased continuously and were lower 

than the TAF (see Fig. 31). Due to one steamline break, the larger steam rushed out for case 5 

and 6 that also caused the water level lower than the TAF. The above results caused the fuel 

temperature increase for the case 4~6 (shown in Fig. 32). The fuel temperature of the case 4 did 

not reach 1088K [6] during 0~1000 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 4 may be larger than 1088 K. Besides, because the 

fuel temperature of the case 5 was larger than 1088 K, the zirconium-water reaction happened. 

The above results indicated that the fuels might be damaged after the zirconium-water reaction 

happened. In case 6, the SBO + LOCA also occurred which was like case 5. However, the 

reactor scram and RCIC were active in case 6. Therefore, the fuel temperature increase was 

delay. The fuel temperature of case 6 went up after 500 sec. It indicated that only RCIC injection 

was not enough for case 6. Fig. 33 shows the dome pressures of TRACE for all cases. There 

were not variation in case 1~3. In case 4, because SRVs opened, there was an oscillation in this 

transient. In case 5~6, because the LOCA occurred, the dome pressure dropped sharply. Finally, 

Fig. 34 depicts the animation model of Kuosheng NPP for case 5.
3-23

Table 5 The sequences of case 1~6 for core shroud leakage transient

Action (sec) Case 1 Case 2 Case 3

Transient started 200 200 200

Core shroud leakage 

started

200

(leakage area: 

0.004m2)

200

(leakage area: 

0.14m2)

200

(leakage area: 

10.248m2)

NRWL reached level 8 - 208 204

End 1000 1000 1000

Action (sec) Case 4

(SBO)

Case 5

(SBO+LOCA)

Case 6

(SBO+LOCA)

Transient started 200 200 200

Core shroud leakage started 200

(leakage area: 

10.248m2)

200

(leakage area: 

10.248m2)

200

(leakage area: 

10.248m2)

Steam line LOCA generated - 200 200

TCV started to close 200.1 200.1 200.1

Feedwater tripped 200.1 200.1 200.1

Reactor scrammed - - 200.1

Recirculation pmup tripped 200.1 200.1 200.1

MSIV started to close 200.1 200.1 200.1

Water level reached level 8 204 204 204

RCIC started - - 234

Water level reached TAF 261 260 291

Water level reached BAF - 431 -

Cladding temperature 

reached 1088K - 550 -

End 1000 1000 1000
3-24

Figure 29 The power results

0.00E+00

1.00E+09

2.00E+09

3.00E+09

4.00E+09

5.00E+09

6.00E+09

7.00E+09

0 200 400 600 800 1000

Power (W)

時間:秒sec

Case1

Case2

Case3

Case4

Case5

Case6

Time (sec) 
3-25

Figure 30 The NRWL results

Time (sec) 
3-26

Figure 31 The vessel water level results

Time (sec) 
3-27

Figure 32 The fuel rod temperature results

Time (sec) 
3-28

Figure 33 The dome pressure results

0

1

2

3

4

5

6

7

8

9

0 200 400 600 800 1000

Dome pressure (MPa)

時間:秒sec

Case1

Case2

Case3

Case4

Case5

Case6

Time (sec) 
3-29

(a) 0 sec

(b) 202 sec
3-30

(c) 583 sec

(d) 981 sec

Figure 34 The animation model of Kuosheng NPP for core shroud leakage transient

(a) 0 sec, (b) 202 sec, (c) 583 sec, (d) 981 sec
4-1

4. CONCLUSIONS

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

rejection and a feedwater pump trip transients were selected to validate the Kuosheng NPP 

TRACE model. The results and sequences of TRACE are similar to the FSAR and startup tests 

data. By the above compared results, it indicates that there is a respectable accuracy in the 

Kuosheng NPP TRACE/SNAP model and it also shows that the Kuosheng NPP TRACE/SNAP

model is satisfying for the purpose of Kuosheng NPP safety analyses with confidence. 

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

There are six 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 Kuosheng

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 1~3) wasn’t 

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

/ one steamline break (LOCA) transient (case 4~6) may cause the cladding temperature larger

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

 
5-1

5. REFERENCES

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

2. K. Y. Lin, 2012. Verification of the Kuosheng BWR/6 TRACE Model with Load Rejection 

Startup Test, ASME 2012 V&V.

3. Taiwan Power Company, 2001. Final Safety Analysis Report for Kousheng Nuclear Power 

Station Units 1&2 (FSAR).

4. R.Y. Yuann, H.T. Lin, 2009. Guideline of Generating Parameters for Reload Licensing 

Analyses for Kuosheng Units 1 and 2, INER report, INER-6529R.

5. J. R. Wang et al, 1989. Kuosheng Startup Tests Transient Analyses, INER report, 

INER-0965.

6. Taiwan Power Company, 2014. Ultimate Response Guideline for Kousheng Nuclear Power 

Station.






UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA

-0450 The Development and Application of Kuosheng (BWR/6)

Nuclear Power Plant TRACE/SNAP Model