NUREG/IA-0427 

International 

 Agreement Report

Application of TRACE V5.0 P2 to China Domestic 

PWR LBLOCA Analysis 

Prepared by:

FENG Jinjun, CHAI Guohan, ZHOU Kefeng, SHI Junying

Nuclear and Radiation Safety Centre

Ministry of Environmental Protection

54 Honglian Nancun

Haiden District, Beijing, 100082, China 

A. Calvo, NRC Project Manager

 

 Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: March 2013 

Date Published: July 2013

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0427 

International 

 Agreement Report

Application of TRACE V5.0 P2 to China Domestic 

PWR LBLOCA Analysis 

Prepared by:

FENG Jinjun, CHAI Guohan, ZHOU Kefeng, SHI Junying

Nuclear and Radiation Safety Centre

Ministry of Environmental Protection

54 Honglian Nancun

Haiden District, Beijing, 100082, China 

A. Calvo, NRC Project Manager

 

 Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: March 2013 

Date Published: July 2013

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

The purpose of this work is to study the behavior of China domestic PWR under LBLOCA

scenario using the TRACE (TRAC/RELAP Advanced Computational Engine) code. The work is 

divided into five parts:

The first part is TRACE model establishment. SNAP(Symbolic Nuclear Analysis Program)

program was used to facilitate system modeling work. Important components such as active 

core, pressurizer, accumulator and steam generator were modeled respectively. These 

components were tested separately and results were compared with design data to check the 

accuracy. Key parameters were indentified and properly adjusted to refine the model further. All 

of the components were incorporated together to build up the integrated TRACE model of China 

domestic PWR. 

The second part is steady state calculation. Steady state of full power operation was simulated 

by TRACE code and calculation results were compared with design data. Hydraulic frictions 

were adjusted to keep calculated and designed flow distribution as close as possible. The 

adjustment work was iterated until all of key parameters were acceptable.

The third part is transient calculation. The LBLOCA scenario was simulated in this part. Restart 

case of accident scenario was prepared based on the steady state TRACE model established 

previously. The transient calculation results showed that safety goal was achieved under the 

assumed accident scenario. 

The forth part is sensitivity analysis. Sensitivity analysis of break spectrum and initial 

accumulator pressure was performed respectively. The most limiting break size and proper initial 

accumulator pressure were found though the sensitivity analysis.

The last part is accident scenario animation. SNAP was used to create the animation of the 

LBLOCA Accident scenario. Better understanding of the calculated physical phenomena and 

transient process was obtained via animation demonstration.
iv
v

FOREWORD

TRACE is an advanced thermal hydraulic code developed by USNRC. It is a kind of 

best-estimate safety analysis code widely used in PWR, BWR and test grid thermal hydraulic 

analysis, such as LOCA, operation transient and other accident scenario. Traditional safety 

analysis codes, TRAC, RELAP and REMONA for example, are incorporated into TRACE code 

to build up a modern integrated safety analysis toolset. In the future, TRACE is going to play a 

very important role in design basis accident analysis and take the place of NRC traditional safety 

analysis codes mentioned above.

China and U.S. have signed an agreement on CAMP (Code Applications and Maintenance 

Program). Both sides are responsible for the development and maintenance of CAMP codes 

such as TRACE. The Nuclear and radiation Safety Center of China (NSC) is a governmental 

organization in China to provide technical support to the National Nuclear Safety Administration 

(NNSA). It is responsible for application of TRACE code in thermal hydraulic safety analysis. 

Users’ experience and bugs found in code running should be carefully recorded and reported to 

USNRC according to CAMP framework. To meet this requirement, we built a TRACE model of 

China domestic PWR, performed steady state calculation, transient calculation and sensitivity 

analysis. Transient scenario was animated with SNAP code. Finally this report was prepared to 

share the code application experience with other CAMP members.
vi
vii

CONTENTS

Page

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

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

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

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

EXECUTIVE SUMMARY ................................................................................................................xi

ABBREVIATIONS......................................................................................................................... xiii

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

2. METHODOLOGY................................................................................................................... 2-1

3. ESTABLISHMENT AND VERIFICATION OF CHINA DOMESTIC PWR TRACE MODEL.. 3-1

3.1 General Description ......................................................................................................... 3-1

3.2 Reactor Vessel Component............................................................................................. 3-1

3.3 Power Component ........................................................................................................... 3-2

3.4 Primary Loop Components .............................................................................................. 3-2

3.5 Secondary Loop Components ......................................................................................... 3-3

4. STEADY STATE CALCULATION ......................................................................................... 4-1

4.1 Model Description............................................................................................................. 4-1

4.2 Steady State Calculation Results..................................................................................... 4-1

5. TRANSIENT CALCULATION ................................................................................................ 5-1

5.1 Transient Calculation Results .......................................................................................... 5-1

5.2 Sensitivity Analysis........................................................................................................... 5-5

6. TRANSIENT ANIMATION ..................................................................................................... 6-1

7. CONCLUSIONS..................................................................................................................... 7-1

8. REFERENCES....................................................................................................................... 8-1
viii
ix

FIGURES

Page

Fig. 1 The China Domestic PWR Site..................................................................................... 1-2

Fig. 2 The Flow Chart of TRACE Model Establishment and Verification ............................... 2-2

Fig. 3 The TRACE Model of the Reactor Vessel Component for China Domestic PWR....... 3-1

Fig. 4 The TRACE Model of the Power Component for China Domestic PWR ..................... 3-2

Fig. 5 The TRACE Model of the Primary Loop Components for China Domestic PWR........ 3-3

Fig. 6 The TRACE Model of the Secondary Loop Components for China Domestic PWR... 3-4

Fig. 7 The TRACE Model of China Domestic PWR................................................................ 4-1

Fig. 8 Reactor Power............................................................................................................... 4-2

Fig. 9 Average Channel Rod Temperature ............................................................................. 4-2

Fig. 10 Average Channel Clad Temperature ............................................................................ 4-3

Fig. 11 Hot Channel Rod Temperature..................................................................................... 4-3

Fig. 12 Hot Channel Clad Temperature.................................................................................... 4-4

Fig. 13 Hot Leg and Cold Leg Temperature ............................................................................. 4-4

Fig. 14 Hot Leg and Cold Leg Pressure.................................................................................... 4-5

Fig. 15 Reactor Power (Transient)............................................................................................ 5-1

Fig. 16 Pressurizer Pressure (Transient).................................................................................. 5-2

Fig. 17 Break Flow (Transient).................................................................................................. 5-2

Fig. 18 Core Flow (Transient) ................................................................................................... 5-3

Fig. 19 Core Collapsed Water Level......................................................................................... 5-3

Fig. 20 Safety Injection Flow ..................................................................................................... 5-4

Fig. 21 Quench Front ................................................................................................................ 5-4

Fig. 22 Hot Rod Clad Temperature........................................................................................... 5-5

Fig. 23 Break Spectrum Analysis .............................................................................................. 5-6

Fig. 24 Initial Accumulator Pressure Sensitivity Analysis.......................................................... 5-6

Fig. 25 Transient Scenario Animation....................................................................................... 6-1
x
xi

EXECUTIVE SUMMARY

Couple of years ago, NNSA(China National Nuclear Safety Administration) and USNRC signed 

an agreement on CAMP(Code Applications and Maintenance Program). NSC (Nuclear and 

radiation Safety Center of China) is a government organization to supply technical support to the 

NNSA(National Nuclear Safety Administration) in China. It has the responsibility to apply the 

TRACE code in thermal hydraulic safety analysis. Users’ experience and bugs found in code 

running should be carefully recorded and reported to USNRC according to CAMP framework. To 

meet this requirement, we built a TRACE model of China domestic PWR, performed steady 

state calculation, transient calculation and sensitivity analysis. We animated the accident 

scenario and finally prepared this report to share the code application experience.

The simulated China domestic PWR is a 2-loop pressurized water reactor designed by a 

domestic institution of China. The plant site is located near Shanghai, east coast of China shown 

in Fig. 1. The reactor rated thermal power is 1930 MW. AFA 3G fuel assembly is used in the 

core. And fuel cycle length is 18 months. Analysis results show that peak clad temperature is 

below the safety limit and other key parameter trends are reasonable during LBLOCA simulation. 
xii
xiii

ABBREVIATIONS

CAMP Code Applications and Maintenance Program

NSC Nuclear and radiation Safety Center of China

NNSA China National Nuclear Safety Administration

PWR Pressurized Water Reactor

LPI Low Pressure Injection

HPI High Pressure Injection

LBLOCA Large Break Loss Of Coolant Accidents

NRC Nuclear Regulatory Commission

SG Steam Generator

SNAP Symbolic Nuclear Analysis Program

TRACE TRAC/RELAP Advanced Computational Engine

US United States
xiv
1-1

1. INTRODUCTION

TRACE is an advanced thermal hydraulic code developed by USNRC. It is a kind of 

best-estimate safety analysis code widely used in PWR, BWR and test grid thermal hydraulic 

analysis, such as LOCA, operation transient and other accident scenario. Traditional safety 

analysis codes, TRAC, RELAP, and REMONA for example, are incorporated into TRACE code 

to build up a modern integrated safety analysis toolset. In the future, TRACE will play a very 

important role in design basis accident analysis and take the place of NRC traditional safety 

analysis codes mentioned above.

China and U.S. have signed an agreement on CAMP (Code Applications and Maintenance 

Program). Both sides are responsible for the development and maintenance of CAMP codes 

such as TRACE. The Nuclear and radiation Safety Center of China (NSC) is a governmental 

organization in China to provide technical support to the National Nuclear Safety Administration 

(NNSA). It is responsible for application of the TRACE code in thermal hydraulic safety analysis. 

Users’ experience and bugs found in code running should be carefully recorded and reported to 

USNRC according to CAMP framework. To meet this requirement, we built a TRACE model of 

China domestic PWR, performed steady state calculation, transient calculation and sensitivity 

analysis. We animated the accident scenario and finally prepared this report to share the code 

application experience.
1-2

Fig. 1 The China Domestic PWR Site
2-1

2. METHODOLOGY

SNAP v 2.1.0 and TRACE v 5.0p2 were used in this work. The methodology of the research is 

as following: 

The first step is TRACE model establishment. SNAP(Symbolic Nuclear Analysis Program)

program was used to facilitate system modeling work. Important components such as vessel, 

cold leg, pressurizer, accumulator and steam generator were modeled respectively. These 

components were tested separately and compared with design data to check the accuracy. Key 

parameters were indentified and properly adjusted to refine the model further. All of the 

components were incorporated together and finally formed the integrated TRACE model.

The second step is steady state calculation. Steady state of full power operation was simulated 

by TRACE code and calculation results were compared with design data. Hydraulic frictions 

were adjusted to keep calculated and designed flow distribution as close as possible. Iterated 

adjustment was performed until all of key parameters were acceptable.

The third step is transient calculation. The accident scenario “Large Break LOCA” was simulated 

in this part. Restart model of accident scenario was prepared based on the steady state TRACE 

model established in the previous parts. The transient calculation results showed that safety goal 

was achieved under the assumed accident scenario. 

The forth step is sensitivity analysis. Sensitivity analysis of initial accumulator pressure and cold 

leg break size was performed respectively.

The last step is accident scenario animation. SNAP was used to create the animation of the 

“Large Break LOCA Accident scenario”. Better understanding of the calculated physical 

phenomena and transient process was obtained via animation demonstration.

The complete process is presented in Fig. 2.
2-2

Fig. 2 The Flow Chart of TRACE Model Establishment and Verification

Processing design data

Setting up and refining the TRACE model 

of important components with SNAP

Verifying the TRACE model of important 

components

Acceptable?

Integrating and refining the TRACE model 

of China domestic PWR

Performing steady state calculation and 

comparing the results with the design data

Verifying the TRACE model of China 

domestic PWR

Acceptable?

Performing transient calculation and 

animating the transient scenario

NO

NO

YES

YES

Preparing the report
3-1

3. ESTABLISHMENT AND VERIFICATION OF CHINA DOMESTIC PWR

TRACE MODEL

3.1 General Description

The simulated China domestic PWR is a 2-loop pressurized water reactor designed by a 

domestic institution of China. The plant site is located near Shanghai, east coast of China shown 

in Fig. 1. The reactor rated thermal power is 1930 MW. AFA 3G fuel assembly is used in the 

core. And fuel cycle length is 18 months. Important components such as core, accumulator, hot 

leg and SG were modeled respectively. These components were tested separately and 

compared with design data to check the accuracy. Key parameters were indentified and properly 

adjusted to refine the model further. After refining, all of the components were incorporated

together to build up the integrated TRACE model of primary and secondary system. 

3.2 Reactor Vessel Component

The height of active core is 3.658m. It consists of 121 fuel assemblies and each assembly is 

made up of 264 fuel rods. The arrangement of fuel rods in an assembly is 17 X 17, in which 24 

rod positions are occupied by control rod guide tubes and 1 rod position is measurement tube. 

The size of fuel assembly is 21.4cm X 21.4cm. The pitch of fuel rods is 12.6mm. Outer diameter 

of fuel rod is 9.5mm. Clad thickness is 0.57mm. The diameter of UO2 pellet is 8.192mm. The 

designed inlet coolant temperature is 292.8°C and outlet coolant temperature is 327.2°C. 

Pipe component was used to build up reactor vessel, include active core, bypass, downcomer, 

upper plenum and lower plenum. The heat structure associated with core was used to simulate 

the average fuel rod and heat transferring from fuel to coolant. The core component and 

associated heat structure is shown in Fig. 3. The fill and break components were used to model 

core inlet and outlet boundary conditions. The reactor vessel component was tested separately

and the difference between calculated results and designed data were checked to confirm the 

accuracy of modeling. 

Fig. 3 The TRACE Model of the Reactor Vessel Component for China Domestic PWR
3-2

3.3 Power Component

Power component was used to generate power in heat structure standing for fuel rods. The 

rated thermal power is 1930 MW. Chopped COSINE distribution was assumed for core axial 

power and flat power distribution was assumed for fuel rod radial power. ANS-94 was used to 

generate decay heat. In the steady state, initial power of 1930MW was maintained to perform 

steady state calculation. Afterwards, “Large Break LOCA” accident was initiated and reactor 

scrammed due to negative void reactivity. After reactor scram, decay heat dominated the heat 

generating of core. The power component is shown in Fig. 4. 

Fig. 4 The TRACE Model of the Power Component for China Domestic PWR

3.4 Primary Loop Components 

Primary loop components mainly consist of cold leg pipe, hot leg pipe, pressurizer, 

accumulators, LPI, HPI, SG tubes and coolant pumps shown in Fig. 5. 

The SG tubes were modeled with pipe component with 8 cells and heat structure was 

associated to simulate heat transferring from primary loop to secondary loop. The pressurizer 

was modeled with prizer component to stabilize system pressure and the pressure setpoint is 

15.5 MPa. The accumulators were modeled with pipe component to supply emergency water 

injection during the refill stage of LBLOCA and actuating setpoint is 4.235 Mpa. The HPI and LPI 

were modeled with fill component to supply safety water injection during reflood stage of 

LBLOCA and the mass flow depends on injection point pressure. 

Break was modeled by break and valve component connected to broken loop cold leg. 

Discharge coefficient of 0.7 was selected according to final safety analysis report.
3-3

Fig. 5 The TRACE Model of the Primary Loop Components for China Domestic PWR

3.5 Secondary Loop Components

Secondary Loop Components mainly consist of SG secondary side, main feed water, auxiliary 

feed water, steam line and related valves.

SG secondary side mainly includes downcomer, boiler, separator and steam dome. Main feed 

water and auxiliary feed water were modeled with fill component connected to a one-cell pipe, 

which was on top of downcomer. Downcomer was modeled with pipe component with 4 cells. 

Feed water flows downwards through the downcomer and goes into boiler at the bottom cell. 

Heat is transferred from SG primary side to the secondary side through SG tube wall and steam 

is generated in the boiler. The mixture of steam and water rises to the separator to be separated. 

Separator was modeled with ideal separator component with 1 cell. Dry steam continues to 

move up to the steam dome and separated water goes down to downcomer again. Proper form 

loss coefficient must be included to keep specific circulation rate. On top of steam dome, there is 

a built-in steam flow limiter. Flow area was decreased to around 0.1 m2 to limit steam flow during 

main steam line break accident. Steam line was modeled with pipe component and isolation 

valves, bypass valves and safety relief valves were connected to related steam lines.
3-4

Fig. 6 The TRACE Model of the Secondary Loop Components for China Domestic PWR
4-1

4. STEADY STATE CALCULATION

4.1 Model Description

There are 2 loops for the simulated China domestic PWR as shown in Fig. 7. Pressurizer was 

connected to the intact loop to delay the safety injection signal which was activated by low low 

pressurizer pressure. While in steady state calculation, there will not be any safety injection 

signal, which will appear in next transient calculation step. In designed steady state, reactor 

power is 1930MW, pressurizer pressure is 15.5 MPa, primary side mass flow is 4796 kg/s per 

loop, bypass flow is 6.5%, main steam pressure is 6.71 MPa, main steam quality is 99.75%, 

steam mass flow is 541.9 kg/s per SG, circulation rate is 3.4%.

Fig. 7 The TRACE Model of China Domestic PWR

4.2 Steady State Calculation Results

Before transient calculation, steady state calculation must be performed first to make sure each 

parameter of the system is close enough to the design value. Frictions must be finely adjusted to 

get the right coolant mass flow distribution. After model refining, key parameters of the system 

have acceptable difference from the design data. Steady state calculation results are shown in 

Fig.8 to 14. 
4-2

Fig. 8 Reactor Power

Fig. 9 Average Channel Rod Temperature
4-3

Fig. 10 Average Channel Clad Temperature

Fig. 11 Hot Channel Rod Temperature
4-4

Fig. 12 Hot Channel Clad Temperature

Fig. 13 Hot Leg and Cold Leg Temperature
4-5

Fig. 14 Hot Leg and Cold Leg Pressure

5-1

5. TRANSIENT CALCULATION 

5.1 Transient Calculation Results

The “Large Break LOCA” accident scenario was simulated in this section. LBLOCA assumption 

is as following: initial reactor power is 100%FP, axial power distribution is chopped cosine, Fq is 

2.35, cold leg break size is 0.7 X 2A, coolant pumps coast down after break initiation, trip signal 

is initiated by low pressurizer pressure, safety injection signal is initiated by low low pressurizer 

pressure, main feed water isolation is delayed by 7s, safety injection signal is delayed by 30s, 

auxiliary feed water startup is delayed by 62s. Realistic model plus conservative assumption 

method is used in this simulation. Normally initial reactor power of 102%FP is assumed in final 

safety analysis report, but in this report we assumed initial reactor power was 100%FP, so this 

LBLOCA simulation result is more optimistic. Transient calculation results are shown in Fig.15 to 

22. 

Fig. 15 Reactor Power (Transient)
5-2

Fig. 16 Pressurizer Pressure (Transient)

 

Fig. 17 Break Flow (Transient)
5-3

Fig. 18 Core Flow (Transient)

Fig. 19 Core Collapsed Water Level
5-4

Fig. 20 Safety Injection Flow

Fig. 21 Quench Front 
5-5

Fig. 22 Hot Rod Clad Temperature

5.2 Sensitivity Analysis

Lots of studies show guillotine break of cold leg is not the most limiting case in large break 

LOCA accident. Break size must be searched to determine which one is the most limiting case. 

This is so called break spectrum analysis. In this report, break size of 0.7 X 2A, 0.6 X 2A, 0.5 X 

2A were analyzed, sensitivity analysis result shows that break size of 0.6 X 2A is the most 

limiting case in our simulation. Break spectrum analysis results are shown in Fig.23. 

Different initial accumulator pressure will also affect peak clad temperature during large break 

LOCA accident. If initial accumulator pressure is too high, more accumulator injection mass flow 

will bypass from the downcomer to the cold leg break, fewer water will be left to inject into the 

active core, the core cooling condition will be worse. If initial accumulator pressure is too low, 

high pressure steam out of the core can easily prevent accumulator injection mass flow 

penetrating downcomer to enter active core. So it is necessary to perform initial accumulator 

pressure sensitivity analysis to search proper pressure actuating setpoint for accumulators. In 

this report, initial accumulator pressures of 2.0 MPa, 2.5 MPa, 3.0MPa, 4.0MPa were analyzed 

separately. Simulation result shows initial accumulator pressures of 2.5 MPa will significantly 

decrease peak clad temperature during large break LOCA scenario compared to other pressure 

setpoints. Analysis results are shown in Fig.24.
5-6

Fig. 23 Break Spectrum Analysis

Fig. 24 Initial Accumulator Pressure Sensitivity Analysis
6-1

6. TRANSIENT ANIMATION

The “Large Break LOCA” accident scenario was animated in this section. SNAP 2.1.0 was used 

to animate the accident scenario. Fluid conditions and key parameters such as temperature and 

mass flow direction can be observed clearly for each part of system during the accident 

progress. Better understanding of the calculated physical phenomena and transient process was 

obtained via animation demonstration.

Fig. 25 Transient Scenario Animation

7-1

7. CONCLUSIONS

In this report, China domestic PWR was successfully modeled by SNAP and TRACE code. The 

model was mainly made up of pipe, fill, break, prizer, separator, valve and power component. 

Key parameters of each component were finely modeled to match the design data. Steady state 

calculation was performed with the integrated TRACE model. Errors between steady state 

calculation results and design data were negligible. Restart file was generated by steady state 

calculation. Based on the steady state restart file, transient case was introduced and transient 

calculation results were generated consequently. After transient calculation, animation work was 

done with SNAP code. Fluid conditions and key parameters could be observed simultaneously

as the transient scenario went on. The animation function gave more interactive information and 

better understanding of accident progress and physical phenomena to analyst. This work 

indicates that SNAP/TRACE code works well to simulate China domestic PWR and the 

calculation results are reasonable.

In China, NSC has the responsibility to apply TRACE code in thermal hydraulic analysis. The 

TRACE application work carried out by NSC is still on the beginning. The work in this report is 

very preliminary, but we are always ready to report our work to NRC and share our experience of 

TRACE code application with other TRACE users according to CAMP agreement. NSC has 

made a plan to model AP1000 with TRACE code next step. The future work will be also reported 

to NRC after completion and we hope it more contributable.

8-1

8. REFERENCES

1. NRC, “TRACE V5.0 User’s Manual”, volume 2, Office of Nuclear Regulatory Research, 

2007

2. Jong-Rong Wang, Hao-Tzu Lin, Chunkuan Shih, “Assessment of the TRACE Code 

Using Transient Data from Maanshan PWR Nuclear Power Plant”, International

Agreement Report, NUREG/IA-0241, Institute of Nuclear Energy Research, Atomic 

Energy Council, R.O.C., 2010. 

3. A. Prošek, B. Mavko, “IJS Animation Model for Krško NPP”, International Agreement

Report, NUREG/IA-0227, Jožef Stefan Institute, 2010. 

4. E. Takasuo, “An Assessment of TRACE V4.160 Code Against PACTEL ATWS-10 – 13 

and ATWS-20 – 21 Pressurizer Experiments”, International Agreement Report, 

UREG/IA-0231, VTT Processes, 2010. 

5. J.I.Sanchez, C.A.Lage, T.Nunez, “LBLOCA Analysis in a Westinghouse PWR 3-Loop 

Design using RELAP5/MOD3”, International Agreement Report, UREG/IA-0195, 

Empresa Nacional del Uranio S.A., 2001. 




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NUREG/IA-0427 Application of TRACE V5.0 P2 to China Domestic PWR LBLOCA Analysis