NUREG/IA-0418 

International 

Agreement Report

Application of TRACE V5.0 P2 to Natural

Circulation Reactor Safety Analysis

 

Prepared by:

FENG Jinjun, CHAI Guohan, SHI Junying, ZHOU Kefeng

 

 Nuclear and Radiation Safety Centre, Ministry of Environmental Protection

 54, Honglian Nancun, Haidian 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: February 2012

Date Published: October 2012

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 under an 

international cooperative agreement for the exchange of 

technical information. 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-0418 

International 

Agreement Report

Application of TRACE V5.0 P2 to Natural

Circulation Reactor Safety Analysis

 

Prepared by:

FENG Jinjun, CHAI Guohan, SHI Junying, ZHOU Kefeng

 

 Nuclear and Radiation Safety Centre, Ministry of Environmental Protection

 54, Honglian Nancun, Haidian 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: February 2012

Date Published: October 2012

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 natural circulation reactor under Accident 

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 core and 

heat exchanger 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 natural circulation reactor.

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 primary side natural circulation mass flow as 

close as possible. The adjustment work was iterated until all of key parameters were 

acceptable.

The third part is transient calculation. The Accident scenario “Loss of Main Feed Water” 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 main feed water flow rate and reactor 

power to primary side natural circulation flow was performed respectively. Special phenomenon 

of natural circulation flow acceleration due to “loss of feed water” was understood better though 

the sensitivity analysis.

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

“Loss of Main Feed Water” Accident scenario. Better understanding of the calculated physical 

phenomena and transient process was obtained via animation demonstration.

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 

Natural Circulation Reactor, 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.

vii

CONTENTS

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

Page

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

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

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

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

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

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

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

3. ESTABLISHMENT AND VERIFICATION OF NATURAL CIRCULATION 

REACTOR TRACE MODEL.................................................................................................3-1

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

3.2 Core Component ............................................................................................................3-1

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

3.4 Heat Exchanger Component ..........................................................................................3-2

3.5 Other 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.........................................................................................................5-3

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

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

8. REFERENCES.....................................................................................................................8-1

ix

FIGURES

Page

1. Fig. 1 The Schematic Diagram of Natural Circulation Reactor .......................................1-2

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

3. Fig. 3 The TRACE Model of the Core Component for NCR............................................3-1

4. Fig. 4 The TRACE Model of the Power Component for NCR .........................................3-2

5. Fig. 5 The TRACE Model of the Heat Exchanger Component for NCR..........................3-3

6. Fig. 6 The TRACE Model of the Atmosphere and Downcomer for NCR.........................3-3

7. Fig. 7 The TRACE Model of NCR System ......................................................................4-1

8. Fig. 8 Difference Between Steady State Calculation Results and Design Data..............4-1

9. Fig. 9 Mass Flow of Primary and Secondary Side (SS) ..................................................4-2

10. Fig. 10 Pressure of Core ...................................................................................................4-2

11. Fig. 11 Core Inlet and Outlet Temperature (SS)................................................................4-3

12. Fig. 12 Rod and Clad Temperature (SS)...........................................................................4-3

13. Fig. 13 HX Primary Side Inlet and Outlet Temperature (SS).............................................4-4

14. Fig. 14 HX Secondary Side Inlet and Outlet temperature (SS) .........................................4-4

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

16. Fig. 16 Primary Coolant Flow (Transient)..........................................................................5-2

17. Fig. 17 Core Inlet and Outlet Temperature (Transient) .....................................................5-2

18. Fig. 18 Rod and Clad temperature (Transient)..................................................................5-3

19. Fig. 19 Feed Water Sensitivity to Natural Circulation Flow ...............................................5-4

20. Fig. 20 Reactor Power Sensitivity to Natural Circulation Flow ..........................................5-4

21. Fig. 21 Transient Scenario Animation ...............................................................................6-1

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 Natural Circulation Reactor, 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.

Natural Circulation Reactor is a kind of concept design developed by domestic institution in 

China. Most of the important components are submerged in a twenty-meter deep water pool 

shown in Fig. 1. The whole system is depressurized. The reactor core is placed at the bottom of 

the pool and the rated thermal power is 120 MW. Fuel rods are cooled by natural circulation

flow. The designed core outlet temperature is 110°C. Hot water out of the core rises upwards to 

a heat exchanger and transfer heat to secondary side feed water. Primary coolant is cooled 

down and flows out of heat exchanger to the downcomer, and finally heads to the inlet of the 

core to complete the whole natural circulation.

The specific natural circulation reactor analyzed in this report has no prototype yet, so no 

experiment data could be collected to do comparison between calculated results and measured 

data. In this report, only theoretical analysis was performed and the rationality of calculation 

results was judged by expert experience. This work indicates that the TRACE code is certainly 

capable to perform thermal hydraulic analysis on Natural Circulation Reactor and generates 

reasonable results.

xiii

ABBREVIATIONS

CAMP Code Applications and Maintenance Program

NSC Nuclear and radiation Safety Center of China

NNSA China National Nuclear Safety Administration

NCR Natural Circulation Reactor

LOCA Loss Of Coolant Accidents

NRC Nuclear Regulatory Commission

HX Heat Exchanger

SNAP Symbolic Nuclear Analysis Program

TRACE TRAC/RELAP Advanced Computational Engine

US United States

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 

Natural Circulation Reactor, 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.

As the mentioned natural circulation reactor is still on design paper on the time being, no real 

data could be collected to do comparison between calculated and measured data. In this report, 

only theoretical analysis was performed and the rationality of calculation results was judged by 

expert experience. Unique phenomenon of natural circulation coolant flow acceleration due to 

“loss of feed water” was observed in transient analysis. Analysis results indicate that the design 

is able to keep the reactor safe under assumed accident scenario.
1-2

Fig. 1 The Schematic Diagram of Natural Circulation Reactor
2-1

2. METHODOLOGY

SNAP v 2.0.8 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 core and 

heat exchanger 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 primary side natural circulation mass flow as 

close as possible. Iterated adjustment was performed until all of key parameters were 

acceptable.

The third step is transient calculation. The accident scenario “Loss of Main Feed Water” 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 main feed water and reactor power 

to primary side natural circulation flow was performed respectively.

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

“Loss of Main Feed Water 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 Natural Circulation Reactor

Performing steady state calculation and 

comparing the results with the design data

Verifying the TRACE model of Natural 

Circulation Reactor

Acceptable?

Performing transient calculation and 

animating the transient scenario

NO

NO

YES

YES

Preparing the report
3-1

3. ESTABLISHMENT AND VERIFICATION OF NATURAL 

CIRCULATION REACTOR TRACE MODEL

3.1 General Description

Natural Circulation Reactor is a kind of concept design developed by domestic institution in 

China. Most of the important components are submerged in a twenty-meter deep water pool 

shown in Fig. 1. The primary coolant loop is depressurized. The reactor core is placed at the 

bottom of the pool and the rated thermal power is 120 MW. Important components such as core 

and heat exchanger 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 NCR system.

3.2 Core Component

The core component is placed at the bottom of the primary loop. The rated thermal power is 120 

MW. The height of core is 1.1m. It consists of 205 fuel assemblies and each assembly is made 

up of 60 fuel rods. The arrangement of fuel rods in an assembly is 8 X 8, and the central 4 rod 

positions are occupied by control rods and water rods. The size of fuel assembly is 10.8cm X 

10.8cm. The pitch of fuel rods is 13.4mm. Outer diameter of fuel rod is 10mm. Zr4 clad 

thickness is 0.7mm. The diameter of UO2 pellet is 8.43mm. The designed core inlet coolant 

temperature is 80°C and outlet coolant temperature is 110°C.

Pipe component was used to build up the core and the associated heat structure 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 core 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 Core Component for NCR
3-2

3.3 Power Component

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

distribution was assumed for core axial power. ANS-94 was used to generate decay heat. In 

transient calculation, Point kinetic method and reactivity feedback of temperature is turned on in 

the power component. In the steady state, initial power of 120MW was maintained to perform 

steady state calculation. Accident “Loss of Main Feed Water” was initiated afterwards. But the 

reactor scram was delayed by 120s. During the delayed 120s, reactivity feedback of fuel 

temperature and coolant temperature was functional. 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 NCR

3.4 Heat Exchanger Component

The heat exchanger was modeled to simulate heat transferring from primary loop 

to secondary loop. The height and volume of heat exchanger was kept as 

accurate as possible since natural circulation capability and heat capacity could 

be affected by these two parameters.

Two pipe components were used to model primary side and secondary side of the 

heat exchanger. A heat structure was built up to simulate tubes to transfer heat. 

For the exchanger primary side, fill and break components were used to model 

inlet and outlet boundary condition. For the exchanger secondary side, fill 

component was used to model main feed water and auxiliary feed water. Break 

component was used to maintain the secondary side outlet boundary condition. 

The heat exchanger model is shown in Fig. 5.
3-3

Heat exchange surface was carefully adjusted to keep calculated outlet temperatures of both 

primary side and secondary side as close as possible to the design value.

Fig. 5 The TRACE Model of the Heat Exchanger Component for NCR

3.5 Other Components

Other components including downcomer and atmosphere were also well modeled with pipe and 

break components. The atmosphere was modeled by a break component to maintain system 

boundary pressure. The volume of downcomer is a very important parameter in transient 

simulation because it affects the heat capacity of primary loop significantly. After all of the 

components were modeled respectively and well tested, they could be incorporated together to 

set up the whole picture of the NCR TRACE model.

Fig. 6 The TRACE Model of the Atmosphere and Downcomer for NCR

4-1

4. STEADY STATE CALCULATION

4.1 Model Description

There is only one primary loop and one secondary loop for the whole NCR system as shown in 

Fig. 7. Fuel rods generate heat in the core which is cooled by natural circulation coolant flow. 

Hot water leaves the outlet of the core, rises through pipes and enters the inlet of heat 

exchanger. Heat transfers from the primary side to secondary side and primary coolant 

temperature drops before getting to outlet of the heat exchanger. Cooled water flows through a 

horizontal pipe to the downcomer and finally heads to the bottom of the core. The NCR TRACE 

model is shown in Fig. 7.

Fig. 7 The TRACE Model of NCR System

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 natural circulation coolant mass flow. After model refining, key parameters of the 

system have acceptable difference from the design data. The difference between steady state 

calculation results and the design data are shown in Fig. 8. More steady state calculation results 

are shown in Fig. 9 to 14.

Fig. 8 Difference Between Steady State Calculation Results and Design Data
4-2

Fig. 9 Mass Flow of Primary and Secondary Side (SS)

Fig. 10 Pressure of Core
4-3

Fig. 11 Core Inlet and Outlet Temperature (SS)

Fig. 12 Rod and Clad Temperature (SS)
4-4

Fig. 13 HX Primary Side Inlet and Outlet Temperature (SS)

Fig. 14 HX Secondary Side Inlet and Outlet temperature (SS)
5-1

5. TRANSIENT CALCULATION

5.1 Transient Calculation Results

The “Loss of Main Feed Water” accident scenario was assumed in this section. Before 500s it 

was steady state. Main feed water was cut off at 500s. Auxiliary feed water was delayed by 

1000s. Reactor scrammed 120s after the initiation of accident. Before reactor trip, primary side 

coolant flow increased and core outlet temperature dropped obviously. It resulted in reactor 

power increasing about 1%FP due to reactivity feedback of coolant temperature. After reactor 

trip, the balance of decay heat generating and the coolant heat capacity in the pool dominated 

the accident process.

Fig. 15 Reactor Power (Transient)
5-2

 

Fig. 16 Primary Coolant Flow (Transient)

Fig. 17 Core Inlet and Outlet Temperature (Transient)
5-3

Fig. 18 Rod and Clad temperature (Transient)

5.2 Sensitivity Analysis

In a short period after the loss of main feed water and before reactor trip, primary side natural 

circulation coolant flow increased significantly. As core inlet coolant temperature didn’t increase 

much at that time, fuel cooling capability became better even than normal operation condition, 

core outlet temperature dropped consequently. It is very interesting to find that reactor power 

increased, main feed water got lost, but core outlet temperature dropped. In order to get better 

understanding of this specific phenomenon, sensitivity analysis was performed in two cases. 

One is feed water sensitivity to natural circulation flow; the other is reactor power sensitivity to 

natural circulation flow.

Sensitivity analysis results indicate that primary natural circulation flow increases significantly as 

feed water flow decreases as well as reactor power increases. As core inlet coolant temperature 

stayed almost constant, core cooling capability temporarily increased shortly after the loss of 

main feed water. This is an important feature special for coolant flow of natural circulation 

system.
5-4

Fig. 19 Feed Water Sensitivity to Natural Circulation Flow

Fig. 20 Reactor Power Sensitivity to Natural Circulation Flow
6-1

6. TRANSIENT ANIMATION

The “Loss of Main Feed Water” accident scenario was animated in this section. SNAP 2.0.8 was 

used to animate the accident scenario. Fluid conditions and key parameter trends such as 

temperature and mass flow 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. 21 Transient Scenario Animation

7-1

7. CONCLUSIONS

In this report, Natural Circulation Reactor was successfully modeled by SNAP and TRACE 

code. The model was mainly made up of pipe, fill, break 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 parameter trends 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 natural circulation reactor system and the 

calculation results are reasonable based on engineering judgment.

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 domestic PWR NPP 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.




UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA

-0418 Application of TRACE V5.0 P2 to 

Natural Circulation Reactor Safety Analysis