NUREG/IA-0404 

International 

Agreement Report

The Development and Assessment of TRACE Model 

for Maanshan Nuclear Power Plant LOCA

 

Prepared by:

Jong-Rong Wang, Hao-Tzu Lin, Jung-Hua Yang*, Chunkuan Shih*

 

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

 1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, TAIWAN

 

*Institute of Nuclear Engineering and Science, National Tsing Hua University,

101 Section 2, Kuang Fu Rd., HsinChu, TAIWAN

 

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: September 2011

Date Published: September 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
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NUREG/IA-0404 

International 

Agreement Report

The Development and Assessment of TRACE Model 

for Maanshan Nuclear Power Plant LOCA

 

Prepared by:

Jong-Rong Wang, Hao-Tzu Lin, Jung-Hua Yang*, Chunkuan Shih*

 

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

 1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, TAIWAN

 

*Institute of Nuclear Engineering and Science, National Tsing Hua University,

101 Section 2, Kuang Fu Rd., HsinChu, TAIWAN

 

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: September 2011

Date Published: September 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 U.S. NRC is developing an advanced thermal hydraulic code named TRACE for safety 

analyses of nuclear power plants (NPPs). According to TRACE user’s manual, 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. 

SNAP is a program with graphic user interface, which processes the input and output of TRACE. 

As TRACE user’s manual describes, TRACE has a greater simulation capability than the other 

old codes, especially for events like loss-of-coolant accident (LOCA). 

The Maanshan NPP operated by Taiwan Power Company is the only Westinghouse-PWR in 

Taiwan. The rated core thermal power is 2775 MWt. The reactor coolant system has three loops, 

each of which includes a reactor coolant pump and a steam generator. The pressurizer is 

connected to the hot-leg piping in loop 2. The Maanshan NPP TRACE model has been 

successfully established [1]-[3]. This study focuses on the establishment of the TRACE large 

break LOCA (LBLOCA) model for Maanshan NPP. Then, the TRACE LBLOCA analysis results 

compare with the FSAR data [4]. In this study, the LBLOCA is defined as a double-ended 

guillotine in cold-leg. The break is located in loop 1, which is one of the two loops that don’t have 

a pressurizer. For a LBLOCA analysis, the most important parameter is the peak cladding 

temperature (PCT). As defined by the 10 CFR 50.46 regulation [5], the PCT does not exceed 

1477.6 K (2200°F). In this LBLOCA, the peak cladding temperature of TRACE calculated was 

1358.8 K (1986°F). 

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 LOCA model of Maanshan NPP has been built. In this report, the 

LBLOCA transient data of Maanshan NPP is utilized and conducted to confirm the accuracy of 

the TRACE model. 

vii 

CONTENTS

Page

Abstract.......................................................................................................................................... iii 

Foreword.........................................................................................................................................v

Executive Summary......................................................................................................................xi 

Abbreviations .............................................................................................................................. xiii 

1. Introduction .......................................................................................................................... 1-1 

2. Description of the codes and Maanshan TRACE LBLOCA model ................................. 2-1 

2.1 Codes............................................................................................................................ 2-1 

2.2 Maanshan TRACE LBLOCA model.............................................................................. 2-1 

2.2.1 Break’s simulation and Choked flow model............................................................ 2-1 

2.2.2 ECCS’ simulation.................................................................................................... 2-2 

2.2.3 CCFL model............................................................................................................ 2-3 

3. Assumptions and description of LBLOCA........................................................................ 3-1 

3.1 Blowdown...................................................................................................................... 3-1 

3.2 Refill .............................................................................................................................. 3-1 

3.3 Reflood.......................................................................................................................... 3-1 

4. Results and discussions..................................................................................................... 4-1 

4.1 The results of thermal-hydraulic parameters................................................................ 4-1 

4.2 Sensitivity studies for LBLOCA model.......................................................................... 4-1 

4.2.1 Accumulators nodalization...................................................................................... 4-2 

4.2.2 Pipe nodalization (near the break).......................................................................... 4-2 

4.2.3 Critical heat flux calculated in TRACE.................................................................... 4-2 

4.2.4 CCFL phenomena in downcomer........................................................................... 4-3 

5. Conclusions.................................................................................................................... 5-1 

6. References...................................................................................................................... 6-1 
viii 

Figures

Page

Fig.2.1 The LBLOCA TRACE model of Maanshan NPP: (a) Overall region 

(b) Vessel region .................................................................................................. 2-4 

Fig.2.2 The break simulation of TRACE LBLOCA model................................................ 2-6

Fig.2.3 The TRACE model of the ECCS control system for Maanshan NPP ................. 2-7 

Fig.2.4 The accumulator simulation of TRACE LBLOCA model ..................................... 2-7 

Fig.2.5 The LPSI simulation of TRACE LBLOCA model ................................................. 2-8 

Fig.2.6 The HPSI simulation of TRACE LBLOCA model................................................. 2-9

Fig.2.7 The LHSI flow rate correspond to the RCS pressure ........................................ 2-10 

Fig.2.8 The HHSI flow rate correspond to the RCS pressure........................................ 2-10 

Fig.3.1 The steam and liquid flow phenomena in vessel................................................. 3-3 

Fig.4.1 The comparisons of power between TRACE and FSAR data ........................... 4-7 

Fig.4.2 The comparisons of vessel pressure between TRACE and 

FSAR data............................................................................................................ 4-8 

Fig.4.3 The comparisons of break mass flow rate between TRACE 

and FSAR data..................................................................................................... 4-9 

Fig.4.4 The comparisons of accumlator mass flow rate between 

TRACE and FSAR data..................................................................................... 4-10 

Fig.4.5 The comparisons of core inlet flow rate between TRACE 

and FSAR data................................................................................................... 4-11 

Fig.4.6 The comparisons of core outlet flow rate between TRACE 

 and FSAR data................................................................................................... 4-12 

Fig.4.7 The blowdown , refill , and reflood phases of LBLOCA..................................... 4-13 

Fig.4.8 Animation of the TRACE LBLOCA model for Maanshan NPP.......................... 4-14 

Fig.4.9 The accumulators flow rate results of the accumulators nodalization 

sensitivity studies................................................................................................ 4-15 

Fig.4.10 The system pressure results of the accumulators nodalization 

sensitivity studies .......................................................................................... 4-16 

Fig.4.11 The break flow rate result of TRACE ................................................................. 4-17 

Fig.4.12 The break flow rate results of the pipe nodalization (near the break) 

sensitivity studies................................................................................................ 4-18 

Fig.4.13 The PCT results of the CHF model sensitivity studies ...................................... 4-19 

Fig.4.14 The vessel water level results of the CCFL model sensitivity 

 studies ................................................................................................................ 4-20 
ix 

Tables

Page

Table 3.1 The initial input parameters in TRACE model..................................................... 3-2 

Table 4.1 The comparison between the steady state data in LBLOCA of 

TRACE and Final safety analysis report (FSAR)................................................. 4-4 

Table 4.2 The LBLOCA sequences of TRACE and FSAR ................................................. 4-5

Table 4.3 The PCT comparison of the sensitivity studies ................................................... 4-6 

xi 

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 LOCA model of Maanshan 

NPP is developed by INER. 

According to the TRACE user’s manual [6], it 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. Therefore, in the future, NRC has ensured that TRACE 

will be the main code used in thermal hydraulic safety analysis, without further development of 

other thermal hydraulic codes such as RELAP5 and TRAC. Besides, the 3-D geometry model of 

reactor vessel is one of the features of TRACE. It can support a more accurate and detailed 

safety analysis of NPPs. TRACE has a greater simulation capability than the other old codes, 

especially for events like LOCA. 

The Maanshan NPP operated by Taiwan Power Company is the only Westinghouse-PWR in 

Taiwan. The rated core thermal power is 2775 MWt. The reactor coolant system has three loops, 

each of which includes a reactor coolant pump and a steam generator. The pressurizer is 

connected to the hot-leg piping in loop 2. The Maanshan NPP TRACE model has been 

successfully established [1]-[3]. This research focuses on the establishment of the TRACE 

LBLOCA model for Maanshan NPP. The codes used in this research are TRACE v 5.0p1 and 

SNAP v 2.0.3. Then, the TRACE LBLOCA analysis results compare with the FSAR data [4]. The 

main components of LBLOCA model include the pressure vessel, pressurizer, steam generators, 

steam piping in the secondary side, the steam dump system, accumulators, and safety injection 

of emergency core cooling system (ECCS). The TRACE LBLOCA model of Maanshan NPP is a 

three-loop model and each loop has a feedwater control system. The pressure vessel of the 

TRACE LBLOCA model is cylindrical. The pressure vessel is divided into 12 levels in the axial 

direction, two rings in the radial direction (internal and external rings) and six equal azimuthal 

sectors in the “θ” direction. The control rod conduit connects the 12th and 7th layers of the 

vessel from end to end. The fuel region is between the third and sixth layers, and heat 

conductors are added onto these structures to simulate the reactor core. In this study, the 

LBLOCA is defined as a double-ended guillotine in cold-leg. The break was located in loop 1, 

which is one of the two loops that don’t have a pressurizer. In transient analysis, the TRACE 

code simulates or calculates the core power by either two methods. One is a constant power or 

input the power curve by “power table” into the TRACE model. Another is to use “point kinetic” 

data (e.g., the delay neutron fraction, Doppler reactivity coefficient, and moderator temperature 

reactivity coefficient) in the TRACE model and let TRACE to calculate the solution of the point 

kinetic equations for the transient analysis. These point kinetic equations specify the time 

behavior of the core power with neutronic reactivity (include the sum of programmed reactivity 

and feedback reactivity). The feedback reactivity calculation of TRACE is based on changes in 

the core-averaged fuel temperature, coolant temperature, gas volume fraction, and boron 

concentration. In this research, we use the TRACE point kinetic to calculate the core power, and 

compare with the FSAR data. 
xii 

By using SNAP/TRACE, this study establishes the TRACE LBLOCA model of the Maanshan 

NPP. Analytical results indicate that the Maanshan NPP TRACE LBLOCA model predicts not 

only the behaviors of important plant parameters in consistent trends with the FSAR data [4], but 

also provides a greater margin for the PCT evaluation. The TRACE model of Maanshan NPP 

can be used in future safety analysis with confidence, such as the applications for different break 

size and other break locations in LOCA. As defined by the 10 CFR 50.46 regulation [5], the PCT 

does not exceed 1477.6 K (2200°F). In this LBLOCA, the peak cladding temperature of TRACE 

calculated was 1358.8 K (1986°F). 
xiii 

ABBREVIATIONS

 

BAF Bottom of the Active Fuel 

CAMP Code Applications and Maintenance Program 

CCFL Counter Current Flow Limitation 

CHF Critical Heat Flux 

ECCS Emergency Core Cooling System 

FSAR Final safety analysis report 

HHSI High Head Safety Injection 

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

LBLOCA Large Break LOCA 

LHSI Low Head Safety Injection 

LOCA Loss Of Coolant Accident 

NPP Nuclear Power Plant 

NRC Nuclear Regulatory Commission 

PCT Peak Cladding Temperature 

PWR Pressurized Water Reactor 

RCS Reactor Coolant System 

RCP Reactor Coolant Pumps 

SIS Safety Injection Signal 

SNAP Symbolic Nuclear Analysis Program 

TAF Top of the Active Fuel 

TRACE TRAC/RELAP Advanced Computational Engine 

US United States 

 

 

 

 

 

 

 

 

 

 

1-1 

1. INTRODUCTION

The U.S. NRC is developing an advanced thermal hydraulic code named TRACE for safety 

analyses of NPPs. According to the reference [6], it 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. Therefore, in the future, NRC has 

ensured that TRACE will be the main code used in thermal hydraulic safety analysis, without 

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

The Maanshan NPP operated by Taiwan Power Company is the only Westinghouse-PWR in 

Taiwan. The rated core thermal power is 2775 MWt. The reactor coolant system has three loops, 

each of which includes a reactor coolant pump and a steam generator. The pressurizer is 

connected to the hot-leg piping in loop 2. The Maanshan NPP TRACE model has been 

successfully established [1]-[3]. This current research focuses on the establishment of the 

TRACE LBLOCA model for Maanshan NPP. Then, the TRACE LBLOCA analysis results 

compare with the FSAR data [4]. The main components of LBLOCA model include the pressure 

vessel, pressurizer, steam generators, steam piping in the secondary side, the steam dump 

system, accumulators, and safety injection of ECCS. In this study, the LBLOCA is defined as a 

double-ended guillotine in cold-leg. The break is located in loop 1, which is one of the two loops 

that don’t have a pressurizer. For a LBLOCA analysis, the most important parameter is PCT. 

As defined by the 10 CFR 50.46 regulation [5], the PCT does not exceed 1477.6K (2200°F). In 

this LBLOCA, the peak cladding temperature of TRACE calculated was 1358.8K (1986°F).

2-1 

2. Description of the codes and Maanshan TRACE LBLOCA model 

2.1 Codes 

The codes used in this research are TRACE v 5.0p1 and SNAP v 2.0.3. SNAP is a program with 

graphic user interface which processes the inputs and outputs of TRACE. 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 NPPs. As the reference [6] describes, TRACE has a 

greater simulation capability than the other old codes, especially for events like LOCA. Models 

for multidimensional two-phase flow, non-equilibrium thermo-dynamics, heat transfer, reflood, 

level tracking, and reactor kinetics are also included in TRACE. 

The results of TRACE LBLOCA analysis are compared with the FSAR data of Maanshan NPP. 

There are four codes used for LBLOCA analysis in FSAR, which are SATAN, BASH, COCO, 

and LOCBART. SATAN is a one-dimensional nodal network code which models the 

thermal-hydraulic phenomena during the blowdown depressurization of the reactor coolant 

system after a postulated large rupture of a primary coolant pipe. The BASH code is used to 

calculate the refill and reflood portions of the LBLOCA. The COCO code can calculate the 

pressure and temperature transients inside the containment during the depressurization and 

post-blowdown phase following a LOCA. The LOCBART code is used to calculate the hot rod 

temperature during the blowdown, refill, and reflood phases of the LBLOCA. 

2.2 Maanshan TRACE LBLOCA model 

Fig 2.1 shows the TRACE LBLOCA model of Maanshan NPP. It is a three-loop model and each 

loop has a feedwater control system. The main structure of this model includes the pressure 

vessel, pressurizer, steam generators, steam piping at the secondary side, the steam dump 

system, accumulators, and safety injection of ECCS. The pressure vessels are cylindrical, and 

its divisions are as shown in Fig 2.1. It is divided into 12 levels in the axial direction, two rings in 

the radial direction (internal and external rings) and six equal azimuthal sectors in the “θ” 

direction. The control rod conduit connects the 12th and 7th layers of the vessel from end to end. 

The fuel region is between the third and sixth layers, and heat conductors were added onto 

these structures to simulate the reactor core. 

Before any transient analysis can begin, a consistent set of parameters used in the TRACE 

model must be obtained in the process of steady-state initialization. In transient analysis, the 

TRACE code simulates or calculates the core power by either two methods. One is a constant 

power or input the power curve by “power table” into the TRACE model. Another is to use “point 

kinetic” data (e.g., the delay neutron fraction, Doppler reactivity coefficient, and moderator 

temperature reactivity coefficient) in the TRACE model and let TRACE to calculate the solution 

of the point kinetic equations for the transient analysis. These point kinetic equations specify the 

time behavior of the core power with neutronic reactivity (include the sum of programmed 

reactivity and feedback reactivity). The feedback reactivity calculation of TRACE is based on 

changes in the core-averaged fuel temperature, coolant temperature, gas volume fraction, and 

boron concentration. In this research, we use the TRACE point kinetic to calculate the core 

power, and compare with the FSAR data. 

2.2.1 Break’s simulation and Choked flow model 

In this study, the LBLOCA is defined as a double-ended guillotine in cold-leg. According to FSAR 

describe, a major pipe break (large break) is defined as a rupture with a total cross sectional 

area equal to or greater than 1.0 ft2

. In this LBLOCA, the cross sectional area is 4.17 ft

2

. Fig 2.2 

shows the initial and boundary conditions of break component in the TRACE LBLOCA model. 

The ECCS is designed to prevent fuel cladding damage. Following a postulated double ended 
2-2 

rupture of a reactor coolant pipe, the ECCS limits the clad temperature to well below the melting 

point and ensures that the reactor core remains intact and in a coolable geometry. 

The break mass flow is one of the most important parameters during the simulation of LBLOCA. 

In this study, the choked flow model is used in the break of Maanshan NPP TRACE LBLOCA 

model which references to our previous study [7]. In our previous study, the TRACE IIST facility 

model also used the choked flow mode to simulate the IIST facility 2% cold-leg-break LOCA 

experiment. The analytical results of TRACE IIST facility models indicate that the TRACE IIST 

facility models predict not only the behaviors of important parameters in consistent trends with 

experiments data, but also their numerical values with respectable accuracy. Choked flow 

occurs when the mass flow in a pipe becomes independent of the downstream conditions [8]. 

Choked flow can also be defined in single-phase system with sonic velocity, choking velocity 

equals sonic velocity. This means that pressure signals can’t be transmitted to higher pressure 

upstream anymore, because transmission speed can’t exceed sonic velocity. The choked flow 

model used in TRACE is actually compounded by three different models, which are the 

subcooled-liquid choked-flow model, the two-phase, two-component choked-flow model, and the 

single-phase vapor choked-flow model. For subcooled single-phase, TRACE applies a modified 

form of the Burnell model and is essentially the same as that used in RELAP5. For the 

two-phase, two-component choked flow model, an extension of a model developed by Ransom 

and Trapp is used. Finally, the single-phase vapor choked flow model is based on isentropic 

expansion of an ideal gas. The TRACE model incorporates an additional inertgas component 

and nonequilibrium effects. In addition, the transition from liquid to two-phase flow, which 

presents a discontinuity, is handled in TRACE by linear interpolation between the subcooled and 

the two-phase regimes. The best value for the coefficient of the choked model was 0.5 in this 

study. 

2.2.2 ECCS’ simulation 

The ECCS includes the accumulators, low head safety injection (LHSI), and high head safety 

injection (HHSI). Fig 2.3 shows the TRACE model of the ECCS and its control systems for 

Maanshan NPP. After the break occurs, low pressurizer pressure caused the ECCS injecting 

water to the reactor coolant system (RCS) cold-leg and preventing excessive clad temperatures. 

As the reactor coolant system pressure decreased below 4.24MPa, accumulator injection started. 

The cold water in the accumulator was expelled into the reactor coolant system by nitrogen gas 

filled in the accumulator. Fig 2.4 shows the simulation of the accumulators by using pipe 

components. The total volume and height were 41.1m3

 and 5.72m, respectively. 

In a LBLOCA of a PWR, most of the initial reactor coolant inventory is rapidly expelled through 

the break and the pressure of the primary system decreases causing most of the liquid inventory 

to flash into steam. After the break occurs, the water of ECCS begins to be injected into the 

reactor coolant system. The purpose of the ECCS injection is to rapidly refill the reactor vessel 

lower plenum and to reflood the reactor core. When the pressure has decreased below the 

safety injection setpoint (11.8MPa), LHSI and HHSI begin to be injected into the reactor coolant 

system with a delay time of 27 sec. Fig 2.5 and Fig 2.6 show the LHSI and HHSI simulation of 

TRACE LBLOCA model. Fig 2.7 and Fig 2.8 show the fill table of LHSI and HHSI which is the 

safety injection flow rate correspond to the RCS pressure. 

2.2.3 CCFL model 

Counter Current Flow Limitation (CCFL) is an important issue related to the safety analysis of 

PWRs. CCFL may occur in the downcomer, the upper core tie plate, the hot legs, the entrance 

of the steam generator inlet plenum, and the pressurizer surge line, where the flow direction or 
2-3 

flow area changes. The CCFL correlations can be represented as Wallis type, Kutateladze type, 

and Bankoff type [8]. In TRACE, the CCFL model basically uses the Bankoff correlation. The 

Bankoff correlation can revert to the Wallis type or Kutateladze type by setting the interpolation 

constant β (β=0 is Wallis type; β=1 is Kutateladze type). The Equation of the Bankoff correlation 

is as follow: 

Hg

1/2 + mHl1/2 = C 

where Hk is dimensionless mass flux (k=gas or liquid), m and C are constants determined from 

the experiments. 

In this study, the CCFL model used the Kutateladze type by setting β = 1. Two correlation 

constants, m and C, are imported by 1.24 and 1.5, respectively. 
2-4 

(a) Overall region
2-5 

(b) Vessel region

Fig. 2.1 The LBLOCA TRACE model of Maanshan NPP.
2-6 

Fig 2.2 The break simulation of TRACE LBLOCA model.
2-7 

Fig 2.3 The TRACE model of the ECCS control system for Maanshan NPP. 

Fig 2.4 The accumulator simulation of TRACE LBLOCA model. 
2-8 

Fig 2.5 The LHSI simulation of TRACE LBLOCA model. 
2-9 

Fig 2.6 The HHSI simulation of TRACE LBLOCA model. 
2-10 

0.0 0.5 1.0 1.5 2.0

0

10

20

30

40

50

60

70

80

90

100

Mass flow rate (kg/s)

Pressure (MPa)

 LHSI flow rate

Fig 2.7 The LHSI flow rate correspond to the RCS pressure. 

0 10 20 30

0

5

10

15

20

25

Mass flow rate (kg/s)

Pressure (MPa)

 HHSI flow rate

Fig 2.8 The HHSI flow rate correspond to the RCS pressure. 
3-1

3. Assumptions and description of LBLOCA 

By the conservative assumptions of analyzing the LOCA [4], [9]-[10], it must be assumed that the 

reactor has been operating continuously at a power level at least 1.02 times the licensed power 

level. Table 3.1 shows the initial input parameters in Maanshan LBLOCA TRACE model. 

The analysis of a LBLOCA is divided into three phases: (1) blowdown, (2) refill, and (3) reflood. 

3.1 Blowdown 

Before the break occurs, the plant is in an equilibrium condition. Should a major break occur, 

depressurization of the RCS results in a pressure decrease in the pressurizer. The reactor trip 

signal subsequently occurs when the pressurizer low pressure trip setpoint is reached. During 

blowdown, heat from fission product decay, hot internals and the vessel continues to be 

transferred to the reactor coolant. As the core becomes uncovered, both turbulent and laminar 

forced convection and radiation are considered as core heat transfer mechanisms. Steam water 

counter-current flow blocks ECC flow in downcomer and turns it towards the broken loop instead 

of the reactor core (see Fig 3.1). Heat stored in the vessel walls boils ECC water and 

strengthens the bypass phenomenon. The important parameter of peaking cladding temperature 

depends on the velocity of blowdown phase in LOCA analysis. The combination of 1D and 3D 

components in TRACE allows an accurate modeling of complex flow networks as well as local 

multidimensional flows. This is important in determining of the flow velocity and primary pressure 

during blowdown. 

3.2 Refill

When the RCS pressure is the same with the pressure of the containment, the blowdown phase 

of the transient terminates. During the refill phase, the mechanisms that are responsible for the 

bypassing of emergency core cooling water injected into the RCS are calculated not to be 

effective. At this time refill of the reactor vessel lower plenum begins. Refill is complete when 

emergency core cooling water has filled the lower plenum of the reactor vessel which is bounded 

by the bottom of the fuel rods. 

3.3 Reflood

The reflood phase of the transient is defined as the time period lasting from the end of refill until 

the reactor vessel has been filled with water to the extent that the core temperature rise has 

been terminated. From the later stage of blowdown, and then the beginning of reflood, the safety 

injection accumulator tanks rapidly discharge borated cooling water into the RCS, contributing to 

the filling of the reactor vessel downcomer. The downcomer water elevation head provides the 

driving force required for the reflooding of the reactor core. The low head and high head safety 

injection pumps aid the filling of the downcomer and subsequently supply water to maintain a full 

downcomer and complete the reflooding process. 
3-2

Table 3.1 The initial input parameters in TRACE model. 

Input Parameters Value 

Licensed Core Power 102% of total thermal power

Loop Flow ( kg/sec ) 4331.8 

Vessel Average Temperature (K) 584.5 

Initial RCS Pressure (MPa) 15.8 

Low Pressurizer Pressure Reactor Trip Setpoint (MPa) 12.8 

Low Pressurizer Pressure SI Setpoint (MPa) 11.8 

Safety Injection Initiation Delay time with loss of offsite power (sec) 27 

Accumlator Water Volume (m3/tank) 27.9 

Accumlator Tank Volume (m3/tank) 41.1 

Minimum Accumlator Gas Pressure (MPa) 4.24 

Accumlator Water Temperature (K) 311 

Nominal RWST Water Temperature (K) 302.6 
3-3

Fig 3.1 The steam and liquid flow phenomena in vessel. 

4-1

4. Results and discussions 

In TRACE, steady-state initialization was performed. The parameters’ results such as the power, 

the pressure of the pressurizer, the Tavg temperature, and the RCS flow rate are compared with 

FSAR data [4]. Table 4.1 shows the comparison between the steady-state results of the TRACE 

and FSAR. The results are clearly mutually quite consistent. 

4.1 The results of thermal-hydraulic parameters 

Following the steady state initialization, the TRACE LBLOCA model’s transient predicted results 

are compared with the FSAR data. Table 4.2 presents the sequence of LBLOCA and the timings 

of the LBLOCA predicted by TRACE. The sequence of TRACE arose from the actuation of the 

related control system, which in turn had to be actuated by physical parameter signals. If the 

parameters predicted by TRACE differ from the FSAR data, then the event sequences will also 

differ. Such deviations can be observed in the comparisons of transient event analyses. 

Fig. 4.1 plots the power curve that calculated from TRACE in the case of LBLOCA, and then 

compares with the FSAR data. In TRACE, the core power can be calculated using the built-in 

point kinetics model, and the power calculated includes decay heat. It displays that the power 

curve of TRACE is almost the same as those of FSAR data. Fig. 4.2 compares the pressures of 

the vessel and suggests that the pressure calculated by TRACE approximately follows the trend 

of the FSAR data. Fig. 4.3 compares the break mass flow rate of cold-leg in loop 1. It reveals 

that break mass flow rate predicted by TRACE agrees closely with the results of the FSAR data. 

Fig. 4.4 shows the comparisons of accumulator mass flow rate of intact loops between TRACE 

model and FSAR data. Fig. 4.5 compares the core inlet flow rate, revealing that the flow rate 

calculated by TRACE is in agreement with the FSAR data except for the period between 6 and 

18 sec. It reveals that the flow rate calculated by TRACE is slightly lower between 6 and 18 sec. 

Fig. 4.6 plots the results for core outlet flow rate. The difference results of core outlet flow before 

6 sec are consideration of the nature flow in TRACE. Fig. 4.7 shows TRACE’s vessel water level 

and PCT results. It also reveals the blowdown, refill, and reflood phases of LBLOCA. In the 

blowdown phase, the vessel water level dropped sharply after the break occurred. The vessel 

water level was lower than the BAF (Bottom of the Active Fuel). At the same time, due to the 

vessel water level decreased, the PCT increased. In the refill phase, the vessel water level 

raised. The Refill phase is complete when the water has filled the lower plenum of the reactor 

vessel which is bounded by the bottom of the fuel rods. The PCT reached the max. value in the 

refill phase. As defined by the 10 CFR 50.46 regulation [5], the PCT does not exceed 1477.6 K 

(2200°F). In this LBLOCA, the peak cladding temperature of TRACE calculated was 1358.8 K 

(1986°F). Finally, in the reflood phase, the vessel water level was higher than the BAF and the 

PCT dropped. 

Furthermore, the animation of this TRACE LBLOCA model is presented using the animation 

function of SNAP/TRACE interface with above models and analysis results. The animation 

model of Maanshan NPP is shown in Fig. 4.8. 

4.2 Sensitivity studies for LBLOCA model

The work performed to improve the Maanshan LBLOCA model was focused on following 

sensitivity studies. The best set of the sensitivity studies have been used in the TRACE model 

and the analysis results are described in section 4.1. 

4.2.1 Accumulators nodalization 

In LBLOCA analysis, ECCS play an important role in preventing fuel cladding damage and 
4-2

ensuring the core maintains integrity. In Maanshan NPP, there is one accumulator in each loop. 

The accumulators provide high flow borated water in a short time after the system pressure 

drops below 4.24 MPa. The nodalization of the accumulators becomes especially important in 

this LOCA analysis [11]. 

The initial nodalization of the accumulator had three cells, simulating nitrogen gas layer, coolant 

water layer, and exit layer, respectively. In order to enhance the accuracy of TRACE predictions, 

the number of cells was increased to five cells. The accumulator geometry was still in 41 m3 of 

total volume and 5.8 m of height, respectively. The mass flow rate of the accumulator injection is 

shown in Fig. 4.9. The time of accumulator injection is earlier in the case of three cells because 

of faster pressure decrease (see Fig. 4.10). Fig 4.9 indicates that accumulator mass flow rates 

calculated by TRACE in the case with 5 cells and 7 cells approximately follows the trend of the 

FSAR data. In the PCT’s comparison, the 5 cells case is higher than 3 cells case and is roughly 

the same with 7 cells case. The above results suggest that using 5 cells of accumulator is good 

enough to predict the trend of accumulator injection. 

4.2.2 Pipe nodalization (near the break) 

One of the most sensitive parameters during the simulation of a LOCA is certainly the evolution 

of the break mass flow. In order to determine the velocity and the pressure at the throat of the 

break, the conditions at the cell-edge, where the choking criterion is applied must be known. The 

homogeneous equilibrium sound speed is calculated to estimate the corresponding cell-edge 

conditions, given the conditions at the cell center. The results of predicting for the TRACE 

choked flow model during the blowdown are presented in Fig 4.11. At approximately 2.7 

seconds, the calculated break flow starts to transition from subcooled liquid flow to two-phase 

flow. 

The initial nodalization of the pipe (near the break) had only one cell. In order to enhance the 

accurately of TRACE predicted, the pipe component was split to two cells. Fig 4.12 shows the 

mass flow rate of break. There are some significant differences between the 1 cell and 2 cells in 

TRACE calculation from 9 sec to 15 sec. As for the case with 2 cells, it is seen that the break 

flow rate from TRACE prediction is in better accuracy, compared with FSAR. The maximum 

mass flow rate in TRACE prediction is almost the same between the 1 cell and 2 cells. As for the 

case with 2 cells, it is seen that the break flow rates predicted by TRACE agrees closely with the 

results of the FSAR data. In the PCT’s comparison, the 2 cells case is is roughly the same with 1 

cell case. The above results indicate that using 2 cells case is good for the TRACE model.

4.2.3 Critical heat flux calculated in TRACE 

Critical heat flux (CHF) is the point where the maximum heat flux occurs in the idealized boiling 

curve. In TRACE [8], the role of the CHF model is two-fold: 1. Determine the transition point for 

the heat transfer regime; 2. Serve as the anchor point for the transition boiling wall heat flux. To 

serve both these roles, the CHF model in TRACE must provide a continuous estimate of the 

CHF over a wide range of conditions with reasonable accuracy. 

According to TRACE manual [8], the 1995 AECL-IPPE CHF look-up table was selected for the 

default CHF model in TRACE. It is based on an extensive database of CHF values obtained in 

tubes with a vertical upflow of a steam-water mixture and provides the value of the critical heat 

flux as a function of the local conditions. This method of determining the value of the critical heat 

flux was selected for TRACE because of its reasonably good accuracy and wide range of 

applicability. Two critical quality correlations were available in the TRAC-B code and have been 

incorporated into TRACE as part of the code consolidation process: the CISE-GE correlation 

and a model derived from the Biasi CHF correlation. Fig 4.13 shows that sensitivity study of PCT 
4-3

prediction. In LBLOCA analysis, the PCT calculated by AECL-IPPE table with critical quality from 

CISE-GE correlation shows the highest peak cladding temperature. The temperatures predicted 

by table and the critical quality from Biasi correlation are almost the same. In this research, the 

AECL-IPPE table with critical quality from CISE-GE correlation was selected in the TRACE 

model. 

4.2.4 CCFL phenomena in downcomer 

In the reactor vessel, CCFL can occur during blowdown as ECC liquid is attempting to fill the 

downcomer (see fig 3.1). Steam is generated which rises upwards and reduces penetration of 

ECC fluid to the core. CCFL sets limits for the ECC penetration and determines how effectively 

the reactor pressure vessel is refilled and how much of the ECC fluid is bypassed directly to the 

broken loop. As the steam generation decreases the ECC fluid starts to fill the lower plenum 

(Nuclear Regulatory Commission, 1988). Fig 4.14 indicates that TRACE with CCF model 

predicted the lower reactor water level between 18 and 30s. The results of peak cladding 

temperature that analyzed with CCF model could increase for 18K.

Finally, the PCT comparison of the sensitivity studies lists in Table 4.3. 
4-4

Table 4.1 The comparison between the steady state data in LBLOCA of TRACE and Final 

safety analysis report (FSAR). 

Parameter FSAR TRACE Error (%) 

Power (MWt) 2830 2830 0 

Tavg * (K) 584.5 584.53 0.0001 

Pressurizer pressure (MPa) 15.858 15.859 0.0001 

Loop Flow (kg/sec) 4331.8 4347 0.0035 

*Tavg = (Hot-leg temperature +Cold-leg temperature)/2 
4-5

Table 4.2 The LBLOCA sequences of TRACE and FSAR. 

LBLOCA 

FSAR 

(sec) 

TRACE 

(sec) 

Break began 

Reactor scram setpoint reached 

SI signal generated 

Accumulators Injection 

Start of Pumped SI 

Accumulators empty 

0.0 

0.50 

1.4 

15.0 

28.4 

52.1 

0.0 

0.50 

1.5 

14.2 

28.5 

59.5
4-6

Table 4.3 The PCT comparison of sensitivity studies.

PCT

Case 3 cells 5 cells 7 cells 

Accumulators nodalization base large than 3 cells 

case

large than 3 cells 

and roughly the 

same with 5 cells 

case

Case 1 cell 2 cells 

Pipe nodalization 

(near the break) 

base roughly the same with 1 cell 

case 

Case AECL-IPPE table AECL-IPPE table 

with Biasi 

correlation 

AECL-IPPE table 

with CISE-GE 

correlation

Critical heat flux base roughly the same 

with AECL-IPPE 

table case 

larger than 

AECL-IPPE table 

case 

Case No use CCFL Use CCFL 

CCFL base larger than no use CCEL 

case 
4-7

0 5 10 15 20 25 30 35

Time (sec)

0

0.2

0.4

0.6

0.8

1

1.2

P

o

w

er (n

o

m

aliz

e

d)

FSAR

TRACE

Fig. 4.1 The comparisons of power between TRACE and FSAR data. 
4-8

0 5 10 15 20 25 30 35

Time (sec)

0

4

8

12

16

20

Pre

s

s

ure (M

P

a)

FSAR

TRACE

Fig. 4.2 The comparisons of vessel pressure between TRACE and FSAR data. 
4-9

0 5 10 15 20 25 30 35

Time (sec)

0

10000

20000

30000

Flo

w rate (k

g/s

e

c)

FSAR

TRACE

Fig. 4.3 The comparisons of break mass flow rate between TRACE and FSAR data. 
4-10

0 5 10 15 20 25 30 35

Time (sec)

0

500

1000

1500

2000

2500

Flo

w rate (k

g/s

e

c)

FSAR

TRACE

Fig. 4.4 The comparisons of accumlator mass flow rate between TRACE and FSAR data.
4-11

0 5 10 15 20 25 30 35

Time (sec)

-8000

-4000

0

4000

8000

12000

16000

Flo

w rate (k

g/s

e

c)

FSAR

TRACE

Fig. 4.5 The comparisons of core inlet flow rate between TRACE and FSAR data.
4-12

0 5 10 15 20 25 30 35

Time (sec)

-4000

0

4000

8000

12000

16000

Flo

w rate (k

g/s

e

c)

FSAR

TRACE

Fig. 4.6 The comparisons of core outlet flow rate between TRACE and FSAR data.
4-13

0 100 200 300

Time (sec)

0

4

8

12

16

V

e

s

s

el w

ater le

v

el (m)

600

800

1000

1200

1400

P

C

T (K)

Water level

BAF

TAF

PCT

Blowdown Refill Reflood

Fig 4.7 The blowdown , refill , and reflood phases of LBLOCA. 
4-14

Fig. 4.8 Animation of the TRACE LBLOCA model for Maanshan NPP. 
4-15

0 5 10 15 20 25 30 35

Time (sec)

0

500

1000

1500

2000

2500

Flo

w rate (k

g/s

e

c)

FSAR

TRACE: 3 cells case

TRACE: 5 cells case

TRACE: 7 cells case

Fig 4.9 The accumulators flow rate results of the accumulators nodalization sensitivity studies. 
4-16

0 5 10 15 20 25 30 35

Time (sec)

0

4

8

12

16

20

Pre

s

s

ure (M

P

a)

FSAR

TRACE: 3 cells case

TRACE: 5 cells case

TRACE: 7 cells case

Fig 4.10 The system pressure results of the accumulators nodalization sensitivity studies. 
4-17

0 5 10 15 20 25 30 35

Time (sec)

0

500

1000

1500

2000

2500

Bre

a

k flo

w rate (k

g/s

e

c)

Total mass flow

Liquid mass flow

Gas mass flow

Fig 4.11 The break flow rate result of TRACE.
4-18

0 5 10 15 20 25 30 35

Time (sec)

0

10000

20000

30000

Flo

w rate (k

g/s

e

c)

FSAR

TRACE: 1 cell case

TRACE: 2 cells case

Fig 4.12 The break flow rate results of the pipe nodalization (near the break) sensitivity studies. 
4-19

0 100 200 300 400 500 600 700

Time (sec)

200

400

600

800

1000

1200

1400

P

C

T (K)

AECL-IPPE Table

AECL-IPPE Table with Biasi correlation

AECL-IPPE Table with CISE-GE correlation

Fig 4.13 The PCT results of the CHF model sensitivity studies. 
4-20

0 20 40 60 80 100

Time (sec)

0

2

4

6

V

e

s

s

el w

ater le

v

el (m)

Without CCFL model

With CCFL model

Fig 4.14 The vessel water level results of the CCFL model sensitivity studies. 
5-1

5. CONCLUSIONS

By using SNAP/TRACE, this study establishes the TRACE LBLOCA model of the Maanshan 

NPP. Analytical results indicate that the Maanshan NPP TRACE LBLOCA model predicts not 

only the behaviors of important plant parameters in consistent trends with the FSAR data, but 

also provides a greater margin for the PCT evaluation. The TRACE model of Maanshan NPP 

can be used in future safety analysis with confidence, such as the applications for different break 

size and other break locations in LOCA. 

6-1 

6. REFERENCES

1. Wang, J.R., et al., “TRACE modeling and its verification using Maanshan PWR start-up 

tests”, Annals of nuclear energy, Vol.36, pp.527-536, 2009. 

2. Wang, J.R., et al., “TRACE analysis of Maanshan PWR for turbine trip test”, ANS 2008 

Winter Meeting and Nuclear Technology Expo, 2008. 

3. Wang, J.R., et al., “Maanshan PWR loss of flow transients analysis with TRACE”, 

Proceedings of the 17th International Conference on Nuclear Engineering, ICONE17, 2009. 

4. Taiwan Power Company, “Final safety analysis report of Maanshan nuclear power station 

Units 1 & 2”, 1982. 

5. Liang, T. K.S., Chang, C.J., Hung, H.J., ”Development of LOCA licensing calculation 

capability with RELAP5-3D in accordance with Appendix K of 10 CFR 50.46”, Nuclear 

Engineering and Design, Vol. 211, pp.69–84, 2002. 

6. NRC, “TRACE V5.0 user’s manual”, Office of Nuclear Regulatory Research, 2010. 

7. Wang, J.R., et al., “The development and verification of TRACE model for IIST 

experiments”, NUREG/IA-0252, 2011. 

8. NRC, “TRACE V5.0 theory manual, Division of Safety Analysis”, Office of Nuclear 

Regulatory Research, 2010. 

9. Marzo, M.D., Bessette, D.E.,”Effect of depressurization on reactor vessel inventory in the 

absence of ECCS injection”, Nuclear Engineering and Design, Vol.193, pp.197–205, 1999. 

10. Haste, T.J., Birchley, J., Richner, M.,”Accident management following loss-of-coolant 

accidents during cooldown in a Westinghouse two-loop PWR”, Nuclear Engineering and 

Design, NED-5651, 2010. 

11. Wang, S.J., Chien, C.S., Chiang, S.C.,”Development of accumulator computational aid for 

determining RCS injection volume”, Nuclear Engineering and Design, Vol. 236, 

pp.330–1333, 2006. 



UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS

NUREG/IA

-0404 The Development and Assessment of TRACE Model for 

Maanshan Nuclear Power Plant LOCA