NUREG/IA-0476

Main Steam Line Break Analysis for 

Lungmen ABWR

Prepared by:

Chunkuan Shih, Jong-Rong Wang, Ai-Ling Ho, Shao-Wen Chen, Show-Chyuan Chiang*, Tzu-Yao 

Yu*

Institute of Nuclear Engineering and Science, National Tsing Hua University; Nuclear and New 

Energy Education and Research Foundation

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

*Department of Nuclear Safety, Taiwan Power Company

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

K. Tien, NRC Project Manager

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: February 2016

Date Published: November 2016

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission

NUREG/IA-0476

Main Steam Line Break Analysis for 

Lungmen ABWR

Prepared by:

Chunkuan Shih, Jong-Rong Wang, Ai-Ling Ho, Shao-Wen Chen, Show-Chyuan Chiang*, Tzu-Yao 

Yu*

Doosan Heavy Industries & Construction

22, DoosanVolvo-ro, Seongsan-gu,

Changwon, Gyeongnam, 642-792, Korea

*Korea Institute of Nuclear Safety

62 Gwahak-ro, Yuseong-gu

Daejeon, 34142, Korea

K. Tien, NRC Project Manager

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: February 2016

Date Published: November 2016

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 object of this paper is to develop methodologies for analyzing the behaviors of fuel rod, 

vessel, and containment during main steamline break (MSLB) transient. The broken area of the 

RPV side was assumed to be 0.0984m2 (flow limiter). And the broken area of the main steam 

header side was assumed to 0.319m2 (main steam line area). According to FSAR, for 

conservative assumption, MSIVs started to close at 0.5sec and fully closed at 5.0sec after the 

transient started. The results of TRACE/PARCS coupling calculation were compared with those 

of both FSAR and GOTHIC data, indicating that the TRACE/PARCS coupling model has the 

ability to predict the MSLB transient, and both RPV integrity and containment integrity criteria are 

met. After that, the output data from TRACE/PARCS calculation was put into FRAPTRAN code 

as boundary conditions to analyze the thermo-mechanical behavior and calculate the stress, 

strain, oxide thickness, etc. The values of these factors were compared with the criteria. And the 

final results show that the fuel rod integrity criteria are met.

v 

FOREWORD

The US NRC is developing an advanced thermal hydraulic code named TRACE for nuclear power 

plant safety analysis. The development of TRACE is based on TRAC, integrating RELAP5 and 

other programs. NRC has determined that in the future, TRACE will be the main code used in 

thermal hydraulic safety analysis, and no further development of other thermal hydraulic codes 

such as RELAP5 and TRAC will be continued. A graphic user interface program, SNAP which 

processes inputs and outputs for TRACE is also under development. One of the features of 

TRACE is its capacity to model the reactor vessel with 3-D geometry. It can support a more 

accurate and detailed safety analysis of nuclear power plants. TRACE 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. To meet 

this responsibility, the TRACE/PARCS/FRAPTRAN model of Lungmen NPP has been built. In 

this report, the TRACE/PARCS/FRAPTRAN model of Lungmen NPP was used to evaluate the 

Lungmen main steamline break transient.

vii

TABLE OF CONTENTS

ABSTRACT...............................................................................................................................III

FOREWORD ............................................................................................................................. V

LIST OF FIGURES ................................................................................................................... IX

EXECUTIVE SUMMARY .......................................................................................................... XI

ABBREVIATIONS .................................................................................................................. XIII

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

2 MODELS OF LUNGMEN ABWR.......................................................................................... 3

2.1 Lungmen TRACE Model.................................................................................................. 3

2.2 Lungmen PARCS Model ................................................................................................. 5

2.3 Lungmen TRACE/PARCS Coupling Model...................................................................... 6

2.4 Lungmen TRACE/PARCS/FRAPTRAN Model ................................................................ 7

3 INITIAL CONDITIONS AND RESULTS ................................................................................ 9

3.1 Assumptions and Initial Conditions.................................................................................. 9

3.2 TRACE/PARCS Calculation Results................................................................................ 9

3.2.1 Blowdown Conditions ............................................................................................. 9

3.2.2 Pressure and Temperature Responses of Containment .......................................10

3.3 FRAPTRAN Calculation Results....................................................................................16

4 CONCLUSIONS..................................................................................................................21

5 REFERENCES....................................................................................................................23

ix

LIST OF FIGURES

Figure 1 Flowchart of combining TRACE/PARCS and FRAPTRAN codes ............................ 1-2

Figure 2 Lungmen TRACE model ......................................................................................... 2-4

Figure 3 Core pattern for Lungmen PARCS model................................................................ 2-5

Figure 4 The procedure of TRACE/PARCS coupling calculation [16] .................................... 2-6

Figure 5 Schematic of fuel rod geometry in FRAPTRAN....................................................... 2-8

Figure 6 Blowdown condition of RPV side........................................................................... 3-10

Figure 7 Blowdown condition of main steam header side.................................................... 3-10

Figure 8 Pressures of (a) UDW and (b) LDW ...................................................................... 3-12

Figure 9 Temperatures of (a) UDW and (b) LDW ................................................................ 3-13

Figure 10 Pressure of WW................................................................................................... 3-14

Figure 11 (a) Airspace and (b) SP Temperatures of WW ..................................................... 3-15

Figure 12 Pressure of RPV .................................................................................................. 3-16

Figure 13 Cladding outside temperatures calculated by TRACE/PARCS coupling model 

and FRAPTRAN model ....................................................................................... 3-17

Figure 14 Fuel surface hoop strain....................................................................................... 3-17

Figure 15 Cladding hoop strain ............................................................................................ 3-18

Figure 16 Fuel centerline temperature ................................................................................. 3-18

Figure 17 Cladding inside temperature ................................................................................ 3-19

xi

EXECUTIVE SUMMARY

An agreement in 2004 which includes the development and maintenance of TRACE has been signed 

between Taiwan and USA on CAMP. NTHU 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/PARCS model of Lungmen NPP is developed. 

According to the user manual, TRACE is the product of a long term effort to combine the capabilities of 

the NRC’s four main systems codes (TRAC-P, TRAC-B, RELAP5 and RAMONA) into one modernized 

computational tool. The 3-D geometry model of reactor vessel, which is one of the representative 

features of TRACE, can support a more accurate and detailed safety analysis of NPPs. On the whole 

TRACE provides greater simulation capability than the previous codes, especially for events like LOCA. 

PARCS is a multi-dimensional reactor core simulator which involves a 3-D calculation model for the 

realistic representation of the physical reactor while 1-D modeling features are also available. PARCS 

is capable of coupling the thermal-hydraulics system codes such as TRACE directly, which provide the 

temperature and flow field data for PARCS during the calculations.

Lungmen NPP is the fourth NPP in Taiwan. It has two identical units of ABWRs with 3,926 MWt rated 

thermal power each, consisted of 872 GE14 assemblies with 205 control rods. The steam flow is 

7.64×106Kg/h at rated power condition. The designed rated core flow is 52.2×106 Kg/h. Compared 

with BWRs, ABWR replaced the recirculation loop by 10 RIPs (reactor internal pumps), eliminating the 

probability of large break LOCA. 10 RIPs could provide 111% rated core flow at the nominal operating 

speed of 151.84 rad/sec.

The object of this paper is to develop a complete flow chart for analyzing the nuclear system transient, 

such as behaviors of fuel rod, vessel, and containment.

The double-ended MSLB transient in Lungmen ABWR was chosen to be a subject of case study in 

this paper. The MSLB is the design-basis accident analysis of containment, presenting in FSAR 

section 6.2 [1]. According to FSAR 6.2, double-ended MSLB transient is the limiting case for DW 

pressure. Lungmen NPP, the fourth NPP in Taiwan, has two identical units of ABWRs with 3,926 

MWt each, consisted of 872 GE14 assemblies (10×10 with two water rods) with 205 control rods. 

Compared with BWR containment, there are two main differences: a) drywell (DW) is divided into 

upper-drywell (UDW) and low-drywell (LDW), which are connected by 10 drywell-connecting-vents 

(DCVs); b) wetwell (WW) is isolated from reactor building, which is connected with DW by 10 vertical 

vents with 3 horizontal vents each.

The codes, TRACE, PARCS, and FRAPTRAN are all developed and provided by US NRC. The 

Lungmen TRACE/PARCS coupling model with only nuclear steam supply system (NSSS) had been 

established and verified that it has respectable accuracy shown in previous papers of our laboratory 

[2][3][4]. In order to develop a complete flow chart for analyzing the nuclear system transient, the 

Lungmen containment model and FRAPTRAN model were established in this research. The results of 

TRACE/PARCS coupling calculation, with containment model, were compared with those of both FSAR 

and GOTHIC [1][5], indicating that the TRACE/PARCS coupling model has the ability to predict the 

MSLB transient, and both RPV(Reactor Pressure Vessel) integrity and containment integrity criteria are 

met. After that, the output data from TRACE/PARCS calculation was putted into FRAPTRAN code as 

boundary conditions to analyze the thermo-mechanical behavior and calculate the stress, strain, oxide 

thickness, etc. The values of these factors were compared with the criteria. And the final results show 

that the fuel rod integrity criteria are met.

xiii

ABBREVIATIONS

CAMP Code Applications and Maintenance Program

DCVs Drywell-Connecting-Vents

DW Drywell

LDW Low-Drywell

LOCA Loss Of Coolant Accidents

MSLB Main SteamLine Break

NPP Nuclear Power Plant

NSSS Nuclear Steam Supply System

NTHU National Tsing Hua University

RIP Reactor Internal Pump

RPV Reactor Pressure Vessel 

SP Suppression Pool

SRV Safety Relief Valve

TCV Turbine Control Valve

TBV Turbine Bypass Valve

UDW Upper-Drywell

WW Wetwell

1 

1 INTRODUCTION

The object of this paper is to develop a complete flow chart for analyzing the nuclear system 

transient, such as behaviors of fuel rod, vessel, and containment, as shown in Figure 1.

The double-ended MSLB transient in Lungmen ABWR was chosen to be a subject of case study 

in this paper. The MSLB is the design-basis accident analysis of containment, presenting in FSAR 

section 6.2 [1]. According to FSAR 6.2, double-ended MSLB transient is the limiting case for DW 

pressure. Lungmen NPP, the fourth NPP in Taiwan, has two identical units of ABWRs with 3,926 

MWt each, consisted of 872 GE14 assemblies (10×10 with two water rods) with 205 control rods. 

Compared with BWR containment, there are two main differences: a) DW is divided into UDW

and LDW, which are connected by 10 DCVs; b) WW is isolated from reactor building, which is 

connected with DW by 10 vertical vents with 3 horizontal vents each.

The codes, TRACE, PARCS, and FRAPTRAN are all developed and provided by US NRC. The 

Lungmen TRACE/PARCS coupling model with only NSSS had been established and verified that 

it has respectable accuracy shown in previous papers of our laboratory [2][3][4]. In order to 

develop a complete flow chart for analyzing the nuclear system transient, the Lungmen 

containment model and FRAPTRAN model were established in this research. The results of 

TRACE/PARCS coupling calculation, with containment model, were compared with those of both 

FSAR and GOTHIC [1][5], indicating that the TRACE/PARCS coupling model has the ability to 

predict the MSLB transient, and both RPV integrity and containment integrity criteria are met. 

After that, the output data from TRACE/PARCS calculation was putted into FRAPTRAN code as 

boundary conditions to analyze the thermo-mechanical behavior and calculate the stress, strain, 

oxide thickness, etc. The values of these factors were compared with the criteria. And the final 

results show that the fuel rod integrity criteria are met.
2 

Figure 1 Flowchart of combining TRACE/PARCS and FRAPTRAN codes
3 

2 MODELS OF LUNGMEN ABWR 

2.1 Lungmen TRACE Model

The preliminary Lungmen TRACE model is established based on the relevant documents, as shown 

in Figure 2 [6]~[9]. There are three major control systems implemented in Lungmen TRACE model: 

feedwater control system, pressure control system, and RIP control system. The core region was 

modeled by 22 thermal-hydraulic channels to simulate the T-H behavior of 872 fuel assemblies. In the 

region around the dropped rod, each channel represented a single assembly in order to reflect 

accurately the T-H reactivity feedback effects following a control rod drop. In other region, each 

channel represented several fuel assemblies. The number of axial nodes in each channel is 11. 

According to the assemblies in the real reactor, the vessel was divided into eleven axial levels, four 

radial rings, and six azimuthal sectors. The six azimuthal sectors are orientated in 36°, 36°, 108°, 36°, 

36°, 108°, 36°apart, and each azimuthal sector is connected with the feed water line inlet (six 

feedwater lines). There are four main steam lines connected to the 36°azimuthal sector of vessel and 

ten RIPs connected to six azimuthal sectors, one for every 36°. The ten RIPs were separated into 

three groups, four RIPs not connect to M/G sets (RIP3) and six RIPs connect to M/G sets (RIP1 and 

RIP2, thee for each). There are four sets of valves included in this model. The MSIVs and Turbine 

control valves (TCVs) are normally opened. The turbine bypass valve (TBV) and six groups of safety 

relief valves (SRVs), simulating eighteen SRVs distributed at the four main steam lines with different 

setpoints, are normally closed. In addition, the Moody choke flow model was adopted for limiting the 

maximum SRVs’ flow.

In addition, the steady state plant parameters from Lungmen TRACE model had been 

successfully verified with those from FSAR and RETRAN02. The verified results reveal that there 

is respectable accuracy in the Lungmen TRACE model [10][11].
4 

Figure 2 Lungmen TRACE model
5 

2.2 Lungmen PARCS Model

PARCS involves 3D reactor core simulator for the realistic representation of physical reactor, and 

it can solve steady-state and time-dependent, multi-group neutron diffusion and SP3 transport 

equations in orthogonal and hexagonal core geometries. Figure 3 shows the core pattern for 

Lungmen PARCS model. For radial mesh, there are 1012 nodes in Lungmen PARCS model: 872 

nodes model 872 fuel assemblies (yellow square); 140 nodes model the reflector outside the core 

(blue square). And the number of axial planes is 25 in the effective fuel region. The cross-section 

data used in PARCS calculation is provided by PMAXS file which is generated by GenPMAXS 

program from the macroscopic cross-section libraries and the results of lattice code, CASMO [12].

The preliminary Lungmen PARCS model is established by our laboratory colleagues, Chen [13] 

and Chang [14]. The kinf calculated from PARCS had been verified by that from SIMULATE. The 

result shows the respectable accuracy in Lungmen PARCS model that the error bar is smaller 

than 10-5. 

Figure 3 is the code pattern of Lungmen PARCS model. The marked positions, (11,9) and (11,28), 

are the fuel assemblies which were chosen for FRAPTRAN analyses.

Figure 3 Core pattern for Lungmen PARCS model
6 

2.3 Lungmen TRACE/PARCS Coupling Model

Figure 4 displays the flowchart of TRACE/PARCS coupling model. During the transient calculation, 

PARCS determines the core power distribution by using T-H conditions provided by TRACE. The 

power information is then transferred back to TRACE to calculate the new T-H conditions for 

PARCS. Thus the TRACE/PARCS coupling model gives the actual core power and T-H distribution 

at any time point.

Based on this preliminary Lungmen TRACE/PARCS coupling model, Feng et al.[15] analyzed the 

loss feed water heater transient and compared the results with plant vender data. It shows that the 

Lungmen TRACE/PARCS coupling model has an ability of transient simulation of Lungmen NPP.

Figure 4 The procedure of TRACE/PARCS coupling calculation [16] PARCS TRACE Lattice

CASMO-4 GenPMAXS PMAXS

Thermal-hydraulic 

Calculation Code

Node-wise 

Cross Section

Node-wise 

Power

History for 

each Region

Cross Section 

Module

Neutronics 

Calculation

Depletion 

Module

Temperature and Flow 

Information
7 

2.4 Lungmen TRACE/PARCS/FRAPTRAN Model

FRAPTRAN is a computer code for analyzing the thermo-mechanical behavior of light water 

reactor fuel rod under transients and accidents, such as LOCAs and RIAs [17]. Figure 5illustrates 

the schematic of fuel rod in FRAPTRAN model. The axial fuel length from bottom to top was 

divided into 12 nodes, and the fuel radial direction was divided into 17 nodes, including 15 nodes 

in the pellet and 2 nodes in the cladding. Although different numbers of axial node were used in 

these codes, important physical parameters could be obtained by simple linear interpolation.

Figure 1 shows the flowchart of combining FRAPTRAN and TRACE/PARCS. The input file of 

FRAPTRAN mainly composes of three parts to define the transient problems: a) Fuel rod 

geometry; b) Power history, including axial pin power shape and pin power history; c) Coolant 

boundary conditions, including coolant temperature, coolant pressure, and cladding-coolant heat 

transfer coefficient. In FRAPTRAN code, there are two modes we can choose to input the coolant 

boundary condition: COOLANT mode and HEAT mode. In this report, HEAT mode was chosen

because the coolant boundary condition can be defined certainly from TRACE/PARCS output 

data. In addition, the reference temperature used in the calculation of fuel and clad enthalpy was 

defined at 298.15K.

The mechanical model used in FRAPTRAN for calculating the mechanical response of the fuel 

and cladding is the FRACAS-I model. This model does not account for stress-induced deformation 

of the fuel and therefore is called the rigid pellet model. This model includes the effects of thermal 

expansion of the fuel pellet; rod internal gas pressure; and thermal expansion, plasticity, and hightemperature creep of the cladding. After the cladding strain has been calculated by the 

mechanical model, the strain is compared with the value of an instability strain obtained from 

MATPRO. If the cladding effective plastic strain is greater than the cladding instability strain, then 

the cladding cannot maintain a cylindrical shape and local ballooning occurs. And the ballooning 

model, BALON2, is used to calculate the localized, nonuniform straining of the cladding. For the 

local region at which instability is predicted, a large deformation ballooning analysis is performed. 

No further strain is calculated for non-ballooning nodes. Modification of local heat transfer 

coefficients is calculated as the cladding ballooning progresses and additional surface area is 

presented to the coolant. 
8 

Figure 5 Schematic of fuel rod geometry in FRAPTRAN
9 

3 INITIAL CONDITIONS AND RESULTS

3.1 Assumptions and Initial Conditions

The assumptions and initial conditions of the analysis are as follows:

 Initial reactor power was 4005 MWt (102% rated power).

 Double-ended MSLB break occurred at 0sec. The broken area of the RPV side was 

0.0984mm2 (flow limiter area). And the broken area of the main steam header side was 

0.319mm2 (main steam line area).

 MSIVs started to close at 0.5sec and fully closed at 5.0sec after MSLB.

 Initial pressure and temperature of DW were 5.17kPaG and 57.2°C, respectively.

 Initial pressure and temperature of WW were 5.17kPaG and 35°C, respectively.

 The initial suppression pool (SP) level was at 7.1m from the SP bottom.

3.2 TRACE/PARCS Calculation Results

3.2.1 Blowdown Conditions

Figure 6 and Figure 7 show the blowdown conditions at both RPV side and main steam header 

side. The blowdown conditions of GOTHIC code at RPV side are generated from two different 

ways: a) obtained by RELAP5 transient analysis (GOTHIC_1); b) calculated by a simplified RPV 

in GOTHIC (GOTHIC_2). Note that, in FSAR analysis (not shown), the RPV side and main steam 

header side are lumped as one single break (a time-varied broken area) on RPV side. The 

TRACE/PARCS results show the same trends with case GOTHIC_1, but the case GOTHIC_2 

displays extremely different behaviors at RPV side. That is because the assumption of GOTHIC_2 

is according to FSAR: because of RPV pressure drop, the core water level would swell and reach 

the elevation of main steam line at 2sec (RPV swell time) after MSLB. In other words, before 2sec, 

RPV side provides the single-phase flow only. After 2 sec, a lot of liquid water would blow down 

into DW from RPV via main steam line.
10

Figure 6 Blowdown condition of RPV side

Figure 7 Blowdown condition of main steam header side

3.2.2 Pressure and Temperature Responses of Containment

Figure 8 and Figure 9 show the pressure and temperature responses of UDW and LDW. The 

TRACE/PARCS results show the same trends with case GOTHIC_1, but both pressure and 

temperature transfer delay-times are slightly longer than GOTHIC_1. That is because both FSAR 

and GOTHIC analyses, for conservative assumption, assume the DW volume to be the sum of 

UDW and 50%LDW. Thus, the transmissions of pressure and temperature in both FSAR and 
11

GOTHIC are faster than TRACE/PARCS. In addition, because FSAR and GOTHIC_2 make the 

same assumption of RPV swell time (2sec), as mentioned in 4.1.1, both pressure and temperature 

of DW drop obviously after a large amount of liquid water blow down into DW. Moreover, in FSAR 

analysis, the results of UDW and LDW are the same because FSAR treats UDW and LDW as 

one volume.

Figure 10 and Figure 11 show the pressure and temperature responses of WW. The 

TRACE/PARCS results show the same trends with both FSAER and GOTHIC except the WW 

airspace temperature, because FSAR assumes WW to be homogeneous mixture and steam to 

be completely condensed by SP.

According to TRACE/PARCS calculation, the peak of RPV dome pressure is 7.03MpaG (Figure 12, 

10.342MPaG for criteria); the peaking values of pressure and temperature in DW are 192.44kPaG 

and 158.82°C(309.9kPaG and 171.1°C for criteria, respectively); the peaking values of WW 

pressure, WW airspace temperature, and SP temperature are about 100kPaG, 80°C and 

38°C(309.9kPaG, 97.2°C and 124.0°C for criteria, respectively). And the peak of DW-WW pressure 

difference is 130.561kPaD(+172.6kPaD for criteria). 
12

(a)

(b)

Figure 8 Pressures of (a) UDW and (b) LDW
13

(a)

(b)

Figure 9 Temperatures of (a) UDW and (b) LDW
14

Figure 10 Pressure of WW
15

(a)

(b)

Figure 11 (a) Airspace and (b) SP Temperatures of WW
16

Figure 12 Pressure of RPV

3.3 FRAPTRAN Calculation Results

Before FRAPTRAN analysis, the cladding outside temperature calculated by FRAPTRAN must 

be compared with that calculated by TRACE/PARCS to re-confirm the correctness of input data, 

as shown in Figure 13. Note that, in FRAPTRAN analysis, MSLB was started at 200sec. Thus, 

the transient started time of FRAPTRAN, x-axis, was shifted to 0sec for comparison with 

TRACE/PARCS data.

Figure 14 and Figure 15 show the hoop strains of fuel surface and cladding. The main factor 

influencing the fuel surface hoop strain is reactor power. As Figure 12 shows, the fuel surface 

hoop strain decreases (i.e., fuel pellet contracts) after reactor power scrammed. The cladding 

hoop strain was calculated based on the following equation:

εθ = �

1

E (σθ − υσz)� + �εθ

P + dεθP� + �� α dT

T

T0

�

where�

1

E (σθ − υσz)� is due to the pressure difference between cladding inside and outside 

surface; �εθ

P + dεθP� is plastic term; �∫ α dT T

T0 � is due to thermal expansion. The FRAPTRAN 

calculation indicates that the plastic term is zero. That is, there is no non-reversible change during 

MSLB transient. The term �

1

E (σθ − υσz)� increases as RPV pressure drops after MSLB. Contrarily, 

the term �∫ α dT T

T0 � decreases after reactor power scrammed. The overall cladding hoop strain 

increases (i.e., cladding expands) with term �

1

E (σθ − υσz)� due to RPV pressure drop except the 

duration of control rod inserted. From 200.5sec to control rod fully inserted, cladding hoop strain 

decreases (i.e., cladding contracts) with term �∫ α dT T

T0 � due to reactor power scram.
17

Figure 16 and Figure 17 show the temperatures of fuel surface and cladding inside surface, both 

indicating that the temperatures decrease as reactor power decreases. The peak temperatures 

of fuel surface and cladding inside surface are 1390.10°C and 609.53°C(2805.0°C and 1200.0°C

for criteria, respectively), respectively.

Figure 13 Cladding outside temperatures calculated by TRACE/PARCS coupling model

and FRAPTRAN model

Figure 14 Fuel surface hoop strain
18

Figure 15 Cladding hoop strain

Figure 16 Fuel centerline temperature
19

Figure 17 Cladding inside temperature

21

4 CONCLUSIONS

A complete flow chart for analyzing the nuclear system transient was performed. And the results 

of TRACE/PARCS coupling calculation were compared with those of both FSAR and GOTHIC, 

indicating that the TRACE/PARCS coupling model has the ability to predict the MSLB transient. 

According to TRACE/PARCS calculation, the peak of RPV dome pressure is 7.03MPaG 

(10.342MPaG for criteria); the peaking values of pressure and temperature in DW are 

192.44kPaG and 158.82°C (309.9kPaG and 171.1°C for criteria, respectively); the peaking values 

of WW pressure, WW airspace temperature, and SP temperature are about 100kPaG, 80°C and 

38°C (309.9kPaG, 97.2°C and 124.0°C for criteria, respectively). And the peak DW-WW pressure 

difference is 130.561kPaD (+172.6kPaD for criteria). Both RPV integrity and containment integrity

criteria are met. According to FRAPTRAN calculation, the peak temperatures of fuel surface and 

cladding inside surface are 1390.10°C and 609.53°C(2805.0°C and 1200.0°C for criteria, 

respectively), respectively. The oxidation under this temperature is insignificant. Therefore, the 

fuel integrity criteria are met.

23

5 REFERENCES

[1] Final Safety Analysis Report:LUNGMEN NUCLEAR POWER STATION UNITS 1 & 2, 

Taiwan Power Company, 2007.

[2] J.R. Wang, H.T. Lin, W.C. Wang, S.M. Yang, and C. Shih, “TRACE models and 

verifications for LUNGMEN ABWR”, American Nuclear Society Winter Meeting, November 

15-19, 2009.

[3] C.Y. Chang, C. K. Shih, and J. R. Wang, “The Establishment and Applications of Lungmen 

TRACE/PARCS Models”, National Tsing-Hua University, 2012.

[4] A.L. Ho, J. R. Wang, H. T. Lin and C. K. Shih, "TRACE/PARCS Analysis of Full Isolation 

Startup Test for LUNGMEN ABWR," in ICONE 20th, ICONE20POWER2012-54174, 2012.

[5] Y.S. Chen, Y.R. Yuann, and L.C. Dai, “Lungmen ABWR containment analyses during shotterm main steam line break LOCA using GOTHIC”, Nuclear Engineering and Design, 

p.106-115, 2012.

[6] Taiwan Power Company, “Lungmen Nuclear Power Station Startup Test Procedure- One 

RIP Trip Test”, STP-28A-HP (2008).

[7] Taiwan Power Company, “Lungmen Nuclear Power Station Startup Test Procedure-Three 

RIPs Trip Test”, STP-28B-HP (2008).

[8] Taiwan Power Company, “Lungmen Nuclear Power Station Startup Test ProcedureReactor Full Isolation”, STP-32-HP (2008).

[9] U. S. Nuclear Regulatory Commission, TRACE v5.0 USER'S MANUAL, 2012. 

[10] J.R. Wang, H.T. Lin, W.C. Wang, S.M. Yang, and C. Shih, “TRACE models and 

verifications for LUNGMEN ABWR”, American Nuclear Society Winter Meeting, November 

15-19, 2009.

[11] J. R. Wang and H. T. Lin, “TRACE Analysis of MSIV Closure Direct Scram Event for 

Lungmen ABWR”, in ICAPP 10, San Diego, CA, USA, 2010.

[12] Y. Xu and T. Downar, “GenPMAXS Code for Generating the PARCS Cross Section 

Interface File PMAXS”, University of Michigan, April, National Tsing-Hua University, 2009.

[13] S.J. Chen, “Study and Application of Neutronic Model in TRACE code”, National Tsing-Hua 

University, 2010.

[14] C.Y. Chang, “The Establishment and Applications of Lungmen TRACE/PARCS Models”, 

National Tsing-Hua University, 2012.

[15] T.S. Feng, J.R. Wang, H.T. Lin, and C. Shih, “Analysis Of Feedwater Heater Transients 

For LUNGMEN ABWR BY TRACE/PARCS”, ICONE 20th, 2012. 

[16] T. Downa., Y. Xu., V. Seke. and N. Hudson, PARCS v3.0 U.S. NRC Core Neutronic 

Simulator USER MANUAL, University of Michigan, 2012. 

[17] K.J. Geelhood, W.G. Luscher, C.E. Beyer, and J.M. Cuta, FRAPTRAN 1.4: A Computer 

Code for the Transient Analysis of Oxide Fuel Rods, NUREG/CR-7023, Vol. 1, 2011.

Main Steam Line Break Analysis for Lungmen ABWR

NUREG/IA-0476

Chunkuan Shih*, Jong-Rong Wang, Ai-Ling Ho, Shao-Wen Chen

Show-Chyuan Chiang*, Tzu-Yao Yu*

Institute of Nuclear Engineering and Science, 

National Tsing Hua Univ.

Nuclear and New Energy Education and Research Foundation 

101 Section 2, Kuang Fu Rd., HsinChu,

K. Tien, NRC Project Manager

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

The object of this paper is to develop methodologies for analyzing the behaviors of fuel rod, vessel, and containment 

during main steamline break (MSLB) transient. The broken area of the RPV side was assumed to be 0.0984m2 (flow 

limiter). And the broken area of the main steam header side was assumed to 0.319m2 (main steam line area). 

According to FSAR, for conservative assumption, MSIVs started to close at 0.5sec and fully closed at 5.0sec after 

the transient started. The results of TRACE/PARCS coupling calculation were compared with those of both FSAR 

and GOTHIC data, indicating that the TRACE/PARCS coupling model has the ability to predict the MSLB transient, 

and both RPV integrity and containment integrity criteria are met. After that, the output data from TRACE/PARCS 

calculation was put into FRAPTRAN code as boundary conditions to analyze the thermo-mechanical behavior and 

calculate the stress, strain, oxide thickness, etc. The values of these factors were compared with the criteria. And the 

final results show that the fuel rod integrity criteria are met.

November 2016

Technical

Advanced Boiling Water Reactor (ABWR)

TRACE/PARCS/FRAPTRAN

Main Steam Line Break (MSLB)

Main Steam Isolation Valve (MSIV)

Lungmen Nuclear Power Plant (NPP)

*Department of Nuclear Safety, Taiwan Power Co.

242, Section 3, Roosevelt Rd., Zhongzheng Dist.

Taipei, Taiwan


NUREG/IA-0476 Main Steam Line Break Analysis for Lungmen ABWR November 2016