NUREG/IA-0429 

International 

 Agreement Report

Analysis of Loss of Feedwater Heater Transients

for Lungmen ABWR by TRACE/PARCS 

Prepared by:

Jong-Rong Wang, Tsung-Sheng Feng*, Hao-Tzu Lin, 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

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: March 2013

Date Published: September 2013

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission
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infringe privately owned rights. 
NUREG/IA-0429 

International 

 Agreement Report

Analysis of Loss of Feedwater Heater Transients

for Lungmen ABWR by TRACE/PARCS 

Prepared by:

Jong-Rong Wang, Tsung-Sheng Feng*, Hao-Tzu Lin, 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

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: March 2013

Date Published: September 2013

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission

iii

ABSTRACT

The TRACE/PARCS model of Lungmen ABWR was used to evaluate the loss of feedwater 

heater (LOFH) transient of the Lungmen startup tests. The Loss of Feedwater heater transient is 

an anticipated operational occurrence (AOO) event. Identifications of the responses of the 

Lungmen models and verification of the plant vendor’s analysis results are crucial in the plant 

licensing analysis. PARCS, a three-dimensional neutronics simulator, is capable of performing 

detail transient three-dimensional core power distributions and can be coupled with TRACE for 

thermal hydraulic feedback analysis. The feedwater enthalpy entering the RPV in this event is 

modeled as a 30 seconds time constant decay curve. When feedwater temperature drops 

approximately 37°C, the feedwater control system (FWCS) triggers reactor internal pump (RIP) 

runback and selected control rod run in (SCRRI) immediately. The colder feedwater temperature 

collapses the voids, which leads to the void reactivity increase and decreases RPV water level. 

SCRRI can reduce the core reactivity and core temperature which then increase the Doppler 

reactivity. The water level would fluctuate between L3 and L8, having enough safety margins to 

avoid either low or high water level scram setpoints. On the other hand, we have also simulated 

another case without RIP runback and SCRRI. The sensitivity studies of this transient include 

different time constants, SCRRI delay times, RIP runback rates and RIP runback delay times. 

Furthermore, the study of 18 CHANs model and 206 CHANs model performance with 

TRACE/PARCS has been evaluated. The SNAP animation model can show three dimensional 

visualized results of different core parameters. 
iv

 
v

FOREWORD

The US NRC (United States Nuclear Regulatory Commission) is developing an advanced 

thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development 

of TRACE is based on TRAC, integrating RELAP5 and other programs. NRC has determined 

that in the future, TRACE will be the main code used in thermal hydraulic safety analysis, and no 

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

continued. A graphic user interface program, SNAP (Symbolic Nuclear Analysis Program) which 

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

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

accurate and detailed safety analysis of nuclear power plants. TRACE has a greater simulation 

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

Taiwan and the United States have signed an agreement on CAMP (Code Applications and 

Maintenance Program) which includes the development and maintenance of TRACE. INER 

(Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) is the organization in 

Taiwan responsible for the application of TRACE in thermal hydraulic safety analysis, for 

recording user’s experiences of it, and providing suggestions for its development. To meet this 

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

TRACE/PARCS model of Lungmen NPP was used to evaluate the loss of feedwater heater 

(LOFH) transient of the Lungmen startup tests. 
vi 
vii

CONTENTS

Page

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

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

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

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

TABLES .......................................................................................................................................... xi 

EXECUTIVE SUMMARY .............................................................................................................. xiii 

ABBREVIATIONS .......................................................................................................................... xv 

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

2. TRACE/PARCS COUPLING MODEL OF LUNGMEN ABWR ............................................. 2-1 

2.1 Lungmen TRACE Model ............................................................................................. 2-1 

2.2 Lungmen PARCS Model ............................................................................................. 2-3 

2.3 Lungmen TRACE/PARCS Coupling Model ............................................................... 2-4 

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

4. SENSITIVITY STUDIES ........................................................................................................ 4-1 

4.1 Different FW Heater Time Constant .......................................................................... 4-1 

4.2 SCRRI Delay Time ....................................................................................................... 4-1 

4.3 RIP Runback Delay Time ............................................................................................ 4-1 

4.4 RIP Runback Rate ....................................................................................................... 4-1 

5. CONCLUSIONS..................................................................................................................... 5-1 

6. REFERENCES....................................................................................................................... 6-1 
viii

 
ix

FIGURES

Page

1. Lungmen TRACE model ................................................................................................ 2-1 

2. Lungmen TRACE/PARCS 18 CHANs mapping model ............................................... 2-2 

3. Lungmen TRACE/PARCS 206 CHANs mapping model ............................................. 2-2 

4. Control rod pattern for Lungmen PARCS model ........................................................ 2-3 

5. History of control rod movements from SCRRI .......................................................... 2-3 

6. The procedure of TRACE/PARCS coupling calculation ............................................ 2-4 

7. The core parameters of the final state in LOFH.......................................................... 3-1 

8. Power responses of the prediction analysis in LOFH ............................................... 3-2 

9. Reactivity responses of the prediction analysis in LOFH ......................................... 3-2 

10. The initial state and final state under the SNAP animation ....................................... 3-3 

11. Power responses of the prediction analysis in LOFH ............................................... 3-4 

12. Dome pressure responses of the prediction analysis in LOFH ............................... 3-4 

13. Core flow responses of the prediction analysis in LOFH ......................................... 3-5 

14. Water level responses of the sensitivity studies for different time constant ......... 4-2 

15. Reactivity responses of the sensitivity studies for SCRRI delay time ..................... 4-2 

16. Power responses of the sensitivity studies for SCRRI delay time ........................... 4-3 

17. Reactivity responses of the sensitivity studies for RR delay time ........................... 4-3 

18. Power responses of the sensitivity studies for RR delay time ................................. 4-4 

19. Reactivity responses of the sensitivity studies for RIP runback rate ...................... 4-4 

20. Power responses of the sensitivity studies for RIP runback rate ............................ 4-5 
x

 
xi

TABLES

Page

1. Summary of TRACE/PARCS steady state condition .................................................. 2-4 

2. Summary of TRACE/PARCS analysis for different time constant studies in 

LOFH ................................................................................................................................ 4-6 

3. Summary of TRACE/PARCS analysis for SCRRI delay time studies in LOFH ........ 4-6 

4. Summary of TRACE/PARCS analysis for RR delay time studies in LOFH .............. 4-6 

5. Summary of TRACE/PARCS analysis for RIP runback rate studies in LOFH ......... 4-6 
xii 
xiii

EXECUTIVE SUMMARY

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

signed between Taiwan and USA on CAMP. INER is the organization in Taiwan responsible for 

applying TRACE to thermal hydraulic safety analysis in order to provide users’ experiences and 

development suggestions. To fulfill this responsibility, the TRACE/PARCS model of Lungmen 

NPP is developed by INER. 

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

capabilities of the NRC’s four main systems codes (TRAC-P, TRAC-B, RELAP5 and RAMONA) 

into one modernized computational tool. NRC has ensured that TRACE will be the main code 

used in thermal hydraulic safety analysis in the future without further development of other 

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

reactor vessel, which is one of the representative features of TRACE, can support a more 

accurate and detailed safety analysis of NPPs. 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 first ABWR plant in Taiwan and still under construction. It has two identical 

units with 3,926 MWt rated thermal power each and 52.2×106 kg/hr rated core flow. The core 

has 872 bundles of GE14 fuel, and the steam flow is 7.637×106

 kg/hr at rated power. There are 

10 RIPs in the reactor vessel, providing 111% rated core flow at the nominal operating speed of 

151.84 rad/sec. 

The TRACE/PARCS model of Lungmen NPP was used to evaluate the LOFH transient of the 

Lungmen startup tests. The Loss of Feedwater heater transient is an anticipated operational 

occurrence (AOO) event. Identifications of the responses of the Lungmen models and 

verification of the plant vendor’s analysis results are crucial in the plant licensing analysis. The 

feedwater enthalpy entering the RPV in this event is modeled as a 30 seconds time constant 

decay curve. When feedwater temperature drops approximately 37°C, the feedwater control 

system (FWCS) triggers reactor internal pump (RIP) runback and selected control rod run in 

(SCRRI) immediately. The colder feedwater temperature collapses the voids, which leads to the 

void reactivity increase and decreases RPV water level. SCRRI can reduce the core reactivity 

and core temperature which then increase the Doppler reactivity. The water level would fluctuate 

between L3 and L8, having enough safety margins to avoid either low or high water level scram 

setpoints. On the other hand, we have also simulated another case without RIP runback and 

SCRRI. The sensitivity studies of this transient include different time constants, SCRRI delay 

times, RIP runback rates and RIP runback delay times. According to the sensitivity studies, it 

could be concluded that every case maintains the water level between L3 and L8, so that the 

scram setpoints would not be triggered. The shorter SCRRI delay time has significant effect on 

fuel rod temperature which leads to higher Doppler and void reactivity. The studies of different 

RIP runback delay times and rates affect the void feedback principally during the runback stage. 

The above sensitivity studies have no significant impacts on the final power and reactivity. 

Furthermore, the study of 18 CHANs model and 206 CHANs model performance with 

TRACE/PARCS has been evaluated. Comparing results between 18 CHANs and 206 CHANs 
xiv

model, 18 CHANs model has ability to perform plant response to the transient and 206 CHANs 

model is proper to local power calculation. The SNAP animation model can show three 

dimensional visualized results of different core parameters. Besides, the startup tests of 

Lungmen NPP will be performed in 2014 and the measured data of Lungmen NPP will be used 

to estimate and modify the TRACE/PARCS model of Lungmen NPP in the future. 
xv

ABBREVIATIONS

ABWR Advanced Boiling Water Reactor 

AOO Anticipated Operational Occurrence 

APLS Advanced Lungmen Plant Specific Simulation 

CAMP Code Applications and Maintenance Program 

FSAR Final Safety Analysis Report 

FWCS Feedwater Control System 

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

LOFH Loss of Feedwater Heater 

NRC Nuclear Regulatory Commission 

PARCS Purdue Advanced Reactor Core Simulator 

RFCS Recirculation Flow Control System 

RIP Reactor Internal Pump 

RR RIP Runback 

SCRRI Selected Control Rod Run In 

SNAP Symbolic Nuclear Analysis Package 

TPC Taiwan Power Company 

TRACE TRAC/RELAP Advanced Computational Engine 
xvi

 
1-1 

1. INTRODUCTION

Lungmen NPP, the first advanced boiling water reactor (ABWR) owned by Taiwan power 

company (TPC) in Taiwan, has two identical units with 3926 MWt rated power and 52.2×106

 kg/h 

rated core flow each, and the reactor core is comprised of 872 GE-14 fuel assemblies with 205 

control rods. TRACE (TRAC/RELAP Advanced Computational Engine) code developed by 

USNRC is an advanced and best-estimate reactor systems code for analysing thermal hydraulic 

behaviour in nuclear power plant. PARCS (Purdue Advanced Reactor Core Simulator) is a 

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

representation. The coupling between the TRACE and the PARCS with 3-D neutronics models 

of the reactor core into system transient has been developed successfully. SNAP (Symbolic 

Nuclear Analysis Package) provides not only a graphical interface for user to develop the 

TRACE/PARCS model but also animation models to visualize 3-D results of different core 

parameters. 

During the Lungmen startup test, the acceptable plant parameters and responses have been 

provided by the plant vendor’s results. In the same time, Yang etc. verified the plant vendor’s 

results by using a dual RELAP5 advanced Lungmen plant specific simulation platform (APLS) 

[1]. Loss of feedwater heater (LOFH) transient is an anticipated operational occurrence (AOO) 

event. There are two reasons to LOFH: First, feedwater heater bypass which causes step 

change in decreasing temperature of downcomer feedwater. Secondly, steam line of feedwater 

heater closure, due to the residual heat in feedwater heater can still heat the feedwater, the 

temperature of downcomer feedwater decreases gradually. 

In this study, TRACE/PARCS applies 30 seconds feedwater time constant decay curve to 

determine feedwater enthalpy entering the RPV. Upon feedwater temperature drops 

approximately 37°C, the feedwater control system (FWCS) will send a signal to recirculation flow 

control system (RFCS), which triggers reactor internal pump (RIP) runback and selected control 

rod run in (SCRRI) immediately. Besides, we compare our 18 CHANs model with 206 CHANs 

model. The compared results show good agreement with power in LOFH transient. The 

sensitivity studies of this transient include different time constants, SCRRI delay times, RIP 

runback rates and RIP runback delay times. And the analysis results of LOFH have been 

compared with RETRAN-3D data. 
1-2 

 
2-1 

2. TRACE/PARCS COUPLING MODEL OF LUNGMEN ABWR 

The study of LOFH startup test is modeled at 100% power and 100% flow by using TRACE 

V5.0p3 and PARCS V3.0 under SNAP V2.2.0. 

2.1 Lungmen TRACE Model

The TRACE model for Lungmen Nuclear Power Plant is shown in Figure 1. The RPV is modeled 

by TRACE 3-D VESSEL component with 11 axial levels, 4 radial rings, and 6 azimuthal sectors 

(separately in 36°, 36°, 108°, 36°, 36°, 108° apart ) for a total of 264 computational cells. The 

vessel is connected with 4 steam lines (36° azimuth al sectors separately) and 6 feedwater lines 

(6 azimuthal sectors of the VESSEL). Each of the 10 RIPs is connected to the VESSEL at every 

36° azimuthal sectors and modeled as PUMP component at the outer ring of the VESSEL. 

These 10 RIPs are arranged into three groups. 6 RIPs are connected to M-G sets; the other 4 

RIPs are connected to non M-G sets. 

The core is modeled by 18 CHAN components (3 inner radial rings of 6 azimuthal sectors 

individually) to represent 872 GE-14 10X10 fuel assemblies. Moreover, we discuss the 

difference between 206 CHANs model and 18 CHANs model as shown in Figure 2 and Figure 3. 

The TRACE model has been verified with the case data of FSAR chapter 15 [2][3]. We expect 

more improvement on the Lungmen TRACE model after the commercial operation of Lungmen 

NPP. 

Figure 1 Lungmen TRACE model 
2-2 

Figure 2 Lungmen TRACE/PARCS 18 CHANs mapping model 

Figure 3 Lungmen TRACE/PARCS 206 CHANs mapping model 
2-3 

2.2 Lungmen PARCS Model

The PARCS model which is based on beginning of cycle consists of 872 fuel assemblies and 

205 control rods. Each fuel assembly is represented by a single neutronics node. The active 

core height is 381cm, with 25 axial levels for active core, and top/bottom axial levels for 

15.24-cm-thick axial reflector regions. The 205 control rod are divided into 19 groups, each 

group has different initial step as shown in Figure 4. The function of SCRRI is simulated in 

PARCS model and the history of control rod movement is shown in Figure 5. The SCRRI would 

be fully inserted within 145 seconds. 

Figure 4 Control rod pattern for Lungmen PARCS model 

0 20 40 60 80 100 120 140 160 180 200

Time (sec)

0

20

40

60

80

100

120

140

160

180

200

Control rod position (steps)

Control rod movement from SCRRI

Bank 7

Bank 12

Bank 13 & 19

Bank 16

Bank 17

Figure 5 History of control rod movements from SCRRI [4] 
2-4 

2.3 Lungmen TRACE/PARCS Coupling Model

CASMO-4, a lattice physics code, generated the cross-section information which would be 

processed into the appropriate format by the program GenPMAXS. Then, the PMAXS cross 

section file that could be read by PARCS. The procedure of TRACE/PARCS coupling calculation 

is shown in Figure 6. During a transient, TRACE provides the thermal-hydraulic conditions for 

PARCS which would response the power distribution for TRACE. Table 1 lists the major 

thermal-hydraulic parameters at steady state condition for Lungmen ABWR. 

Figure 6 The procedure of TRACE/PARCS coupling calculation [5]

Table 1 Summary of TRACE/PARCS steady state condition 

Parameters FSAR TRACE/PARCS model 

Thermal power (MW) 3926 3926 

Steam flow rate (kg/s) 2122 2121.2 

Core flow rate (kg/s) 14500 14610 

Dome pressure (MPa) 7.2 7.16 

Narrow range water level (m) 13.42 13.43 
3-1 

3. RESULTS 

The purpose of LOFH startup test is to demonstrate acceptable plant response and associated 

assumptions used for this transient in the plant licensing analysis. The study of LOFH transient 

is under the condition of 100% power and 100% flow. Upon sensing the reduction of 37°C in 

feedwater temperature, the FWCS sends out two signals, which initiate RIP runback and SCRRI 

in order to reduce the core power. Under normal operation, the RIP runback rate is designed at 

5% of rated speed per second, and the delay time is 0.275 second; the delay time for SCRRI 

initiation is designed to be 0.09 second. Figure 7 shows that the core parameters of the final 

state of LOFH under SNAP animation. Moreover, the water level fluctuate between L3 and L8, 

having enough safety margins to avoid either low or high water level scram setpoints. 

When the LOFH occurs, the colder feedwater temperature collapses the voids, which leads to 

the void reactivity increase and decreases RPV water level and followed by the increasing core 

power in Figure 8. SCRRI and RIP runback can reduce the core reactivity and core temperature 

which increase the Doppler reactivity as shown in Figure 9. 

Figure 7 The core parameters of the final state in LOFH 
3-2 

0 50 100 150 200

Time (sec)

30

40

50

60

70

80

90

100

110

Normalized power (%)

LOFH Power

TRACE/PARCS 18-CHANs

TRACE/PARCS 206-CHANs

Figure 8 Power responses of the prediction analysis in LOFH 

0 50 100 150 200 250 300

Time (sec)

-2

-1.5

-1

-0.5

0

0.5

1

Reactivity ($)

LOFH Reactivity

TRACE/PARCS 18-CHANs

TRACE/PARCS 206-CHANs

Void reactivity

Doppler reactivity

Total reactivity

Control rod reactivity

Figure 9 Reactivity responses of the prediction analysis in LOFH 
3-3 

TRACE/PARCS 18 CHANs model shows that the power is similar with 206 CHANs model in 

Figure 8. Figure 9 illustrates that 206 CHANs model has higher void reactivity than 18 CHANs 

model .The corresponding SCRRI reactivity has higher control rod worth because lower void 

fraction leads to more thermal neutrons absorbed by control rod. Due to more channels in RPV, 

206 CHANs model can detail the core phenomena like maximum fuel rod temperature. However, 

it needs smaller timestep to calculate LOFH in 105 hrs. Therefore, 18 CHANs model is suitable 

for NPP system transient because it has the same trends with 206 CHANs model but less 

time-consuming (4 hrs). 

We also simulate the case without RIP runback and SCRRI to get a range of power and 

temperature for startup test. Figure 10 presents that this case has higher void fraction which can 

provide negative reactivity. Hence, the power can reach 110% without triggering reactor scram 

(120%). Besides, the TRACE/PARCS model results are compared well with RETRAN-3D as 

shown in Figure 11~ Figure 13. As a result of slight variance in dome pressure, the power 

difference between TRACE/PARCS and RETRAN-3D is two percent. 

Figure 10 The initial state and final state under the SNAP animation 
3-4 

0 200 400 600 800

Time (sec)

92

96

100

104

108

112

116

120

Normalized power (%)

LOFH Power

TRACE/PARCS 

RETRAN-3D

Figure 11 Power responses of the prediction analysis in LOFH 

0 200 400 600 800

Time (sec)

6

6.4

6.8

7.2

7.6

8

Dome Pressure (MPa)

LOFH Dome Pressure

TRACE/PARCS 

RETRAN-3D

Figure 12 Dome pressure responses of the prediction analysis in LOFH 
3-5 

0 200 400 600 800

Time (sec)

99

99.5

100

100.5

101

101.5

102

102.5

103

Core Flow (%)

LOFH Core Flow

TRACE/PARCS 

RETRAN-3D

Figure 13 Core flow responses of the prediction analysis in LOFH 
3-6 

 
4-1 

4. SENSITIVITY STUDIES 

The following sensitivity studies are simulated by TRACE/PARCS 18 CHANs model. It is 

believed that 18 CHANs model with less CPU time is more suitable for demonstrating the 

acceptable plant response to the LOFH. The sensitivity studies of this transient include different 

time constants, SCRRI delay times, RIP runback delay times and RIP runback rates. 

4.1 Different FW Heater Time Constant

The feedwater heater time constant for Lungmen NPP is not known, so we run a set of 

sensitivity cases to demonstrate the effect of different time constant. The summary of different 

time constant studies is shown in Table 2. With larger time constant, the enthalpy decay more 

slowly and then a temperature reduction of 37°C will be delayed. In the same time, two signals 

from the FWCS will also be delayed. As shown in Figure 14, the 90 second time constant has 

the highest water level at 1375 cm, however, still having safety margins to keep water level 

between L3 (1280 cm) and L8 (1395 cm). 

4.2 SCRRI Delay Time

Table 3 lists the analysis results for SCRRI delay times in LOFH. The curves of control rod 

reactivity resulted from SCRRI shift parallel with SCRRI delay times in Figure 15. It also shows 

that earlier initiating SCRRI will lead to lower fuel rod temperature which causes higher Doppler 

and void reactivity. The lowest power is, however, due to control rod reactivity. Moreover, 

different SCRRI delay times have the same state in the end of this study. 

4.3 RIP Runback Delay Time

Table 4 exhibits the analysis results for RIP runback (RR) delay times in LOFH. With longer RR 

delay times, more cold feedwater collapses the voids which leads to the void reactivity increase 

in Figure 17. Before the RIP runback, the power increase leads to Doppler reactivity decrease. 

However, different RR delay times do not affect the final state in this study. 

4.4 RIP Runback Rate

Table 5 summarizes the sensitivity study on RIP runback rates in LOFH. Figure 19 presents that 

the RIP runback mechanism lessens the core inlet flow, the void in core increase such that the 

void reactivity rises in proportional to the RIP runback rate. Besides, faster RIP runback rate 

reduces the core temperature with corresponding to Doppler reactivity increase. Different RIP 

runback rates have significant effect on power history where power drops faster as RIP runback 

rate increases. 
4-2 

0 100 200 300 400

Time (sec)

1320

1330

1340

1350

1360

1370

1380

Height (cm)

LOFH Water Level

Time constant = 15s

Time constant = 30s

Time constant = 60s

Time constant = 90s

Figure 14 Water level responses of the sensitivity studies for different time constant 

0 50 100 150 200 250 300

Time (sec)

-1.5

-1

-0.5

0

0.5

1

Reactivity ($)

LOFH Reactivity

SCRRI delay time = 0.09s

SCRRI delay time = 1.09s

SCRRI delay time = 10.09s

SCRRI delay time = 20.09s

SCRRI delay time = 30.09s

Void reactivity

Doppler reactivity

Total reactivity

Control rod reactivity

Figure 15 Reactivity responses of the sensitivity studies for SCRRI delay time 
4-3 

0 50 100 150 200 250 300 350 400

Time (sec)

30

40

50

60

70

80

90

100

110

Normalized power (%)

LOFH Power

SCRRI delay time = 0.09 s

SCRRI delay time = 1.09 s

SCRRI delay time = 10.09 s

SCRRI delay time = 20.09 s

SCRRI delay time = 30.09 s

Figure 16 Power responses of the sensitivity studies for SCRRI delay time 

0 25 50 75 100

Time (sec)

-1

-0.5

0

0.5

Reactivity ($)

LOFH Reactivity

RR Delay time = 0.275s

RR Delay time = 1.2s

RR Delay time = 10s

RR Delay time = 20s

RR Delay time = 30s

Void reactivity

Doppler reactivity

Figure 17 Reactivity responses of the sensitivity studies for RR delay time 
4-4 

0 50 100 150

Time (sec)

40

50

60

70

80

90

100

110

Normalized power (%)

LOFH Power

RR delay time = 0.275s

RR delay time = 1.2s

RR delay time = 10s

RR delay time = 20s

RR delay time = 30s

Figure 18 Power responses of the sensitivity studies for RR delay time 

0 50 100 150

Time (sec)

-1.5

-1

-0.5

0

0.5

1

Reactivity ($)

LOFH Reactivity

RIP runback rate = 4%

RIP runback rate = 5%

RIP runback rate = 6%

RIP runback rate = 10%

Void reactivity

Doppler reactivity

Total reactivity

Control rod reactivity

Figure 19 Reactivity responses of the sensitivity studies for RIP runback rate 
4-5 

0 25 50 75 100

Time (sec)

40

60

80

100

120

Normalized power (%)

LOFH Power

RIP runback rate = 4%

RIP runback rate = 5%

RIP runback rate = 6%

RIP runback rate = 10%

Figure 20 Power responses of the sensitivity studies for RIP runback rate 
4-6 

Table 2 Summary of TRACE/PARCS analysis for different time constant studies in LOFH 

Time constant (s) 15 30 60 90 

Max. power (%)/ time (s) 103/17 106/34 107/67 107/101 

Max. water level (cm) 1363.9 1369.1 1375.7 1375.8 

Table 3 Summary of TRACE/PARCS analysis for SCRRI delay time studies in LOFH 

SCRRI delay time (s) 0.09 1.09 10.09 20.09 30.09 

Max. power after RR (%) 67.8 67.9 69.5 70.9 72.0 

Max. water level (cm) 1369.0 1369.2 1370.1 1370.5 1370.6 

Table 4 Summary of TRACE/PARCS analysis for RR delay time studies in LOFH 

RR delay time (s) 0.275 1.2 10 20 30 

Min. void reactivity ($) -0.625 -0.618 -0.56 -0.471 -0.373 

Max. power after RR (%) 67.8 67.8 66.2 64.9 62.3 

Table 5 Summary of TRACE/PARCS analysis for RIP runback rate studies in LOFH 

RIP runback rate (%/s) 4 5 6 10 

Min. power within 100sec (%) 61.1 59.64 58.3 54.7 

Min. void reactivity ($) -0.579 -0.625 -0.667 -0.786 
5-1

5. CONCLUSIONS

The Lungmen TRACE/PARCS coupling model indicates that the response results of the plant is 

consistent with the plant vendor’s results in LOFH transient. Two main mechanisms, RIP 

runback and SCRRI, reduce the power without reactor scram successfully as feedwater 

temperature drops approximately 37°C. In summary, comparing results between 18 CHANs and 

206 CHANs model, 18 CHANs model has ability to perform plant response to the transient and 

206 CHANs model is proper to local power calculation. According to the sensitivity studies of 

different time constants, SCRRI delay times, RIP runback delay times and RIP runback rates, it 

could be concluded that every case maintains the water level between L3 and L8, so that the 

scram setpoints would not be triggered. The time constant is crucial for predicting water level. 

The 90 second time constant has the highest water level at 1375 cm. The shorter SCRRI delay 

time has significant effect on fuel rod temperature which leads to higher Doppler and void 

reactivity. The studies of different RIP runback delay times and rates affect the void feedback 

principally during the runback stage. The above sensitivity studies have no significant impacts on 

the final power and reactivity. The SNAP animation model demonstrates our three dimensional 

results visually of the plant parameters and phenomena. 
5-2

 
6-1 

6. REFERENCES

1. C.Y Yang, L.C. Dai, K.S. Liang, B.S. Pei, C.K. Shih, L.C. Wang, and S.C. Chiang, 2010, 

“ABWR Power Tests Simulation by Using a Dual RELAP5 Nuclear Power Plant Simulation 

Platform”, Nuclear Engineering and Design, In Press, Corrected Proof, Available online 17 

September 2011. 

2. Taiwan Power Company (TPC), “Final Safety Analysis Report for Lungmen Nuclear Power 

Station Units 1 and 2”, T.P.C, Taipei, Taiwan, 2007. 

3. 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. 

4. C.Y. Chang, H.T. Lin, J.R. Wang, and C. Shih, “Studies on SCRRI Performances in ABWR 

with TRACE/PARCS”, TopSafe 2012 Programme Committee, Helsinki, Finland, April 2012. 

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

Interface File PMAXS”, University of Michigan, April, 2009. 

6. Taiwan Power Company (TPC), “Lungmen Nuclear Power Station Startup Test ProcedureLoss of feedwater heating”, STP-26-HP, T.P.C, Taipei, Taiwan, 2008. 




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-0429 Analysis of Loss of Feedwater Heater Transients for 

Lungmen ABWR by TRACE/PARCS