NUREG/IA-0439 

International 

 Agreement Report

TRACE Analysis on Heat Removal Decrease 

Accidents for AP1000

 

Prepared by:

Jong-Rong Wang, Kai Wang*, 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: September 2013

Date Published: March 2014

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission
NRC Reference Material 

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

NRC’s Public Electronic Reading Room at 

http://www.nrc.gov/reading-rm.html. Publicly released 

records include, to name a few, NUREG-series 

publications; Federal Register notices; applicant, 

licensee, and vendor documents and correspondence; 

NRC correspondence and internal memoranda; bulletins 

and information notices; inspection and investigative 

reports; licensee event reports; and Commission papers 

and their attachments. 

NRC publications in the NUREG series, NRC 

regulations, and Title 10, “Energy,” in the Code of

Federal Regulations may also be purchased from one 

of these two sources. 

1. The Superintendent of Documents

 U.S. Government Printing Office

Mail Stop SSOP

Washington, DC 20402–0001

Internet: bookstore.gpo.gov

Telephone: 202-512-1800

Fax: 202-512-2250 

2. The National Technical Information Service 

Springfield, VA 22161–0002

www.ntis.gov

1–800–553–6847 or, locally, 703–605–6000 

A single copy of each NRC draft report for comment is 

available free, to the extent of supply, upon written 

request as follows:

Address: U.S. Nuclear Regulatory Commission

 Office of Administration 

Publications Branch

 Washington, DC 20555-0001

E-mail: DISTRIBUTION.RESOURCE@NRC.GOV

Facsimile: 301–415–2289 

Some publications in the NUREG series that are 

posted at NRC’s Web site address 

http://www.nrc.gov/reading-rm/doc-collections/nuregs 

are updated periodically and may differ from the last 

printed version. Although references to material found on 

a Web site bear the date the material was accessed, the 

material available on the date cited may subsequently be 

removed from the site.

Non-NRC Reference Material 

Documents available from public and special technical 

libraries include all open literature items, such as books, 

journal articles, transactions, Federal Register notices, 

Federal and State legislation, and congressional reports. 

Such documents as theses, dissertations, foreign reports 

and translations, and non-NRC conference proceedings 

may be purchased from their sponsoring organization. 

Copies of industry codes and standards used in a 

substantive manner in the NRC regulatory process are 

maintained at—

The NRC Technical Library

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852–2738 

These standards are available in the library for reference 

use by the public. Codes and standards are usually 

copyrighted and may be purchased from the originating 

organization or, if they are American National Standards, 

from—

American National Standards Institute

11 West 42nd Street

New York, NY 10036–8002

www.ansi.org

212–642–4900 

AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

Legally binding regulatory requirements are stated only 

in laws; NRC regulations; licenses, including technical 

specifications; or orders, not in NUREG-series 

publications. The views expressed in contractorprepared publications in this series are not necessarily 

those of the NRC.

The NUREG series comprises (1) technical and 

administrative reports and books prepared by the staff 

(NUREG-XXXX) or agency contractors

(NUREG/CR-XXXX), (2) proceedings of conferences 

(NUREG/CP-XXXX), (3) reports resulting from 

international agreements (NUREG/IA-XXXX), (4) 

brochures (NUREG/BR-XXXX), and (5) compilations of 

legal decisions and orders of the Commission and 

Atomic and Safety Licensing Boards and of Directors’

decisions under Section 2.206 of NRC’s regulations 

(NUREG-0750).

DISCLAIMER: This report was prepared under an 

international cooperative agreement for the exchange of 

technical information. Neither the U.S. Government nor 

any agency thereof, nor any employee, makes any 

warranty, expressed or implied, or assumes any legal 

liability or responsibility for any third party’s use, or the 

results of such use, of any information, apparatus, 

product or process disclosed in this publication, or 

represents that its use by such third party would not 

infringe privately owned rights. 
NUREG/IA-0439 

International 

 Agreement Report

TRACE Analysis on Heat Removal Decrease 

Accidents for AP1000

 

Prepared by:

Jong-Rong Wang, Kai Wang*, 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: September 2013

Date Published: March 2014

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

This research presents the applicability of TRACE to simulate AP1000's heat removal decrease 

accidents. The AP1000 nuclear power plant (NPP) TRACE model containing the essential 

components of the primary, secondary loop and passive safety systems with corresponding 

control systems is established through the interface code -SNAP based on the Westinghouse 

design. The steady-state calculation of TRACE is conducted to testify the accuracy of model 

and the results show a good coherent with the design parameters. Two condition II events 

categorized as the decrease in heat removal by secondary system are simulated and TRACE‘s 

results are consistent with Westinghouse's LOFTRAN results. The results of TRACE also reveal 

that the availability of reactor coolant pumps has a significant influence on the passive heat 

removal performance. Moreover, even without any AC power source, the passive core cooling 

system is capable of extracting all the core decay heat without the operator intervention. In 

conclusion, the passive safety system has a strong capability coping with the long-term heat 

removal decrease by secondary systems, further preventing the occurrence of severe 

consequences.
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 model of AP1000 NPP has been built. In this report, we focus on the 

TRACE analysis of the heat removal decrease accidents for AP1000. 
vi 

 
vii 

CONTENTS

Page

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

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

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

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

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

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

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

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

2. AP1000 TRACE MODEL ..................................................................................................... 2-1 

2.1 Model Description ..................................................................................................... 2-1 

2.2 Steady State Calculation .......................................................................................... 2-2 

3. SIMULATION OF TWO HEAT REMOVAL DECREASE ACCIDENTS ............................... 3-1 

3.1 Accident Assumptions and Description ................................................................. 3-1 

3.2 Loss of Normal Feedwater Flow ............................................................................ 3-1 

3.3 Loss of AC Power to Plant Auxiliaries .................................................................. 3-2 

3.4 Discussion ............................................................................................................... 3-2 

3.5 The Sensitivity Study on Influence of Different Decay Heat Generation ............ 3-3 

4. CONCLUSIONS................................................................................................................... 4-1 

5. REFERENCES..................................................................................................................... 5-1 
viii 

 
ix 

FIGURES

Page

Figure 1 The overview of AP1000 TRACE model ............................................................... 2-3 

Figure 2 The control systems in AP1000 model ................................................................. 2-4 

Figure 3 AP1000 SNAP animation model ............................................................................ 2-4 

Figure 4 Normalized RCS volumetric flow rate (loss of normal feedwater flow) ............ 3-4 

Figure 5 Normalized PRHR heat flux and CMT injection flow rate (loss of normal 

 feedwater flow) ....................................................................................................... 3-4 

Figure 6 RCS temperature in loop containing the PRHR (loss of normal feedwater 

 flow) ........................................................................................................................ 3-5 

Figure 7 Normalized PRHR and core power (loss of normal feedwater flow) ................. 3-5 

Figure 8 Normalized RCS volumetric flow rate (loss of AC power to the plant 

 auxiliaries) .............................................................................................................. 3-6 

Figure 9 Normalized PRHR heat flux and mass flow rate of PRHR and CMT injection 

 (loss of AC power to the plant auxiliaries) .......................................................... 3-6 

Figure 10 SG inventory (loss of AC power to the plant auxiliaries) ................................. 3-7 

Figure 11 RCS temperature in loop containing the PRHR (loss of AC power to the 

 plant auxiliaries) ................................................................................................... 3-7 

Figure 12 Normalized PRHR and core power (loss of AC power to the plant 

 auxiliaries) ............................................................................................................ 3-8 

Figure 13 Sensitivity study on different decay heat generation (loss of AC power 

 to the plant auxiliaries) ....................................................................................... 3-8
x 

 
xi 

TABLES

Page

Table 1 Comparison of steady state calculated results ....................................................... 2-5 

Table 2 Sequence for loss of normal feedwater flow event ................................................. 3-9 

Table 3 Sequence for loss of AC power to the plant auxiliaries ......................................... 3-9 
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 model of Maanshan Nuclear 

Power Station is developed by INER. 

According to the TRACE user’s manual, 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. 

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. 

Especially after the Fukushima NPP event occurred, the importance of NPP safety analysis has 

been raised and there is more concern for the safety of the NPPs in the world. AP1000 

(Generation III + reactor) developed by Westinghouse Company is a two-loop 1000MWe 

pressurized water reactor (PWR). Passive safety systems are used to provide significant 

improvements in plant simplification, safety, reliability, investment protection and plant costs. 

AP1000’s safety performance has been verified by extensive testing, safety analysis and 

probabilistic safety assessments. The safe shutdown conditions could be established and 

maintained in AP1000 NPP for 72 hours without operator actions. 

This research presents the applicability of TRACE to simulate AP1000's heat removal decrease 

accidents. The AP1000 NPP TRACE model containing the essential components of the primary, 

secondary loop and passive safety systems with corresponding control systems is established 

through the interface code -SNAP based on the Westinghouse design. The steady-state 

calculation of TRACE is conducted to testify the accuracy of model and the results show a good 

coherent with the design parameters. Two condition II events categorized as the decrease in 

heat removal by secondary system are simulated and TRACE‘s results are consistent with 

Westinghouse's LOFTRAN results. The results of TRACE also reveal that the availability of 

reactor coolant pumps has a significant influence on the passive heat removal performance. 

Moreover, even without any AC power source, the passive core cooling system is capable of 

extracting all the core decay heat without the operator intervention. In conclusion, the analysis 

results of TRACE indicate that the PXS provides the sufficient capacity to establish and 

maintain the long-term core cooling for the plant without human intervention and AC power. 

Severe consequences in Fukushima disaster, such as the core uncovery, could be prevented or 

mitigated. 
xiv 

 
xv 

ABBREVIATIONS

AC Alternating current 

ACC Accumulator 

ADS Automatic depressurization system 

CAMP Code Applications and Maintenance Program 

CMT Core makeup tank 

DVI Direct vessel injection 

ESF Engineering safety feature 

HX Heat exchanger 

INER Institute of Nuclear Energy Research 

IRWST Incontainment refueling water storage tank 

LOCA Loss-Of-Coolant Accident 

MSIV Main Steam Isolation Valve 

NPP Nuclear Power Plant 

PCS Passive containment cooling system 

PMS Protection and safety monitoring system 

PRHR Passive residual heat removal 

PWR Pressurized water reactor 

PXS Passive core cooling system 

RCP Reactor coolant pumps 

RCS Reactor coolant system 

RPV Reactor pressure vessel 

SG Steam generator 

SNAP Symbolic Nuclear Analysis Package 

TRACE TRAC/RELAP Advanced Computational Engine 

USNRC U.S. Nuclear Regulatory Commission 

 
xvi 

 
1-1 

1. INTRODUCTION

The 2011 accident of Fukushima NPP in Japan was the worst nuclear disaster since the 1986 

Chernobyl accident. The tsunami following an earthquake wiped out any available AC power 

sources and the core residual heat was not removed successfully, which further induced a 

series of severe consequences such as reactor core melting and radioactive release. Therefore, 

after the Fukushima NPP event occurred, the importance of NPP safety analysis has been 

raised and there is more concern for the safety of the NPPs in the world. 

AP1000 (Generation III + reactor) developed by Westinghouse Company is a two-loop 

1000MWe pressurized water reactor (PWR). Passive safety systems, which utilize natural 

driving forces such as gravity, natural circulation, and compressed gas-simple physical 

principles, have been applied into the AP1000 NPP design to achieve the objective that it could 

establish the safety shutdown condition and maintain long-term core cooling effectively and 

sufficiently, without the operator intervention or even any AC power support [1]. 

In the previous research, RELAP5/MOD3 code was mainly adopted to implement the AP1000's 

simulation [2][3]. As a new generation and advanced system code developed by the US NRC, 

TRACE (TRAC/RELAP Advanced Computational Engine) code is utilized in this study to 

analyze thermal-hydraulic behavior in AP1000. This code has been world-widely applied and 

demonstrated its excellent accuracy [4][5]. SNAP (Symbolic Nuclear Analysis Package) as a 

graphic user interface code is employed for editing the input decks and exhibiting the simulation 

results straightforwardly. Code versions used in this study are SNAP v 2.2.1 and TRACE v 

5.0p3. 

Former analyses mainly focused on the AP1000's ability coping with loss of coolant accident 

(LOCA) and the relevant thermal hydraulic phenomena [2][6]. Passive containment cooling 

performance responding to LOCA was analyzed by applying different containment codes [7]. 

Long-term heat removal issue in AP1000, however, has not been fully discussed in open 

literature. Considering the Fukushima accident, it is worthwhile to investigate the effectiveness 

of AP1000's passive safety systems resolving the long-term cooling problem and evaluate its 

safety margin. Therefore, two condition II events, loss of normal feedwater flow and loss of AC 

power to the station auxiliaries, are simulated, both of which are categorized as the decrease in 

heat removal by secondary system based on Westinghouse description [8]. Not considering the 

actuation of normal control systems, the passive safety systems become the only approach to 

remove the long-term decay heat and prevent the reactor from overpressurized and core 

melting. 

Based on Westinghouse design information, the AP1000 TRACE model including the primary 

loop, secondary loop and passive core cooling system (PXS) has been developed. The 

corresponding control systems are added to obtain a steady-state condition as well as to 

simulate the accidental transient. In order to achieve more realistic results, a novel two-vessel 

approach is adopted to model the reactor pressure vessel (RPV) with 3D geometry nodalization. 

A steady-state calculation is thereafter executed to investigate the model conformity with the 

design. Two condition II events mentioned previously are simulated and the results are 

compared with that calculated by Westinghouse's LOFTRAN code to gain the information about 

the reproducibility. The impact of the availability of the reactor coolant pumps on the cooling 

performance of passive safety systems is discussed. Sensitivity study about the different decay 

heat generation is implemented to evaluate the influence of the differing fuel burnup conditions 

on the passive residual heat removal (PRHR) capability. 
1-2 

 
2-1 

2. AP1000 TRACE MODEL 

2.1 Model Description 

AP1000 NPP TRACE model can be divided into three parts: the primary loop, the secondary 

loop and passive safety systems. The primary loop or the reactor coolant system (RCS) 

consists of two heat transfer circuits, with each circuit containing one steam generator (SG), two 

reactor coolant pumps (RCP), a single hot leg and two cold legs. The pressurizer is attached to 

the loop A's hot leg. The secondary loop includes the necessary valves like main steam isolate 

valves (MSIV) and safety valves connected to the main steam piping. The passive safety 

systems are made up of two sub-systems: the passive core cooling system (PXS) and the 

passive containment cooling system (PCS). PXS comprises PRHR loop, the safety injection 

loop from core makeup tanks (CMT), accumulators (ACC), incontainment refueling water 

storage tank (IRWST), and automatic depressurization system (ADS) loop. These loops provide 

core residual heat removal, safety injection, and controlled depressurization for reactor. PCS 

affords the safety-related ultimate heat sink for the plant and the residual heat is finally 

transferred to the outside atmosphere by the natural circulation. An overview of the AP1000 

NPP model is presented in Figure 1. 

AP1000 RPV adopts a special four-inlet-two-outlet structure, which is simulated with 3D 

cylindrical geometry in TRACE. Due to the hot leg, cold leg and direct vessel injection (DVI) pipe 

at the different elevations, a two-vessel approach is applied to fulfill a more realistic geometry. 

The vessel simulating the downcomer portion is connected to that simulating the core part 

through the vessel junction. Connected to the hot legs, the vessel simulating the core portion 

divides into 15 levels in the axial direction, 2 rings in the ‘‘r” direction and 8 equal azimuthal 

sectors in the ‘‘θ” direction. The guide tubes between the 12th and 16th layer simulate core 

bypass flow and the heat structure component generating the core power is added into the fuel 

region between the 1st and 10th layer. With the cold legs and two DVI lines attached, the vessel 

modeling the downcomer portion has the identical division method except the different layer's 

width and elevation. DVI lines connect PXS to the vessel for achieving the passive safety 

function. The RPV node diagram is exhibited in Figure 1. 

SG model includes several parts: the primary side with U tubes, and the secondary side 

including the downcomer, boiler part, upper head. These parts are simulated by PIPE 

components with the specific geometries. A separator component is used to achieve the 

vapor-water separation. Hundreds of heat transfer tubes in SG are lumped into one single PIPE 

component with a heat structure attached. The primary and secondary loop components such 

as hot leg and pressurizer are modeled with the PIPE and VALVE components in accordance 

with their specific geometries and features. 

The PRHR loop, including a heat exchanger (PRHR HX), is connected through RCS's hot leg to 

SG's channel head chamber in loop A. The IRWST provides the heat sink for the PRHR HX, 

absorbing decay heat from RCS and transferring it to the containment atmosphere when water 

reaches boiling temperature. When RCS suffers the leak and rupture of various sizes and 

locations, PXS uses three sources of water- the water of the CMTs, the ACC, and the IRWSTto prevent the occurrence of core uncovery through safety injection. The PXS also provides core 

depressurization via the four stages of ADS to permit a relatively slow, controlled RCS pressure 

reduction. These passive safety systems are modeled with the PIPE or VALVE components as 

well. PCS simulation is not discussed in this study. The reason is that the containment cooling 

involves the complex thermal–hydraulics and their propagation during accidents, which requires 

the special containment codes such as CONTAIN and COCOSYS to simulate its evaporation 

and condensation process. The simulation is not the concern in this research and it is replaced 
2-2 

by a simplified PCS makeup flow offered by the FILL component, which compensates the 

IRWST inventory loss when the water evaporates out of IRWST. 

The control systems are categorized into three types: normal plant control, reactor trip and 

engineering safety feature (ESF) trip. The normal plant control system is a nonsafety-related 

system that provides control and coordination of the plant during startup, ascent to power, 

power operation, and shutdown conditions. In this study, pressurizer pressure and water level 

control are modeled to acquire a steady-state operation condition. Reactor trip is a protective 

function performed by the protection and safety monitoring system (PMS). When it anticipates a 

potential threat which generates the signal deviating from the reactor safety limit, the reactor trip 

along with the ESF trips actuates a series of related passive safety systems to bring the reactor 

into safety shutdown condition and prevent any possible core damage. These logic functions 

are achieved by using diverse signal variables, control blocks and trip units. An overview of the 

plant control systems are shown in Figure 2. 

2.2 Steady State Calculation

Since some detail design information on AP1000 are proprietary, the reasonable assumptions 

are employed in the analysis and a trial-and-error approach is adopted to adjust some unknown 

parameters for a desired condition. This steady-state condition is achieved through the 

pressurizer water level and pressure control systems. Constant feedwater boundary is used so 

far to supply the water flow at 100% rated power. It will be evolved into SG water level control 

system achieving the match between feedwater flow and steam flow in future. The comparison 

of the results calculated by TRACE code with the design is listed in Table 1, which 

demonstrates a good agreement. In addition, these parameters can be observed directly from 

the animation mode shown in Figure 3, which helps better understand the simulation results. 

After the desired condition is achieved, the model is modified slightly to simulate transient 

accidents from a restart time setpoint of previous steady-state calculation. 
2-3 

Figure 1 The overview of AP1000 TRACE model
2-4 

Figure 2 The control systems in AP1000 model

Figure 3 AP1000 SNAP animation model 
2-5 

Table 1 Comparison of steady state calculated results

Parameters TRACE value Rated value 

Licensed core power (MW) 3415 3415 

Coolant volume flow per loop (kg/sec) 7572 7578.3 

Pressurizer pressure (MPa) 15.42 15.45 

Pressurizer water (volume/m3

) 28.317 28.347 

Cold leg temperature (K) 553.7 553.4 

Cold leg pressure(MPa) 15.927 16.00 

Hot leg temperature (K) 594 594.4 

Hot leg pressure(MPa) 15.499 15.56 

Coolant average temperature (K) 573.85 573.9 

Steam flow per SG (kg/sec) 943.72 943.4 

SG steam outlet pressure (MPa) 5.764 5.76 

SG power (MWt/unit) 1707.5 1707.3 

SG secondary water mass(kg) 79722.49 76965.77 

SG secondary steam volume (m3

) 147.871 151.03 
2-6 

 
3-1 

3. SIMULATION OF TWO HEAT REMOVAL DECREASE ACCIDENTS 

3.1 Accident Assumptions and Description 

As condition II events, loss of normal feedwater flow and loss of AC power to the plant 

auxiliaries are categorized as the events which result in the decrease of the secondary systems' 

capability on removing the heat from reactor core. The loss of normal feedwater can be caused 

by pump failures, valve malfunctions, or loss of AC power sources. The loss of AC power to the 

plant auxiliaries is possibly caused by a complete loss of the offsite grid accompanied by a 

turbine-generator trip and the onsite standby AC power system is not credited to mitigate the 

accident. Several conservative assumptions used in analysis are as follows: 

 The plant is initially operating at 102% of the rated power. 

 Core residual heat generation is based on ANSI/ANS-5.1-1979, which is a conservative 

representation of the decay energy release rates. 

 Feedwater loss is assumed at 10 sec for both of events. 

 The startup feedwater system is unavailable, and normal plant control systems are not 

considered to function. 

 One of two parallel valves in the PRHR and CMT outlet line fails to open. This is the 

worst single failure. 

 Secondary system steam relief is achieved through the SG safety valves. The 

condenser is unavailable for turbine bypass. 

 For loss of AC power to the plant auxiliaries accident, the offsite AC power is assumed 

to lose along with the reactor trip at 72 sec, which further induces RCPs to coastdown 

simultaneously. This assumption is more conservative than the case in which offsite 

power is lost at time zero. That is because under this assumption SG has a comparable 

lower water level when the reactor trip. 

 The reactor coolant pumps (RCP) leakage is not assumed to happen. 

For both of events, plant vital instruments and related valves' actuation are supplied by the 

Class 1E DC power. Based on the assumptions above, PXS becomes the only system which 

affords emergency core decay heat removal and brings the plant into the safety shutdown 

condition. The only difference between two events is that for loss of AC power to the plant 

auxiliaries, RCP stop rotating at the beginning which minimizes PRHR HX's heat removal 

performance. In addition, coolant flow necessary for core cooling and the residual heat removal 

has to initially rely on the natural circulation in the primary loop and PRHR loop. In order to 

assess the adequacy of PXS coping with the long-term accident, the simulation adopts the time 

interval of 36,000 seconds and a logarithmic time scale is utilized to observe the accidental 

transient behavior. 

3.2 Loss of Normal Feedwater Flow

Event sequence for loss of normal feedwater flow is presented in Table 2. Based on the 

assumptions mentioned, the SG water inventory decreases as a consequence of the continuous 

steam supply to the turbine which leads to the reactor trip on a low SG water level signal at 72 

sec. Due to unavailability of startup feedwater pump and normal plant control systems, a low SG 

water level signal (narrow range), coincident with a low startup feedwater flow rate signal, 

quickly activate the PRHR HX to transfer the core decay heat into the IRWST at about 130 sec. 

Because of the operation of RCPs, the reactor coolant volumetric flow plotted in Figure 4 

remains at a high level until RCPs are tripped by the low Tcold “S” signal (The signal indicates 

the cold leg temperature reaches 533K following continuous cooling). After that, the coolant flow 

necessary for core cooling is maintained by the natural circulation between SG and reactor. 

Since the heat flux of PRHR HX in TRACE is higher than that in LOFTRAN, the termination time 

of RCPs calculated by TRACE is approximately 200 sec earlier than that by LOFTRAN. This 

can be seen from Figure 5. More residual heat is therefore extracted from the RCS, which 
3-2 

further lowers the coolant temperature at a faster speed. Figure 6 shows the coolant 

temperature from TRACE lies slightly below that from LOFTRAN during a period between 130 

sec and 1000 sec. It therefore takes less time to reach low Tcold “S” signal. Another important 

behavior of PRHR HX during this period is that its heat flux slowly declines due to the RCS 

cooldown, but is still higher than the residual heat generation rate shown in Figure 7. 

Once the low Tcold “S” signal is reached, the RCPs start to coastdown followed by the MSIV 

closure and CMT injection with several seconds delay. It can be seen in Figure 6 that the 

coolant temperature continues to decrease from 1000 sec to 2000 sec. During this transient, the 

CMTs operate in water recirculation mode. The CMT injection flow slowly decreases as the 

CMT fluid temperature increases due to water recirculation. The flow eventually disappears 

when the temperature difference between CMT fluid and RCS's coolant vanishes. This can be 

observed in Figure 5. Furthermore, it is worth noting that the CMT injection significantly 

diminishes the heat flux of the PRHR HX, which is represented in Figure 5 by a drastic drop 

appearing at the time of CMT injection. Due to this injection, the heat removal rate shown in 

Figure 7 goes below the core decay heat produced, which leads to the RCS heating up again. 

As the RCS temperature is elevated, the heat removal rate of the PRHR HX increases again 

until the heat extraction rate matches the core decay heat produced at about 17500 sec. Since 

then, this match presented in Figure 7 can be sufficiently maintained till the end of simulation. 

3.3 Loss of AC Power to Plant Auxiliaries 

From the decay heat removal point of view, in the long term this event is more severe than loss 

of normal feedwater flow. This is because, for this case, the decrease in heat removal by the 

secondary system is accompanied by a reactor coolant flow coastdown at the beginning, which 

further reduces the capability of the primary coolant to remove heat from the core. Table 3 lists 

the event sequence for loss of AC power to the plant auxiliaries accident. The reactor starts to 

shut down at 72 sec when the low SG water level signal is reached. Not considering the onsite 

AC power to support, offsite AC power is assumed to lose along with the reactor trip at 72 sec, 

which immediately result in the RCPs' coastdown. Low SG water level signal (narrow range), 

coincident with a low startup feedwater flow rate signal, motivates the PRHR loop to function at 

about 130s. Upon the loss of power to RCPs at the beginning, the primary loop has to initially 

depend on the natural circulation to sustain the essential coolant flow. This is reflected in Figure 

8 by an obvious decline of RCS volumetric flow at 72 sec and roughly 8% of the initial flow 

maintained thereafter. The heat removal rate of PRHR presented in Figure 12 is far lower than 

the core heat generation rate between 130 sec and 2500 sec. The excessive heat need to be 

extracted by the SG through the natural circulation in the primary loop until the decay heat 

generation declines below the PRHR's removal rate at 2500 sec. The PRHR heat removal 

ability shown in Figure 9 remains relatively stable before the CMT injection. This is due to the 

relative constant natural circulation flow existing in primary loop and PRHR loop. 

Although the PRHR HX shows a higher heat flux compared with the results predicted by 

LOFTRAN, the time of cold leg temperature reaching low Tcold “S” signal is approximately 

2000s later than that calculated by LOFTRAN. The reason is that SG heat transfer efficiency 

predicted by TRACE model is lower than that by LOFTRAN. It can be seen in Figure 10 that the 

SG inventory from TRACE simulation lies higher than the results from LOFTRAN, which 

indicates less heat is extracted by SG's water. Therefore, a slightly higher cold leg temperature 

simulated by TRACE can be observed in Figure 11. Accompanied with the continuous core 

power decline and PRHR HX cooling, CMT injection is induced at about 7000 sec, which largely 

lowers the PRHR HX's heat flux. The injected flow reduces along with the rise of CMT fluid 

temperature due to water recirculation as well. RCS starts to heat up again as the result of the 

removal rate dropping below decay heat generation rate. As the RCS coolant temperature rises, 

PRHR HX's heat flux is subsequently elevated to match the decay heat generation. This match 

is achieved at roughly 19500s and a stable cooling condition is thereafter sustained. 
3-3 

3.4 Discussion 

The analysis results above indicate that the availability of RCPs significantly affects the cooling 

performance of PRHR HX. Compared with the loss of AC power to auxiliaries accident, the heat 

transfer rate of PRHR HX can be raised from 0.02% to 0.05% in loss of normal feedwater flow 

accident through 130 sec to 1000 sec. Due to a higher extraction rate from systems, it takes 

less time to actuate the CMT injection on the low Tcold “S” signal. Moreover, the CMT injection 

accelerates the cooldown of plant, which is reflected by a drastic drop in both cold leg and hot 

leg temperature following the injection. 

As the plant gradually cools down, PRHR HX heat removal rate exceeds the decay heat 

produced before CMT injection for both cases. The cold water injected by the CMTs greatly 

lowers the heat transfer of PRHR HX since the injection provides alternative core cooling 

through the water recirculation. The process further benefits the system reaching the match 

between decay heat removal and production. It is worthwhile to stress that PRHR HX has the 

ability to accommodate this heat flux change caused by CMT injection. When the RCS coolant 

temperature increases due to the decreasing cooling effect from CMT, the heat transfer rate of 

PRHR HX is automatically elevated until it can completely match the decay heat produced. The 

operator intervention is entirely not required during the whole process, which reveals the 

inherent safety feature of PXS system. After the match is achieved, a long-term stable shutdown 

condition is established and maintained by PRHR loop. As long as the IRWST water level can 

be maintained, this stable condition can be indefinitely extended. Cooperating with PCS, it is 

achievable to sustain the IRWST inventory for a long time. Besides, this study proves that even 

though loss of any AC power supply occurs at the beginning, the PXS system has the adequacy 

in establishing and maintaining the long-term core cooling, preventing excessive heatup of RCS 

coolant. 

3.5 The Sensitivity Study on Influence of Different Decay Heat Generation

From the beginning of cycle to the end of cycle, the differing fuel burnup has different decay 

heat generation rate. Therefore the sensitivity study is conducted to evaluate this influence on 

the PXS cooling performance. Three decay heat models with different heat generation curves 

presented in Figure 13 are adopted in the analysis. Model A represents the most conservative 

decay heat generation rate which has already been used in the analysis of section 3.2 and 3.3 

Due to more severe condition the loss of AC power to plant auxiliaries accident has, it is 

essential to take this event for example. The CMT injection is actuated at different time because 

less time is taken when removing the smaller amount of the decay heat from the system and 

reaching the low Tcold “S” signal. Although the injection time differs, the match of decay heat 

production with PRHR HX removal rate is eventually achieved for all of models. PRHR cooling 

performance is even slightly stronger in Model B and C which indicates the adequate safety 

margin for cooling. 
3-4 

Figure 4 Normalized RCS volumetric flow rate (loss of normal feedwater flow) 

Figure 5 Normalized PRHR heat flux and CMT injection flow rate (loss of normal 

feedwater flow) 
3-5 

Figure 6 RCS temperature in loop containing the PRHR (loss of normal feedwater flow) 

Figure 7 Normalized PRHR and core power (loss of normal feedwater flow) 
3-6 

Figure 8 Normalized RCS volumetric flow rate (loss of AC power to the plant auxiliaries) 

Figure 9 Normalized PRHR heat flux and mass flow rate of PRHR and CMT injection 

(loss of AC power to the plant auxiliaries) 
3-7 

Figure 10 SG inventory (loss of AC power to the plant auxiliaries) 

Figure 11 RCS temperature in loop containing the PRHR (loss of AC power to the plant 

auxiliaries) 
3-8 

Figure 12 Normalized PRHR and core power (loss of AC power to the plant auxiliaries) 

Figure 13 Sensitivity study on different decay heat generation (loss of AC power to the 

plant auxiliaries) 
3-9 

Table 2 Sequence for loss of normal feedwater flow event 

Event Time(s) 

TRACE LOFTRAN 

Feedwater is lost 10.0 10.0 

Low SG water level reactor trip set point is reached 68.9 70.4 

Rods begin to drop 70.9 72.4 

PRHR HX actuation on low SG water level (narrow range 

coincident with low start up flow rate) 125.7 132.4 

Cold leg temperature reaches low Tcold setpoint 908.1 1154.6 

Reactor coolant pump trip on low Tcold “S” signal 916.1 1160.6 

Steam line isolation on low Tcold “S” signal 926.2 1166.6 

Core makeup tank actuation on low Tcold “S” signal 928.2 1171.6 

PRHR heat exchanger extracted heat matches decay heat ~17500 ~17620 

Table 3 Sequence for loss of AC power to the plant auxiliaries 

Event Time(s) 

TRACE LOFTRAN 

Feedwater is lost 10.0 10.0 

Low SG water level reactor trip set point is reached 70.2 70.4 

Rods begin to drop, AC power is lost, RCPs start to coast down 72.2 72.4 

PRHR HX actuation on low SG water level (narrow range 

coincident with low start up flow rate) 133.6 132.4 

Core makeup tank actuation on low Tcold “S” signal 7078.3 4753 

Steam line isolation on low Tcold “S” signal 7086.3 4765 

PRHR heat exchanger extracted heat matches decay heat ~19500 ~19100 
4-1

 
4-1

4. CONCLUSIONS

In this study, AP1000 TRACE model is constructed and assessed against the Westinghouse 

data. In the evaluation of the steady state condition, the analytical results of TRACE show that 

the parameters have a good consistency with the Westinghouse design data. In transients’ 

analysis, the comparison of TRACE and LOFTRAN results shows that the AP1000 TRACE 

model has a certain level of confidence for the simulation of condition II events. Some important 

conclusions can be drawn as follows: 

 The availability of RCPs has a significant impact on the cooling performance of PRHR 

HX. The heat transfer rate of PRHR HX can be drastically raised if RCPs continue to 

operate after the reactor trip, which accelerates the plant cooldown. 

 CMT injection provides another core cooling approach and speeds up the cooldown of 

plant which further benefits the match between decay heat production and removal. 

 PRHR HX can accommodate the change of heat flux and finally extract all of the decay 

heat generation without the operator intervention and any AC power support. 

 Natural circulation is successful established in both the primary loop and PRHR loop, 

providing the sufficient coolant flow necessary for core cooling. 

 PXS can sustain this long-term core cooling efficiently without the operator intervention 

and any AC power support if the adequate water inventory of IRWST is ensured. 

 PXS can cover the different heat generation rate during the differing fuel life cycle and 

shows a good safety margin. 

In summary, the results of TRACE indicate that the PXS provides the sufficient capacity to 

establish and maintain the long-term core cooling for the plant without human intervention and 

AC power. Severe consequences in Fukushima disaster, such as the core uncovery, could be 

prevented or mitigated. 
4-2

 
5-1

5. REFERENCES

1. Schulz, T.L., 2006. Westinghouse AP1000 advanced passive plant. Nuclear Engineering 

and Design, vol.236, pp.1547–1557. 

2. W.W. Wang, et. al, 2011. Thermal hydraulic phenomena related to small break LOCAs in 

AP1000. Annals of Nuclear Energy, vol. 53, pp. 407-419. 

3. D. Lioce, et. al, 2012. AP1000 passive core cooling system pre-operational tests procedure 

definition and simulation by means of Relap5 Mod. 3.3 computer code. Nuclear Science and 

Engineering, vol. 250, pp. 538–547. 

4. J. Freixa, and A. Manera, 2010. Analysis of an RPV upper head SBLOCA at the ROSA 

facility using TRACE. Nuclear Engineering and Design, vol. 240, pp. 1779–1788. 

5. Jong-Rong Wang, Hao-Tzu Lin , Yi-Hsiang Cheng, Wei-Chen Wang, Chunkuan Shih, 2011. 

TRACE modeling and its verification using Maanshan PWR start-up tests. Annals of Nuclear 

Energy, 36, pp. 527–536. 

6. J. Yang, et. al, 2012. Simulation and analysis on 10-in. cold leg small break LOCA for 

AP1000. Annals of Nuclear Energy, vol. 46, pp. 81–89. 

7. Philipp Broxtermann, Hans-Josef Allelein., 2012. Simulation of AP1000’s passive 

containment cooling with the German Containment Code System COCOSYS. Nuclear 

Engineering and Design, In Press, Corrected Proof. Available online 13 December. 

8. The AP1000 European DCD, 2007a. UK AP1000 Safety, Security and Environmental 

Report. [Online]Available at: https://www.ukap1000application.com/doc_pdf_library.aspx. 




UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA-0439 TRACE Analysis on Heat Removal Decrease Accidents for AP1000