NUREG/IA-0438 

International 

 Agreement Report

ATWS Analysis of Lungmen ABWR for MSIV Closure 

Transient

 

Prepared by:

Jong-Rong Wang, Ai-Ling Ho*, 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-0438 

International 

 Agreement Report

ATWS Analysis of Lungmen ABWR for MSIV Closure 

Transient

 

Prepared by:

Jong-Rong Wang, Ai-Ling Ho*, 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

The objective of this report is to analyses the MSIV closure ATWS transient for Lungmen ABWR. 

There are three parts in ATWS analysis: ARI, FMCRD run-in and SLCS initiation. 

The ATWS analyses show that the TRACE/PARCS coupling model established in this report 

indeed have ability to analyze the ARI and FMCRD initiation transient. And the design (RRCS) of 

Lungmen ABWR is verified to have an ability to mitigate the ATWS transient. In addition, it also 

shows the importance of control rod. Reactor power will decrease rapidly as control rod run-in. If 

the ARI and FMCRD run-in fail simultaneously, the peak reactor power would still be controlled 

by pressure, RVs, void fraction and RIP rotation speed. However, the reactor core shutdown will 

then rely on the SLCS injection after 300sec. 

The peak pressure of ARI, FMCRD run-in, and SLCS initiation analyses is 9.12, 9.12, 9.40 

MPaG respectively, which is below the 10.342 MPaG limit. And the peak cladding temperature is 

309.5, 309.5, 591.78Ԩ respectively, which is below the 1204Ԩ limit. The oxidation under these 

temperatures is insignificant. Therefore, the primary system criteria and the fuel integrity critria of 

10CFR50.46 are met. 

 
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 coupling model of Lungmen NPP has been built. In this 

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

ATWS transient. 
vi 
vii 

CONTENTS

Page

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

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

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

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

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

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

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

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

2. 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 ............................................................. 2-5 

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

3.1 ARI Analysis .............................................................................................................. 3-2 

3.2 FMCRD Run-In Analysis ........................................................................................... 3-6 

3.3 SLCS Initiation Analysis ......................................................................................... 3-10 

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

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

 
viii 

 

 
ix 

FIGURES

Page 

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

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

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

Figure 4 The procedure of TRACE/PARCS coupling calculation ..................................... 2-5 

Figure 5 The initial condition and animation model of Lungmen ..................................... 3-1 

Figure 6 The core power (ARI) ............................................................................................. 3-2 

Figure 7 The dome pressure (ARI) ....................................................................................... 3-3 

Figure 8 The steam flow (ARI) .............................................................................................. 3-3 

Figure 9 The water level (ARI) .............................................................................................. 3-4 

Figure 10 The maximum average cladding temperature (ARI) ......................................... 3-4 

Figure 11 The core power (FMCRD run-in) ......................................................................... 3-6 

Figure 12 The dome pressure (FMCRD run-in) ................................................................... 3-7 

Figure 13 The steam flow (FMCRD run-in) .......................................................................... 3-7 

Figure 14 The water level (FMCRD run-in) .......................................................................... 3-8 

Figure 15 The maximum average cladding temperature (FMCRD run-in)........................ 3-8 

Figure 16 The core power (SLCS initiation) ...................................................................... 3-11 

Figure 17 The dome pressure (SLCS initiation) ............................................................... 3-11 

Figure 18 The steam flow (SLCS initiation) ...................................................................... 3-12 

Figure 19 The void fraction feedback reactivity (SLCS initiation) .................................. 3-12 

Figure 20 The water level (SLCS initiation) ....................................................................... 3-13 

Figure 21 The core average boron concentration (SLCS initiation) ............................... 3-13 

Figure 22 The maximum average cladding temperature (SLCS initiation) .................... 3-14 

 
x 

 

 
xi 

TABLES 

Page

Table 1 The transient sequences of ARI analysis .............................................................. 3-5 

Table 2 The transient sequences of FMCRD run-in analysis ............................................ 3-9 

Table 3 The transient sequences of SLCS initiation analysis ......................................... 3-15 

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 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 x 106 Kg/h at rated power condition. The designed rated core flow is 52.2 x 106 Kg/h. 

Compared with BWRs, ABWR replaced the recirculation loop by 10 RIPs (reactor internal 

pumps), eliminating the probability of large LOCA. 10 RIPs provide 111% rated core flow at the 

nominal operating speed of 151.84 rad/sec. 

The object of this report is to analyze the ATWS (Anticipated Transient Without Scram) transient 

of Lungmen ABWR. Because the normal scram system fails, the reactor cannot be shutdown. 

ATWS event might lead to severe damage in nuclear reactor. To decrease the risk of ATWS 

transient, in addition to SLCS (Standby Liquid Control System) found in BWR design, ABWR 

added diverse scram system logic, ARI (Alternate Rod Insertion), and an additional insertion 

mechanism, FMCRD (Fine Motion Control Rod Drive) run-in. Thus, in order to know the ability of 

ATWS mitigation of different control system, ATWS analysis in this report was performed with 

three cases: ARI, FMCRD run-in, and SLCS initiation. The first case shows the effectiveness of 

the ARI design. The second case shows the backup capability of FMCRD run-in. The third case 

shows the indepth ATWS mitigation capability of the ABWR. And it was selected the most 

unfavorable of the different scenarios that could lead to an ATWS accident: a closure of the main 

steam isolation valves (MSIVs). 
xiv 

 

 
xv 

ABBREVIATIONS

ABWR Advanced Boiling Water Reactor 

ADS Automatic Depressurization System 

AOO Anticipated Operational Occurrence 

ARI Alternate Rod Insertion 

ATWS Anticipated Transient Without Scram 

CAMP Code Applications and Maintenance Program 

FMCRD Fine Motion Rod Drive 

FSAR Final Safety Analysis Report 

FWCS Feedwater Control System 

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

NRC Nuclear Regulatory Commission 

PARCS Purdue Advanced Reactor Core Simulator 

RIP Reactor Internal Pump 

RPS Reactor Protection System 

RR RIP Runback 

RRCS Redundant Reactivity Control System 

SLCS Standby Liquid Control System 

SNAP Symbolic Nuclear Analysis Package 

SRV Safety Relief Valve 

TPC Taiwan Power Company 

TRACE TRAC/RELAP Advanced Computational Engine 

 
xvi 

 

 
1-1 

1. INTRODUCTION

The issue of Anticipated Transient Without Scram (ATWS) is one of the first for probabilistic risk 

analysis (PRA) and cost-benefit analysis used by the Nuclear Regulatory Commission (NRC) 

Staff in reaching a resolution. ATWS event especially MSIV closure might produce a severe 

accident such as core melt because the failure of automatic scram system would lead reactor to 

high neutron flux, heat flux and vessel pressure. Thus, the object of this report is to analyse the 

MSIV closure ATWS transient of Lungmen ABWR. 

In order to know the ability of ATWS mitigation of different control system, three cases were 

analysed for MSIV closure ATWS analysis. The first one shows the ATWS performance with 

ARI. This case is intended to show the effectiveness of the ARI design. The second case, which 

uses FMCRD run-in, assuming a total failure of ARI, was performed to show the backup 

capability of FMCRD run-in. The third case was analyzed to show the indepth ATWS mitigation 

capability of the ABWR. In this case, both ARI and FMCRD run-in are assumed to fail. Automatic 

boron injection with a 180-second delay is relied upon to mitigate the transient event.

U.S. NRC approved a final ATWS rule, 10CFR50.46, on June 1, 1984. According to 

10CFR50.46, the design should meet the flowing requirement [1]: 

(1) Fuel integrity: The long-term core cooling capability shall be assured by meeting the 

cladding temperature (൏ 1204Ԩ) and oxidation criteria (൏ 17% of the total cladding 

thickness). 

(2) Containment integrity: The long-term containment capability shall be maintained (the 

maximum containment pressure ൏ 0.310MPaG; the suppression pool bulk temperature 

൏ 97.2Ԩ). 

(3) Primary system: The system transient pressure shall be limited (the maximum primary 

stress within the reactor coolant pressure boundary ൏ 10.342MPaG). 

(4) Long-term shutdown cooling: The reactor shall be bought to safe shutdown condition, and 

cooled down and maintained in a cold shutdown condition. 

In addition, the multitude of equipment and procedures for ATWS mitigation used in this analysis 

is consistent with those described in the Lungmen FSAR Chapter 15E, ATWS Performance 

Evaluation. The scenario assumes that after MSIVs closure an ATWS is initiated, and the high 

reactor pressure activates the ARI and FMCRD run-in. If the ARI and FMCRD run-in fail at the 

same time, the boron solution will be injected about 300sec from the transient. 

 
1-2 

 
2-1 

2. MODEL OF LUNGMEN ABWR

The study of MSIV closure ATWS analyses 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 preliminary Lungmen TRACE model is established based on the relevant documents, as 

shown in Figure 1. There are three major control systems implemented in Lungmen TRACE 

model, including feedwater control system, pressure control system, and RIP control system. 

The core is modeled by 18 channels in order to simulate the 872 fuel assemblies. According to 

the assembly in the real reactor, the Vessel is 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° apart, and each azimuthal sector is connected with the feed water line inlet (six feed water 

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 are 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 

is adopted for limiting the maximum SRVs’ flow. 

TRACE provides three types of core power calculation mode: (1) Power table; (2) Point kinetic; 

(3) Coupling PARCS. The MSIV closure ATWS analyses were performed by using point kinetic 

and coupling PARCS. 

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 [2][3]. 
2-2 

Figure 1 Lungmen TRACE model 

 
2-3 

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 2 shows the core 

pattern for Lungmen PARCS model. 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). 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 [4]. Figure 3 shows the control rod pattern for Lungmen PARCS 

model. The 205 control rod are divided into 19 groups, each group has different initial step. 

The preliminary Lungmen PARCS model is established by our laboratory colleagues, Shu-Juan 

Chen and Chia-Ying Chang [5][6]. The k୧୬୤ calculated from PARCS had been verified by that 

from CASMO. The result shows the respectable accuracy in Lungmen PARCS model that the 

error bar is small than 10ିହ. 

Figure 2 Core pattern for Lungmen PARCS model 
2-4 

Figure 3 Control rod pattern for Lungmen PARCS model 

 
2-5 

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 thermal-hydraulic (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. 

Base on this preliminary Lungmen TRACE/PARCS coupling model, T.S. Feng et al. analyzed the 

loss feed water heater transient and compared the results with plant vender data [7]. It shows 

that the Lungmen TRACE/PARCS coupling model has an ability of transient simulation of 

Lungmen NPP. 

Note that the case ARI and FMCRD run-in was performed by TRACE/PARCS coupling model, 

but in case SLCS initiation we substituted PARCS for point kinetics because Lungmen PMAXS 

file used in core power calculation has problems in boron reactivity calculation. 

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

 
2-6 

 
3-1 

3. RESULTS 

Figure 5 displays the initial steady sate condition and the animation model of Lungmen NPP. 

Three cases were analyzed in MSIV closure ATWS analysis: 

 ARI: to show the effectiveness of the ARI design. 

 FMCRD run-in, assuming a total failure of ARI: to show the backup capability of FMCRD 

run-in. 

 SLCS initiation, assuming a total failure of both ARI and FMCRD run-in: to show the 

in-depth ATWS mitigation capability of the ABWR. 

Figure 5 The initial condition and animation model of Lungmen 

 
3-2 

3.1 ARI Analysis

Table 1 shows the sequence of ARI analysis. The MSIV fully closure time is assumed to be 4sec. 

Figure 6~10 show the results of TRACE/PARCS analysis. After MSIVs close, the steam is kept in 

vessel, increasing the dome pressure to reach the scram signal setpoint 7.76MPaG (about 

4sec). The normal scram system fails. Reactor is soon into ATWS transient and initiates the ARI 

signal. According to the data from Taiwan Power Company, it would need 25sec to let the control 

rod all-in: 15sec for initiating the Hydraulic Control Unit (HCU) to insert the control rod into core; 

10sec for control rod insertion. For simulation, ARI (10sec all-in) is assumed to be inserted 15sec 

later (about 19sec) after system receives the high pressure signal. The control rods are all-in 

about 29sec. In Figure 6, compared with the dotted line, representing the core power without 

control rod, the solid line (with control rod) tends to decline about 20sec. It reveals that the core 

power decrease because of the control rods insertion. And the reactor is bought to a safe 

shutdown condition (the core power < 6% rated power) about 28sec. Moreover, the peak dome 

pressure is 9.12MPaG, which is below the 10.342MPaG limit. Figure 10 shows the maximum 

average cladding temperature. The peak average cladding temperature is 309.5Ԩ, which is 

below the 1204Ԩ limit. The oxidation under this temperature is insignificant. Therefore, the 

primary system criteria and the fuel integrity criteria of 10CFR50.46 are met.

Figure 6 The core power (ARI) 
3-3 

Figure 7 The dome pressure (ARI) 

Figure 8 The steam flow (ARI) 
3-4 

Figure 9 The water level (ARI) 

 

Figure 10 The maximum average cladding temperature (ARI) 
3-5 

Table 1 The transient sequences of ARI analysis 

Time (sec) 

0 Transient starts 

4 System reaches the high pressure setpoint 

(7.76MPaG): RIP3 (without M/G set) trips; RIP1 and 

RIP2 (with M/G set) runback (-5% rated speed) to 

minimum speed (47.12 rad/sec); ARI signal is initiated 

19 ARI is initiated 

29 Control rods are all-in. 

 
3-6 

3.2 FMCRD Run-In Analysis

Table 2 shows the sequence of FMCRD run-in analysis. The MSIV fully closure time is assumed 

to be 4sec. Figure 11~15 show the results of TRACE/PARCS analysis. After MSIVs close, the 

steam is kept in vessel, increasing the dome pressure to reach the scram signal setpoint 

7.76MPaG (about 4sec). The normal scram system fails. Reactor is soon into ATWS transient 

and initiates the ARI signal. ARI is assumed to fail, and then system would initiate the FMCRD 

run-in signal. According to the data from Taiwan Power Company, it would need 120sec to let the 

control rod all-in. For FMCRD run-in analysis, because the control rods are driven by electric 

power, there is no need to wait for initiation. Thus, FMCRD run-in is initiated (20sec) 

instantaneously after ARI fail. The control rods are all-in about 140sec. In Figure 11, compared 

with the dotted line, representing the core power without control rod, the solid line (with control 

rod) tends to decline about 20sec. It reveals that the core power decrease because of the control 

rods insertion. And the reactor is bought to a safe shutdown condition (the core power < 6% 

rated power) about 104sec. Moreover, the peak dome pressure is 9.12MPaG, which is below the 

10.342MPaG limit. Figure 15 shows the maximum average cladding temperature. The peak 

average cladding temperature is 309.5Ԩ, which is below the 1204Ԩ limit. The oxidation under 

this temperature is insignificant. Therefore, the primary system criteria and the fuel integrity 

criteria of 10CFR50.46 are met. 

Figure 11 The core power (FMCRD run-in) 
3-7 

Figure 12 The dome pressure (FMCRD run-in) 

Figure 13 The steam flow (FMCRD run-in) 
3-8 

Figure 14 The water level (FMCRD run-in) 

Figure 15 The maximum average cladding temperature (FMCRD run-in) 
3-9 

Table 2 The transient sequences of FMCRD run-in analysis 

Time (sec) 

0 Transient starts 

4 System reaches the high pressure setpoint 

(7.76MPaG): RIP3 (without M/G set) trips; RIP1 and 

RIP2 (with M/G set) runback (-5% rated speed) to 

minimum speed (47.12 rad/sec); ARI signal is initiated 

19 ARI initiation failed. FMCRD starts to run-in after 1sec 

delay. 

140 Control rods are all-in. 

 
3-10 

3.3 SLCS Initiation Analysis

Table 3 shows the sequence of SLCS initiation analysis. The MSIV fully closure time is assumed 

to be 4sec. Figure 16~22 show the results of TRACE analysis. After MSIVs close, the steam is 

kept in vessel, increasing the dome pressure to reach the scram signal setpoint 7.76MPaG 

(about 4sec). The normal scram system fails. Reactor is soon into ATWS transient. Both ARI and 

FMCRD run-in are assumed to fail. The reactor is at high pressure and SRNM (Startup Range 

Neuron Monitor) ATWS permissive for 180sec. After a 180sec delay, Standby boron liquid is 

rejected to bring the reactor shutdown. According to the data from Taiwan Power Company, it 

would need 96sec to inject the boron liquid. Thus, for SLCS initiation analysis, the standby boron 

liquid is assumed to be injected 276sec (180sec ATWS signal delay time and 96sec boron 

injection time) later after receiving the scram signal. 

Figure 19 shows the void fraction feedback reactivity. Before boron injection (0~300sec), the 

high dome pressure, because of MSIV close, leads RVs to open, decreasing the dome pressure 

and increasing the void fraction. The increase of void fraction, then, gives the negative reactivity 

feedback to decrease the core power. Conversely, RV close leads the dome pressure to 

increase, decreasing the void fraction and increasing the core power. About 148sec, the water 

level, as shown in Figure 20, drops to low water level L2 because the feedwater pump trips, 

initiating RIP1 and RIP2 trip. It causes the increase of void fraction and gives another negative 

reactivity feedback to the core power. Thus, the peak reactor power, in this region, would still be 

controlled by pressure, RVs, void fraction and RIP rotation speed. Note that according to the 

design of Lungmen ABWR, the steam-driven feedwater pump would loss the power and trip 

immediately after MSIVs close. For ATWS analyses, we used the conservative assumption that 

the feedwater pump won’t trip until 120sec after receiving the scram signal. After boron injection 

(300~800sec), standby boron liquid absorbs the neutron, gives the reactivity feedback more 

negative, and brings the reactor to shutdown.

The peak dome pressure is 9.40MPaG, which is below the 10.342MPaG limit. Figure 22 shows 

the maximum average cladding temperature. The peak average cladding temperature is 

591.78Ԩ, which is below the 1204Ԩ limit. The oxidation under this temperature is insignificant. 

Therefore, the primary system criteria and the fuel integrity criteria of 10CFR50.46 are met.
3-11 

Figure 16 The core power (SLCS initiation) 

Figure 17 The dome pressure (SLCS initiation) 
3-12 

Figure 18 The steam flow (SLCS initiation) 

Figure 19 The void fraction feedback reactivity (SLCS initiation) 
3-13 

Figure 20 The water level (SLCS initiation) 

Figure 21 The core average boron concentration (SLCS initiation)
3-14 

Figure 22 The maximum average cladding temperature (SLCS initiation)

 
3-15 

Table 3 The transient sequences of SLCS initiation analysis 

Time (sec) 

0 Transient starts 

4 System reaches the high pressure setpoint 

(7.76MPaG): RIP3 (without M/G set) trips; RIP1 and 

RIP2 (with M/G set) runback (-5% rated speed) to 

minimum speed (47.12 rad/sec); ARI signal is initiated 

19 ARI initiation failed. FMCRD starts to run-in after 1sec 

delay. 

20 Control rod insertion is failed. Core power is 

maintained upon 6% rated power. 

124 Feed water pump trips. 

148 Water level decreases to low water level L2: initiating 

RIP1 and RIP2 trip signal (RIP1 trip instantaneously, 

and RIP2 trip 6sec later). 

184 Automated initiation of ADS is inhibited; SLCS is 

initiated 

280 Standby boron liquid is injected into core. 

 
3-16 

 
4-1 

4. CONCLUSIONS

The objective of this paper is to analyse the MSIV closure ATWS transient for Lungmen ABWR. 

The analyses show the importance of control rod. Reactor power will decrease rapidly as control 

rod run-in, bringing the reactor to shutdown quickly (about 28sec and 104sec for ARI and 

FMCRD run-in respectively). If the ARI and FMCRD run-in fail simultaneously, the peak reactor 

power would still be controlled by pressure, RVs, void fraction and RIP rotation speed before 

boron injection (about 300sec). The reactor power would remain upon 6% rated power, and the 

reactor cannot be shutdown safely. However, the reactor core shutdown will then rely on the 

SLCS injection after 300sec. Moreover, the results of MSIV closure ATWS analyses are showed 

below: 

(1) Fuel integrity: The peak cladding temperature is 309.5, 309.5, 591.78Ԩ respectively, which 

is below the 1204Ԩ limit. The oxidation under these temperatures is insignificant. 

(2) Primary system: The peak pressure of ARI, FMCRD run-in, and SLCS initiation analyses is 

9.12, 9.12, 9.40 MPaG respectively, which is below the 10.342 MPaG limit. 

Therefore, the primary system criteria and the fuel integrity critria of 10CFR50.46 are met. The 

RRCS system has an ability of ATWS mitigation on MSIV closure ATWS transient. 

 
4-2 

 
5-1 

5. REFERENCES

1. U.S. NRC, "Requirements for Reduction of Risk from Anticipated Transients Without Scram 

(ATWS) Events for Light-Water-Cooled Nuclear Power Plants," Code of Federal Regulation 

10 CFR 50.62, June 1, 1984. 

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. J. R. Wang and H. T. Lin, "TRACE Analysis of MSIV Closure Direct Scram Event for 

Lungmen ABWR," ICAPP 10, San Diego, CA, USA, 2010. 

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

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

5. S.J. Chen, “Study and Application of Neutronic Model in TRACE code”, Thesis, National 

Tsing-Hua University, Taiwan, 2010. 

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

Thesis, National Tsing-Hua University, Taiwan, 2012. 

7. 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. 






UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA-0438 ATWS Analysis of Lungmen ABWR for MSIV Closure Transient