NUREG/IA-0436 

International 

 Agreement Report

Assessment of LONF ATWS for Maanshan 

PWR Using TRACE Code 

Prepared by:

Jong-Rong Wang, Che-Hao 

Chen*, 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: February 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
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NUREG/IA-0436 

International 

 Agreement Report

Assessment of LONF ATWS for Maanshan 

PWR Using TRACE Code 

Prepared by:

Jong-Rong Wang, Che-Hao 

Chen*, 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: February 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 study is to utilize TRACE code to analyze the reactor coolant system (RCS) 

pressure transients under LONF (Loss of Normal Feedwater) ATWS (Anticipated Transient 

Without Scram) for Maanshan PWR. TRACE is an advanced thermal hydraulic code for nuclear 

power plant safety analysis which is developed by U.S. NRC. Maanshan nuclear power plant 

(NPP) is a Westinghouse three-loop PWR in Taiwan. The rated core thermal power of 

Maanshan with MUR-PU (Measurement Uncertainty Recapture Power Uprate) is 2822 MWt. 

According to Westinghouse anticipated transients without trip report, LONF ATWS was regarded 

as the most severe plant condition. The ASME Code Level C service limit criteria of 22.06 MPa 

(3200 psig) was assumed to be an unacceptable plant condition in SECY-83-293. 

In order to conform to 10 CFR 50.62, Maanshan NPP has set up AMSAC that is diverse from 

reactor trip system to automatically initiate the auxiliary feedwater system and initiate a turbine 

trip under conditions indicative of an ATWS. Since the ATWS analysis is not specified in the 

FSAR, we use TRACE code to assess the RCS pressure for Maanshan NPP. The results 

indicate that RCS pressure could keep within 22.06 MPa with sufficient negative moderator 

temperature coefficient (MTC) and normal work of AMSAC and valves. 
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 ATWS model of Maanshan NPP has been built. In this report, we 

focus on the TRACE analysis of LONF ATWS. 
vi

 
vii

CONTENTS

Page

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

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

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

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

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

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

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

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

2. DESCRIPTION OF MAANSHAN TRACE ATWS MODEL ................................................. 2-1 

2.1 Reactor Pressure Vessel ......................................................................................... 2-1 

2.2 Pressurizer ............................................................................................................... 2-1 

2.3 Steam Generator with Feedwater Control System ................................................. 2-1 

2.4 Steam Dump Control System .................................................................................. 2-1 

2.5 PORVs and SVs of Pressurizer and Main Steam Line ........................................... 2-2 

2.6 AMSAC ..................................................................................................................... 2-2 

3. DESCRIPTION AND ASSUMPTIONS OF LONF ATWS .................................................... 3-1 

3.1 MTC Variations ........................................................................................................ 3-1 

3.2 RCP Trip ................................................................................................................... 3-1 

3.3 Failure of Partial Motor-Driven AFW ....................................................................... 3-1 

4. RESULTS AND DISCUSSIONS ........................................................................................ 4-1 

4.1 MTC Variations ........................................................................................................ 4-1 

4.2 RCP Trip ................................................................................................................... 4-5 

4.3 Failure of Partial Motor-Driven AFW ..................................................................... 4-11 

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

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

 
ix

FIGURES

Page

Figure 1 The TRACE model for Maanshan NPP ........................................................... 2-3 

Figure 2 Cross section and the TRACE model of vessel for Maanshan NPP ............. 2-4 

Figure 3 TRACE model of steam generator and feedwater control system for 

Maanshan NPP ................................................................................................ 2-5 

Figure 4 The TRACE model of steam dump control system for Maanshan NPP ........ 2-6 

Figure 5 AFW system of Maanshan NPP ...................................................................... 3-2 

Figure 6 MTC variations – steam generator inventory ................................................. 4-2 

Figure 7 MTC variations – cold-leg temperature .......................................................... 4-2 

Figure 8 MTC variations – hot-leg temperature ............................................................ 4-3 

Figure 9 MTC variations – pressurizer water level ....................................................... 4-3 

Figure 10 MTC variations – RCS pressure ...................................................................... 4-4 

Figure 11 MTC variations – core power .......................................................................... 4-4 

Figure 12 RCP trip – steam generator inventory ............................................................ 4-6 

Figure 13 RCP trip – cold-leg temperature ..................................................................... 4-6 

Figure 14 RCP trip – hot-leg temperature ....................................................................... 4-7 

Figure 15 RCP trip – pressurizer water level .................................................................. 4-7 

Figure 16 RCP trip – RCS pressure ................................................................................. 4-8 

Figure 17 RCP trip – core power ..................................................................................... 4-8 

Figure 18 RCP trip – RCP inlet subcooling .................................................................... 4-9 

Figure 19 RCP trip – integrated mass flow of steam dump system .............................. 4-9 

Figure 20 RCP trip – core water level ........................................................................... 4-10 

Figure 21 Failure of partial MDAFW – steam generator inventory .............................. 4-12 

Figure 22 Failure of partial MDAFW – cold-leg temperature........................................ 4-12 

Figure 23 Failure of partial MDAFW – hot-leg temperature ......................................... 4-13 

Figure 24 Failure of partial MDAFW – pressurizer water level .................................... 4-13 

Figure 25 Failure of partial MDAFW – RCS pressure ................................................... 4-14 

Figure 26 Failure of partial MDAFW – core power ....................................................... 4-14 

Figure 27 Failure of partial MDAFW – integrated mass flow of pressurizer 

PORVs and SVs ............................................................................................. 4-15 

Figure 28 Failure of partial MDAFW – core water level ................................................ 4-15 

Figure 29 The animation model of Maanshan NPP ...................................................... 4-16 
x

 
xi

TABLES

Page

Table 1 The parameters of the pressurizer for Maanshan NPP ................................... 2-2 

Table 2 Sequences of events of LONF ATWS – MTC variations .................................. 4-1 

Table 3 Sequences of events of LONF ATWS – RCP trip ............................................. 4-5 

Table 4 Sequences of events of LONF ATWS – Failure of partial MDAFW ............... 4-11 
xii

 
xiii

EXECUTIVE SUMMARY

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

between Taiwan and USA on CAMP. INER (Institute of Nuclear Energy Research Atomic Energy 

Council, R.O.C.) 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 NPP is developed. 

Maanshan NPP is the first PWR in Taiwan. Its reactor is made by Westinghouse Company and 

has the rated power of 2822 MWt. The reactor coolant system has three loops and each loop 

has a reactor coolant motor and a steam generator. Besides, the pressurizer is connected with 

the hot-leg piping in loop 2. 

The codes used in this research are TRACE v5.0p3 and SNAP v2.2.1. The Maanshan PWR 

TRACE model is based on Wang et al. [1][2] that V. & V. with FSAR [3] and start-up tests. In 

order to simulate the conditions of ATWS, we establish PORVs, SVs, spray system, AFW 

system, and AMSAC setting etc. into the model. Before transient simulation, it is necessary for

testing the convergence of steady state of the Maanshan NPP TRACE model and comparing the 

TRACE data in a steady state. After completing the components in TRACE model, then 

introduce the AMSAC setting against Westinghouse Anticipated transients without trip report. 

Furthermore, we make some sensitivity studies as MTC variations, RCP trip, and failure of 

partial motor-driven AFW. The results indicate that RCS pressure could keep within 22.06 MPa 

with sufficient negative MTC and normal work of AMSAC and valves. 
xiv

 
xv

ABBREVIATIONS

AFW Auxiliary Feedwater 

AMSAC ATWS Mitigation System Actuation Circuitry 

ATWS Anticipated Transient Without Scram 

CAMP Code Applications and Maintenance Program 

CST Condensate Storage Tank 

DST Demineralized Water Storage Tank 

ESFAS Engineered Safety Features Actuation System 

FSAR Final Safety Analysis Report 

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

LONF Loss of Normal Feedwater 

MTC Moderator Temperature Coefficient 

NPP Nuclear Power Plant 

NRC Nuclear Regulatory Commission 

PORVs Power-Operated Relief Valves 

RCS Reactor Coolant System 

RPS Reactor Protection System 

RTS Reactor Trip System 

S/G Steam Generator 

SNAP Symbolic Nuclear Analysis Program 

SVs Safety Valves 

TAF Top of Active Fuel 

TPC Taiwan Power Company 

TRACE TRAC/RELAP Advanced Computational Engine 

US United States 
xvi

 
1-1 

1. INTRODUCTION

Maanshan nuclear power plant is the only Westinghouse PWR of Taiwan Power Company 

(Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the 

power of Maanshan NPP [4]. The assessments include NSSS (Nuclear Steam Supply System) 

parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of 

auxiliary systems, and transient analyses, etc. Maanshan NPP finally uprates to 2822 MWt since 

2009. 

The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear 

power plant safety analysis. The development of TRACE is based on TRAC and integrating with 

RELAP5 and other programs. SNAP is an integrated suite of programs designed to provide an 

efficient framework for the user of nuclear safety analysis codes. Taiwan and the United States 

have signed an agreement on CAMP (Code Applications and Maintenance Program) which 

includes the development and maintenance of TRACE. Since Taiwan participates in the 

activities of CAMP, we have to re-analyze these transients with TRACE/SNAP to confirm their 

credibility. ATWS for Maanshan with MUR is one of the tasks. 

The LONF ATWS results in a reduction in capability of the secondary system to remove the heat 

generated in the reactor core. If the turbine fails to trip immediately, the secondary water 

inventory will decrease significantly before the actuation of AFW system. The heat removal from 

the primary side decreases, and this leads to increases of primary coolant temperature and 

pressure. The water level of pressurizer also increases subsequently. The heat removal through 

the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation

from primary side. The pressurizer will be filled with water finally, and the RCS pressure might 

rise above the set point of relief valves for water discharge. Then the transient proceeds by the 

negative reactivity feedback due to the temperature increase of coolant. The RCS pressure may 

reach its peak after core power reduction [5]. 

The peak RCS pressure depends on steam generator inventory, primary coolant temperature, 

negative reactivity feedback, and core power, etc. 
1-1 

 
2-1

2. DESCRIPTION OF MAANSHAN TRACE ATWS MODEL

The code versions adopted in this study are TRACE v5.0p3 and SNAP v2.2.1. The Maanshan 

PWR TRACE model is based on Wang et al. [1][2] that verified with start-up test. It is a 

three-loop model, the main components include the reactor pressure vessel, pressurizer, steam 

generators, steam piping in the secondary side (including four sets of steam dump and vent 

valves) and the steam dump control system. Figure 1 shows the whole TRACE model for 

Maanshan NPP, and individual description below. 

2.1 Reactor Pressure Vessel 

The vessel in TRACE is a unique model with three-dimensional geometry rather than 

one-dimensional that most conventional simulation tools are. We define the vessel model for 

Maanshan as 12 levels in the axial direction, two rings in the radial direction (internal and 

external rings) and six equal azimuthal sectors. The control-rod conduit connects the 12th and 

7th layer of the vessel from end to end. The fuel region is between the third and sixth layer, and 

heat conductors are added onto the structures to simulate the reactor core. Each volume has 

individual thermal-hydraulic parameters to assist us in refining the core. Its divisions are shown 

in Figure 2. 

2.2 Pressurizer

The pressurizer is indispensable in ATWS transients for the purpose of pressure regulation at 

primary side. It is a vertical and cylindrical tank with carbon steel, and the components include 

heater, spray, power-operated relief valves, and safety valves, etc. Since the pressurizer plays 

an essential role in a PWR, its design parameters listed in Table 1 must be conformed to the 

plant data. 

2.3 Steam Generator with Feedwater Control System

The steam generator of Maanshan NPP is a Model F, vertical U-tube heat exchanger, with a 

total of 5624 U-tubes. Figure 3 shows the TRACE model of the steam generator, and the 

U-Tube in the primary side is divided into 18 volumes. A FILL component represents ‘‘Hot-leg 

fluid inflow,” and a BREAK component is used to represent ‘‘Cold-leg fluid outflow.” Their inputs 

were derived from real plant temperature and pressure time histories [6][7]. On the secondary 

side, the division of volume is seven for boiler, 13 for downcomer, 13 for steam dome and 

separator. Furthermore, a FILL component is added to represent ‘‘Feedwater inflow,” and a 

BREAK component is added to represent ‘‘steam outflow.” Plant data for feedwater flow and 

other input parameters derived from velocity. Temperature and pressure are used to set initial 

conditions. Feedwater flow is controlled by a three-variable feedwater control system after the 

transient began. 

2.4 Steam Dump Control System

The steam dump control system is composed of ten atmospheric venting valves, six turbine 

bypass valves and the associated piping control apparatus. Figure 4 shows the TRACE model of 

the steam dump control system. This model was established mainly as described in the report of 

INER [8]. The ten atmospheric venting valves and six turbine by-pass valves are grouped into 

four sets in this model: three turbine bypass valves comprise the first set; the other three are 

formed as the second set; five atmospheric venting valves are considered as the third set, and 

the fourth set consists of the rest. 
2-2

2.5 PORVs and SVs of Pressurizer and Main Steam Line

In order to regulate the pressure rise in an ATWS transient, the PORVs and SVs are indeed 

necessary. There are three PORVs and one SV of the pressurizer; because the PORVs work 

normally in the ATWS transients, they could be lumped into one set in the TRACE model. 

Besides, the amount of PORVs and SVs of each main steam line are one and five, respectively. 

The SVs of the main steam line cannot be lumped on account of different set points. The above 

valves established in the TRACE model based on plant specific parameters, especially the rated 

flow rates and boundary conditions. 

2.6 AMSAC 

In order to conform to 10 CFR 50.62, Maanshan NPP has set up AMSAC that is diverse from the 

reactor trip system to automatically initiate the auxiliary feedwater system and initiate a turbine 

trip under conditions indicative of an ATWS. AMSAC is a backup system that initiates if the RTS 

(Reactor Trip System) and ESFAS (Engineered Safety Features Actuation System) fail to work 

following an ATWS. It will be initiated under three separate conditions: (1) Low steam generator 

water level, (2) Low main feedwater flow, (3) Main feedwater pump trip or main feedwater valve 

closure. It has a delay time for making sure that the RPS is failed, the amount depends on the 

core power at that time which is about 30 seconds at hot full power. 

Table 1 The parameters of the pressurizer for Maanshan NPP 

Parameter 

Pressurizer volume 39.64 m3

Full power water level 56.5% 

Pressurizer flow length 11.74 m

Spray valve max. flow rate (total 2 valves) 44 L/sec 

Heater (proportional heater and backup heater) 1400 W 

Safety valve set point 17.24 MPa

Power-operated relief valve set point 16.20 MPa 
2-3

Figure 1 The TRACE model for Maanshan NPP 
2-4

Figure 2 Cross section and the TRACE model of vessel for Maanshan NPP 
2-5

Figure 3 TRACE model of steam generator and feedwater control system for Maanshan 

NPP 
2-6

Figure 4 The TRACE model of steam dump control system for Maanshan NPP 
3-1

3. DESCRIPTION AND ASSUMPTIONS OF LONF ATWS 

The PWR sequence starts with an anticipated transient and the electrical or mechanical failure 

of the RPS. In a PWR, the ATWS transient results in a RCS pressure rise, the magnitude and 

timing of which is dependent on the MTC, the relief capacity, and the energy removal capacity of 

the steam generators [9]. 

Before any transient analysis using TRACE whole plant model, a consistent set of parameters 

used in TRACE must be obtained from the process of steady-state initialization. The parameters 

computed from steady-steady initialization such as feedwater/steam flows and water level of the 

steam generators, water level and pressure of the pressurizer. And the hot-leg temperatures 

were then compared with real plant data. After completing the steady-state initialization, the 

ATWS transient analysis began with point-kinetic power calculation which reactivity coefficients 

defined by the parameters: fuel temperature coefficient (Doppler coefficient), coolant 

temperature coefficient (optional), gas volume coefficient (optional), and solute mass coefficient 

(optional). 

The followings are the general assumptions of Maanshan ATWS: 

(1) LONF transient was initiated at 10 sec after the beginning of calculation 

(2) Main feedwater flow descended to zero in the first four seconds of transient 

(3) According to the design basis of AMSAC, the maximum time of signal delay at full power 

was 30 sec; therefore, the turbine was assumed to trip at 40 sec, and the AFW was 

initiated at 70 sec 

(4) Normal operation of pressurizer pressure control, including heaters, spray, PORV and 

SV 

(5) Normal operation of main steam valves, including PORV, SV and steam dump system 

(6) No credit for automatic reactor trip 

(7) No credit for automatic control rod insertion as reactor coolant temperature rises 

Besides, sensitivity studies as well as MTC variations, RCP trip, failure of partial motor-driven 

AFW were taken into consideration specifically. 

3.1 MTC Variations 

The MTC taken for calculation of Maanshan NPP in early days were -12.6 pcm/K (-7 pcm/°F) at 

1% burn-up and -14.4 pcm/K (-14.4 pcm/°F) at 10% burn-up. Since 1993, the Taiwan Power 

Company (TPC) in order to extend fuel cycle to 18 months, the maximum MTC was tightened to 

-7.2 pcm/K (-4 pcm/°F). Therefore, we took these MTC settings into account to assess the 

resulting pressure, and chose maximum part as the conservative condition of ATWS 

calculations. 

3.2 RCP Trip

RCP trip may result in lower RCS pressure and coast down of loop flow that causes lower 

capacity of heat removal from reactor core; furthermore, Lower RCS pressure also brings about 

lower saturation temperature. The set-point of RCP trip was set to be 3.3 K (6 °F) of inlet 

subcooling to prevent impeller cavitation, and the pumping curve of RCP after trip was 

calculated by TRACE built-in Westinghouse Curves. 

3.3 Failure of Partial Motor-Driven AFW

The AFW system of Maanshan NPP has two motor-driven pumps and one turbine-driven pump 

and relative piping to each steam generators, the water sources can be drained from either 
3-2

condensate storage tank (CST), demineralized water storage tank (DST) or raw water (Figure 5). 

AFW played the important role to maintain inventory of steam generators to fulfill the capacity of 

heat removal from primary side. Hence, one motor-driven AFW was cancelled due to electrical 

or mechanical failure as valve closure to simulate partial loss of AFW. 

Figure 5 AFW system of Maanshan NPP[10] 
4-1

4. RESULTS AND DISCUSSIONS 

4.1 MTC Variations

Table 2 shows the sequences of events, Transients began at 10 sec with following loss of 

feedwater flow in 4 seconds that caused the steam generator inventory to decrease gradually 

(Figure 6). Main feedwater pump trip brought about AMSAC standby. The secondary side began 

to lose its heat-sink property because of decreasing heat removal from steam generators and 

made temperature at primary side rise (Figure 7 and Figure 8). Turbine tripped at 40 sec that 

initiated by AMSAC, loss of ultimate heat sink induced system pressure rise rapidly. The 

pressurizer spray system began to drain coolant from cold-leg to the upper plenum of 

pressurizer to mitigate the rise of temperature and pressure at 41 sec. Furthermore, the RCS 

temperatures rose to their peak values that led coolant density in the primary side descend. 

Density drop of primary coolant made its volume increase to fill all the room of pressurizer 

(Figure 9); meanwhile, the pressurizer valves as well as main steam line valves were initiated to 

mitigate the pressure rise. AFW entered steam generator at 70 sec to supply inventory, but the 

RCS pressures still rose to their peak values of the ATWS transients (Figure 10, 20.92 MPa for 

-7.2 pcm/K, 19.33 MPa for -12.6 pcm/K, and 18.99 MPa for -14.4 pcm/K). Although the timing for 

pressurizer with water filled was later with more negative MTC, the peak RCS pressure reached 

earlier. Because more negative MTC causes slower pressure rise, which means slower growth 

of pressurizer water level and stronger mitigation of pressure rise. After 300 sec, there were 

several small fluctuations for -12.6 pcm/K and -14.4 pcm/K because of the opens of pressurizer 

PORVs (set-point at 16.20 MPa). With the effort of PORVs, SVs, pressurizer spray system, 

and descend of core power (Figure 11, the core power decreased to about 15% of hot full power 

in 5 min), the RCS pressures were kept within 22.06 MPa. For the following discussions, we took 

-7.2 pcm/K as the conservative condition of MTC. 

Table 2 Sequences of events of LONF ATWS – MTC variations 

MTC setting (pcm/K)

Transient (sec) -7.2 -12.6 -14.4 

Transient initiates 10

Main feedwater trips 10 – 14 

Turbine trips 40

Pressurizer sprays actuate 41 41 41 

Pressurizer PORVs open 43 44 44

Main steam line PORVs open 44 44 44 

Main steam line SVs open 52 51 51

Full AFW flow actuates 70 

Pressurizer safety valves open 79 81 82

Pressurizer fills with water 83 85 86 

Peak RCS pressure reaches 112 109 109
4-2

Figure 6 MTC variations – steam generator inventory 

Figure 7 MTC variations – cold-leg temperature 
4-3

Figure 8 MTC variations – hot-leg temperature 

Figure 9 MTC variations – pressurizer water level 
4-4

R

C

S

pressure (M

P

a)

Figure 10 MTC variations – RCS pressure 

0 200 400 600

Time (sec)

0

1000

2000

3000

Core power (MWt)

MTC = -7.2 pcm/K

MTC = -12.6 pcm/K

MTC = -14.4 pcm/K

Figure 11 MTC variations – core power
4-5

4.2 RCP Trip

Table 3 shows the sequences of events, Transient began at 10 sec with following loss of 

feedwater flow in 4 seconds that caused the steam generator inventory to decrease gradually 

(Figure 12). Main feedwater pump trip brought about AMSAC standby. The secondary side 

began to lose its heat-sink property because of decreasing heat removal from steam generators 

and made temperature at primary side rise (Figure 13 and Figure 14). Turbine tripped at 40 sec 

that initiated by AMSAC, loss of ultimate heat sink induced system pressure rise rapidly. The 

pressurizer spray system began to drain coolant from cold-leg to the upper plenum of 

pressurizer to mitigate the rise of temperature and pressure at 41 sec. Furthermore, the RCS 

temperatures rose to their peak values that led coolant density in the primary side descend. 

Density drop of primary coolant made its volume increase to fill all the room of pressurizer 

(Figure 15); meanwhile, the pressurizer valves as well as main steam line valves were initiated 

to mitigate the pressure rise. AFW entered steam generator at 70 sec to supply inventory, the 

RCS pressures rose to its peak value of 20.92 MPa (Figure 16) and the core power descended 

to about 18% (Figure 17) of hot full power in 5 min. As a result of timing of RCP trip was behind 

that of peak pressure (Figure 18), RCP trip had done no effort to pressure rise that resulted in 

the same value as no RCP trip. But RCP trip made RCS temperature descend which resulted in 

less mass dumped by steam dump system (Figure 19), the accumulated heat evaporated 

coolant that brought about lower core water level to approach TAF (Top of Active Fuel) that 

threatened the integral of fuel rods (Figure 20).

Table 3 Sequences of events of LONF ATWS – RCP trip 

Transient (sec) No RCP trip RCP trip 

Transient initiates 10 

Main feedwater trips 10 – 14 

Turbine trips 40 

Pressurizer sprays actuate 41 41 

Pressurizer PORVs open 43 43 

Main steam line PORVs open 44 44 

Main steam line SVs open 52 52 

Full AFW flow actuates 70 

Pressurizer safety valves open 79 79 

Pressurizer fills with water 83 83 

Peak RCS pressure reaches 112 112 

RCPs trip None 124 
4-6

Figure 12 RCP trip – steam generator inventory 

Figure 13 RCP trip – cold-leg temperature 
4-7

Figure 14 RCP trip – hot-leg temperature 

Figure 15 RCP trip – pressurizer water level
4-8

0 600 1200 1800

Time (sec)

2

6

10

14

18

22

No RCP trip

RCP trip

Figure 16 RCP trip – RCS pressure 

Figure 17 RCP trip – core power
4-9

0 600 1200 1800

Time (sec)

0

20

40

60

80

No RCP trip

RCP trip

Figure 18 RCP trip – RCP inlet subcooling 

Integrated mass flow (m.t.)

Figure 19 RCP trip – integrated mass flow of steam dump system 
4-10

Figure 20 RCP trip – core water level 
4-11

4.3 Failure of Partial Motor-Driven AFW

Table 4 shows the sequences of events, Transient began at 10 sec with following loss of 

feedwater flow in 4 seconds that caused the steam generator inventory to decrease gradually 

(Figure 21). Main feedwater pump trip brought about AMSAC standby. The secondary side 

began to lose its heat-sink property because of decreasing heat removal from steam generators 

and made temperature at primary side rise (Figure 22 and Figure 23). Turbine tripped at 40 sec 

that initiated by AMSAC, loss of ultimate heat sink induced system pressure rise rapidly. The 

pressurizer spray system began to drain coolant from cold-leg to the upper plenum of 

pressurizer to mitigate the rise of temperature and pressure at 41 sec. Furthermore, the RCS 

temperatures rose to their peak values that led coolant density in the primary side descend. 

Density drop of primary coolant made its volume increase to fill all the room of pressurizer 

(Figure 24); meanwhile, the pressurizer valves as well as main steam line valves were initiated 

to mitigate the pressure rise. AFW generated by one turbine-driven pump and one motor-driven 

pump entered steam generator at 70 sec to supply inventory, the RCS pressures rose to its peak 

value of 21.80 MPa (Figure 25) and the core power descended to about 10% (Figure 26) of hot 

full power in 5 min. Trip of one motor-driven AFW pump caused decrease of AFW flow that 

resulted in less heat removal capacity by steam generator and further rise of RCS temperature 

and pressure. During the opening of pressurizer PORVs and SVs, the increase of pressure 

difference made larger amount of coolant dump through valves (Figure 27). Therefore, the 

pressurizer water level (Figure 24) and core water level (Figure 28) became lower as coolant 

density shrunk behind peak pressure. Furthermore, the animation of the TRACE model is 

presented using the animation function of SNAP/TRACE interface with the analysis results. The 

animation model of Maanshan NPP is shown in Figure 29.

Table 4 Sequences of events of LONF ATWS – Failure of partial MDAFW

Transient (sec) No AFW trip One MDAFW trip

Transient initiates 10 

Main feedwater trips 10 – 14 

Turbine trips 40 

Pressurizer sprays actuate 41 41 

Pressurizer PORVs open 43 43 

Main steam line PORVs open 44 44 

Main steam line SVs open 52 51 

Full AFW flow actuates 70 

Pressurizer safety valves open 79 79 

Pressurizer fills with water 83 83 

Peak RCS pressure reaches 112 112 
4-12

Figure 21 Failure of partial MDAFW – steam generator inventory 

Figure 22 Failure of partial MDAFW – cold-leg temperature
4-13

Figure 23 Failure of partial MDAFW – hot-leg temperature 

Figure 24 Failure of partial MDAFW – pressurizer water level 
4-14

0 600 1200 1800

Time (sec)

10

14

18

22

No AFW trip

One MDAFW trip

Figure 25 Failure of partial MDAFW – RCS pressure 

Figure 26 Failure of partial MDAFW – core power 
4-15

Figure 27 Failure of partial MDAFW – integrated mass flow of pressurizer PORVs and 

SVs 

Figure 28 Failure of partial MDAFW – core water level 
4-16

Figure 29 The animation model of Maanshan NPP 
5-1

5. CONCLUSIONS

By using SNAP/TRACE, this study predicts the transient phenomena of LONF ATWS and 

assesses the peak pressures. According the simulation results, the peak pressures were 20.92 

MPa for -7.2 pcm/K, 19.33 MPa for -12.6 pcm/K, and 18.99 MPa for -14.4 pcm/K; moreover, 

even the conservative MTC condition of -7.2 pcm/K was employed, the RCS pressure could still 

keep within the ASME Code Level C service limit criteria of 22.06 MPa. The negative MTC, 

normal operations of PORVs and SVs, and heat removal capacity by steam generator (which 

inventory maintained by AMSAC function and relative facilities) are the important parts to 

mitigate pressure fluctuations and make coolant cover fuel rods. 

In order to simulate further severe situations, we choose RCP trip for primary loop and failure of 

partial MDAFW for secondary loop that both result in less heat removal during transients. As 

RCPs trip at 3.3 K of inlet subcooling to prevent impeller cavitation, it results in lower RCS 

pressure (but no effort to peak pressure), but RCP trip brings about lower core water level to 

approach TAF that threatens the integral of fuel rods. RCPs should be mandatorily kept working 

as ATWS takes place. AFW is necessary to steam generator inventory, trip of one MDAFW 

results in less inventory and following pressure rise due to accumulated heat at primary side. It 

also causes lower coverage of core water level. Therefore, regularly inspecting the facilities of 

AFW is helpful to mitigation of ATWS. 
5-2

 
6-1 

6. REFERENCES

1. J.R. Wang, H.T. Lin, Y.H. Cheng, W.C. Wang, C. Shih, “TRACE modeling and its verification 

using Maanshan PWR start-up tests”, Annals of Nuclear Energy, Vol. 36 pp. 527-536, 2009. 

2. J.R. Wang, H.T. Lin, C. Shih, “Assessment of the TRACE Code Using Transient Data from 

Maanshan PWR Nuclear Power Plant”, NUREG/IA-0241, 2010. 

3. Taiwan Power Company, “Final Safety Analysis Report of the Maanshan Nuclear Power 

Station”, 1982. 

4. Atomic Energy Council, “The Investigation Report for Maanshan MUR”, Taiwan, 2008. 

5. P.H. Huang, L. Kao, "ATWS Analysis for Maanshan Units 1 and 2," Taiwan Power Company, 

Taiwan, 1993. 

6. J.R. Wang, C.Y. Liu, Y.S. Chen, S.F. Wang, “Maanshan nuclear power plant startup tests 

and transient events documentation”, INER Report, INER-T1320, Institute of Nuclear Energy 

Research, Atomic Energy Council, Taiwan, 1989. 

7. T.C. Lyie, T.C. Cheng, C.H. King, “A tape data management system for Maanshan nuclear 

power plant”, INER Report, INER-OM-0338, Institute of Nuclear Energy Research, Atomic 

Energy Council, Taiwan, 1997. 

8. J.R. Wang, Y.S. Chen, S.F. Wang, “Maanshan unit2 load reduction and net load trip tests 

transient analyses”, INER Report, INER-0868, Institute of Nuclear Energy Research, Atomic 

Energy Council, Taiwan, 1988. 

9. U.S. Nuclear Regulatory Commission, “Regulatory Effectiveness of the Anticipated Transient 

Without Scram Rule”, NUREG-1780, USA, 1978. 

10. Taiwan Power Company, “Training Center Report of the Maanshan Nuclear Power Station”, 

1995. 




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NUREG/IA-0436 Assessment of LONF ATWS for Maanshan PWR Using TRACE Code