NUREG/IA-0440 

International 

 Agreement Report

The Alternate Mitigation Strategies Study of Chinshan 

BWR/4 by Using the LOCA and SBO Analysis of 

TRACE

 

Prepared by:

Jong-Rong Wang, Chun-Yu 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: 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
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NUREG/IA-0440 

International 

 Agreement Report

The Alternate Mitigation Strategies Study of Chinshan 

BWR/4 by Using the LOCA and SBO Analysis of 

TRACE

 

Prepared by:

Jong-Rong Wang, Chun-Yu 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: 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

Chinshan nuclear power plant is the first NPP in Taiwan which is the BWR/4 plant. This 

research focuses on the development of the Chinshan NPP TRACE model and the LOCA 

combined SBO accident analysis. From the accident at the Japanese Fukusima NPP, an 

extreme event beyond the design basis is realized to be possible. The current mitigation 

strategies for the emergency core cooling systems (ECCSs) can be easily voided in the event of 

an extended station blackout (SBO), where all the onsite and offsite electrical power is failed. 

Although the electrical power of the critical control systems can be recovered by portable 

electrical generators, the electrical pumps are difficult to recover by any portable device. The 

only possible driving force of the pumps in SBO is the steam generated by residual heat. The 

current strategies in an extended SBO are mostly focused on low pressure injection, but the 

reactor water level will decrease sharply while the reactor pressure is reduced and that results 

in a higher PCT. In this report, the alternate mitigation strategies adopting the turbine driven 

pumps, the high pressure injection systems, are analyzed to maintain an “enough” water level 

before the reactor pressure is reduced. Three break sizes, 100%, 10% and 1%, on the 

recirculation suction line of Chinshan NPP which is the most serious LOCA in BWR/4 reactor 

are analyzed with three sensitivity studies: (1) the scram time, (2) the increase of RCIC injection 

flow rate, and (3) the earlier HPCI injection. 
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 Chinshan NPP has been built. In this report, we focus on the 

LOCA combined SBO accident of TRACE analysis. 
vi 

 
vii 

CONTENTS

Page

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

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

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

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

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

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

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

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

2. METHODOLOGY AND MODELING.................................................................................... 2-1 

2.1 The TRACE Code....................................................................................................... 2-1 

2.2 Chinshan TRACE Model ......................................................................................... 2-1 

2.3 Test Cases and Sensitivity Studies ......................................................................... 2-2 

3. RESULTS AND DISCUSSION ............................................................................................ 3-1 

3.1 The Mitigation Strategies of ECCS System ............................................................ 3-1 

3.2 The Extreme Event of LOCA Combined SBO ......................................................... 3-1 

3.3 The Various Break Area on Recirculation Suction Line ........................................ 3-2 

3.4 The Sensitivity of Scram Signals ............................................................................. 3-3 

3.5 The Sensitivity to Increase RCIC Flow .................................................................... 3-3 

3.6 The Sensitivity on Earlier Injecting of HPCI Flow ................................................. 3-4 

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

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

 
ix 

FIGURES

Page

Figure 1 The TRACE model of Chinshan nuclear power plant ......................................... 2-4 

Figure 2 The initial condition and animation model of Chinshan nuclear power plant .. 2-5 

Figure 3 ECCS system mitigation strategy of Chinshan NPP ........................................... 3-5 

Figure 4 The break flow rate with various break area ........................................................ 3-6 

Figure 5 The reactor water level with various break area ................................................. 3-7 

Figure 6 The dome pressure with various break area ....................................................... 3-8 

Figure 7 The steam flow rate with various break area ....................................................... 3-9 

Figure 8 The ECCS injecting flow rate with various break area ..................................... 3-10 

Figure 9 The SCRAM signals in 1% break area ................................................................ 3-11 

Figure 10 The reactor water level in two SCRAM signals and various break area ....... 3-12 

Figure 11 The peak cladding temperature in two SCRAM signals and various 

 break area ........................................................................................................... 3-13 

Figure 12 The peak cladding temperature with various break area ............................... 3-14 

Figure 13 The total break flow with various RCIC flow .................................................... 3-15 

Figure 14 The reactor water level with various RCIC flow and HPCI injecting time ..... 3-16 

Figure 15 The ECCS injecting flow with various RCIC flow and HPCI injecting time ... 3-17 

Figure 16 The dome pressure with various RCIC flow and HPCI injecting time ........... 3-18 

Figure 17 The peak cladding temperature with various RCIC flow and HPCI 

 injecting time ...................................................................................................... 3-19 
x 

 
xi 

TABLES

Page

Table 1 Setpoints of Chinshan NPP .................................................................................. 3-20 

Table 2 Event sequences (scrammed by high drywell pressure) ................................... 3-21 

Table 3 Event sequences (scrammed by low reactor water level) ................................. 3-22 
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 Chinshan NPP 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. 

In the NPP safety, the safety analysis of the NPP is very important work. Especially in 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. Chinshan NPP was building in 

1970. It is the first NPP in Taiwan which is the BWR/4 plant and the original rated power for 

each unit is 1775 MWt. After the project of MUR (Measurement Uncertainty Recovery) for 

Chinshan NPP, Unit 2 started MURPU (Measurement Uncertainty Recovery Power Uprate) from 

April 6, 2008 for Cycle 23 and Unit 1 started MURPU from November 8, 2008 for Cycle 24. The 

thermal power of Chinshan NPP is 1828MWt now. 

This research focuses on the development of the Chinshan NPP TRACE model and the LOCA 

combined SBO accident analysis. From the accident at the Japanese Fukusima NPP, an 

extreme event beyond the design basis is realized to be possible. The current mitigation 

strategies for the emergency core cooling systems (ECCSs) can be easily voided in the event of 

an extended station blackout (SBO), where all the onsite and offsite electrical power is failed. 

Although the electrical power of the critical control systems can be recovered by portable 

electrical generators, the electrical pumps are difficult to recover by any portable device. The 

only possible driving force of the pumps in SBO is the steam generated by residual heat. The 

current strategies in an extended SBO are mostly focused on low pressure injection, but the 

reactor water level will decrease sharply while the reactor pressure is reduced and that results 

in a higher PCT. In this report, the alternate mitigation strategies adopting the turbine driven 

pumps, the high pressure injection systems, are analyzed to maintain an“enough”water level 

before the reactor pressure is reduced. Three break sizes, 100%, 10% and 1%, on the 

recirculation suction line of Chinshan NPP which is the most serious LOCA in BWR/4 reactor 

are analyzed with three sensitivity studies: (1) the scram time, (2) the increase of RCIC injection 

flow rate, and (3) the earlier HPCI injection. Through this report, the alternate mitigation strategy 

using the turbine driven pumps and residual steam is evaluated for the emergency operational 

procedures (EOPs) and the severe accident mitigation guidelines (SAMGs). 
xiv 

 
xv 

ABBREVIATIONS

ADS Automatic Depressurize System 

CAMP Code Applications and Maintenance Program 

CHAN A Component of TRACE to Simulate Fuel Bundles 

CS Containment Spraying 

ECCS Emergency Core Cooling System 

EOPs Emergency Operational Procedures 

FSAR Final Safety Analysis Report 

HPCI High Pressure Core Injecting 

HTC Heat-Transfer Coefficient 

JETPUMP A Component of TRACE to Simulate Jet Pumps 

LBLOCA Large Break Loss of Coolant Accident 

LOCA Loss-Of-Coolant Accident 

LPCI Low Pressure Core Injecting 

LPCS Low Pressure Core Spraying 

MSIV Main Steam Isolation Valve 

NPP Nuclear Power Plant 

PCT Peak Cladding Temperature 

SAMGs Severe Accident Mitigation Guidelines 

SBO Station Blackout 

SEPD A Component of TRACE to Simulate Separators and Dryers

SRV Safety Relieve Valve 

TAF Top of Active Fuels 

TBV Turbine Bypass Valve 

TCV Turbine Control Valve 

USNRC U.S. Nuclear Regulatory Commission 

VESSEL A Component of TRACE to Simulate the Reactor Vessel 

 

 

 

 

 
xvi 

 
1-1 

1. INTRODUCTION

In the NPP safety, the safety analysis of the NPP is very important work. Especially in 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. The TRACE code, the TRAC/ 

RELAP Advanced Computational Engine, is the latest component-based and best-estimate 

reactor system code being developed by the U.S. nuclear regulatory commission (USNRC) for 

analyzing the neutronic and thermal-hydraulic behaviors, operational transients and other 

accident scenarios in light water reactors. Through the international cooperative program, the 

Code Applications and Maintenance Program (CAMP), many organizations participate and 

adopt the TRACE code for various applications. 

From the accident at the Japanese Fukusima NPP, an extreme event beyond the design basis 

is realized to be possible. The current mitigation strategies for the emergency core cooling 

systems (ECCS) can be easily voided in the event of an extended station blackout (SBO) 

because the low pressure injecting system is out of service without AC power and causes a 

serious low water level. In the Fukusima accident, following the current EOPs and SAMGs, the 

reactor pressure was released for the low pressure injecting system. In fact, the reactor 

pressure was still too high for the low pressure injecting system and caused the fuel bundles to 

become uncovered and damaged. It is necessary to analyze more different cases in the 

extreme accident and provide more safety precautions and more operation strategies in an 

accident. 

The most extreme event is the LOCA combined with the station blackout (SBO) event, where all 

the onsite and offsite AC electric power is failed except the DC battery power for the control 

system. Since the low pressure ECCS system relies on electrical pumps, it will fail in SBO and 

the automatic depressurize system (ADS) would not be activated because of an interlock with 

the low pressure core injecting system (LPCI). In an SBO event, the electrical power of the 

critical control systems can be recovered by portable electrical generators, but the electrical 

pumps are difficult to recover by any portable device. The only possible driving force for the 

pumps in SBO is the steam that is generated by residual heat. If the reactor pressure is 

depressurized, the turbine pumps of the RCIC and the HPCF will fail. Making the best use of 

residual steam as a driving force for the high pressure core injecting systems is a possible 

mitigation strategy in the SBO. The two most important targets after the reactor scram are to 

maintain the reactor coolant above the TAF and to remove the residual heat. The turbine driven 

pump can fulfill these two requirements. The current strategies in the extended SBO are mostly 

focused on low pressure injection, but the reactor water level will decrease sharply while the 

reactor pressure is released and that results in a higher PCT. In this report, the alternate 

mitigation strategies using the turbine driven pumps and the high pressure injection system are 

analyzed to maintain an “enough” water level before the reactor pressure is released. With 

sufficient coolant, the fuel temperature will be maintained to prevent fuel bundles being 

damaged. 

Chinshan nuclear power plant with the same BWR/4 reactor as the Fukusima NPP is the first 

nuclear power plant in Taiwan. After the MUR (Measurement Uncertainty Recovery) project in 

2008, the rated thermal power is 1828MWt with the rated steam flow at 3.5 Mkg/h, the core flow 

at 24 Mkg/h and the reactor pressure at 6.98 MPa. The power and the core flow analyzed are 

1864 MWt and 18 Mkg/h, respectively (102% power and 75% core flow). There are 408 fuel 

bundles, two recirculation loops, 20 Jetpumps and 130 separators. In Chinshan NPP, the RCIC 

system and the HPCI system are equipped with turbine driven pumps that provide the injecting 

mass flow at 25.2 liters/sec and 267.8 liters/sec, respectively. 

The TRACE model of Chinshan NPP, consisting of the specific components like 3D VESSEL, 

JETPUMP, SPED and CHAN, was developed using the plant design data and benchmarked 
1-2 

with the steady state of FSAR, the start-up data and the transient results of the RETRAN data. 

In this research, a more serious case than the Fukusima accident, the LOCA combined SBO, is 

analyzed with the two turbine driven pumps, the RCIC pump and the HPCI pump. Three break 

sizes, 100%, 10% and 1%, on the recirculation suction line which is the most serious LOCA in 

BWR/4 reactor are analyzed with three sensitivity studies: (1) the scram time, (2) the increase of 

RCIC injection flow rate, and (3) the earlier HPCI injection. Different break sizes will result in 

different discharging flows, a different declining curve of the reactor pressure, a different scram 

time and a different recovery time of the reactor water level. The reactor scram can be triggered 

by the three signals, the drywell high pressure, the reactor low water level and the high reactor 

pressure. The first two scram signals are analyzed with the three different break areas. The 

scram signal of the drywell high pressure is found to be activated earlier than the other two 

scram signals and the impact is clearer in a small break LOCA. If the drywell high pressure 

signal fails, the reactor will be scrammed by the reactor low water level signal (Level-3) which is 

delayed by several seconds. The SRV is activated to maintain the reactor pressure in the 1% 

break area. 

Although the discharging flow of the 1% break area is less than the other two, the time below 

the TAF level is much longer because the reactor pressure remains high and inhibits the HPCI 

flow injecting. The reactor water level will found not recover until the success injection of the 

HPCI flow with the original RCIC flow. The HPCI flow will not be injected into the reactor until 

the HPCI injecting valve opens at the reactor pressure of 3.44 MPa. The impact on the earlier 

HPCI injection by increasing the setpoint from 3.44 MPa to 6.21 MPa is found to be significant 

for the recovery of the reactor water level. The current RCIC flow is not sufficient even in a 1% 

break area. In this paper, to increase the RCIC flow is also tested with double and triple RCIC 

flow in a 1% break area. It is found that the reactor water level could recover earlier with the 

RCIC flow increasing and is more significant in a small break LOCA. 

With the sensitivity studies on the three break sizes, the scram signals, the opening setpoint of 

HPCI injecting valve and the mass flow of the RCIC system in the paper, a different strategy in 

the extreme accident with SBO is developed. To make the best use of the residual steam as a 

driving force with the high pressure core injecting systems is a possible mitigation strategy in 

the SBO. Through the tests, the emergency operating procedures (EOPs) and severe accident 

mitigation guidelines (SAMGs) can be evaluated again.
2-1 

2. METHODOLOGY AND MODELING 

2.1 The TRACE Code 

TRACE, the TRAC/RELAP Advanced Computational Engine developed by USNRC, has been 

designed to perform best-estimate analyses of loss-of-coolant accidents (LOCAs), operational 

transients, and other accident scenarios in reactor systems. Through the International 

Cooperative Program, Code Applications and Maintenance Program (CAMP), many 

organizations participate and adopt TRACE for various analysis applications. TRACE is a 

modernized code with the capability to simulate the reactor system and model the 

thermal-hydraulic phenomena in three-dimensional space. TRACE is a component-based code 

for fast and integrated inputs of reactor systems. The reactor vessel, fuel bundles, separators 

and dryers and jetpumps are modeled by the specific components, VESSEL, CHAN, SEPD and 

JETPUMP. In addition, TRACE is integrated into SNAP (Symbolic Nuclear Analysis Package), a 

graphic user interface, to assist users developing TRACE input decks. Instead of those 

out-of-date codes like TRAC and RELAP, the TRACE code provides a new man-machine 

interface and will become the NRC’s flagship thermal-hydraulic analysis tool. The SNAP v2.0.4 

and TRACE v5.0p2 were employed in this research. 

2.2 Chinshan TRACE Model

The TRACE model of Chinshan Nuclear Power Plant (1) is developed and based on the plant 

design data; (2) consists of different modules to simulate the reactor systems; and (3) analyzes 

the 3D thermal-hydraulic phenomena through the 3D VESSEL component (shown in Figure 1). 

The reactor vessel is divided into 88 cells with eleven axial elevations, four radial rings and two 

azimuthal sectors. The volume and height of each cell is determined by the plant design data. 

The form loss, the hydraulic diameter and the flow area can be set individually on each of the 

six interfaces of a cell to specify fluid-dynamics and heat transfer calculations. 

The loss coefficient K for an abrupt expansion is calculated as 

2

1

expansion K 1 















 

j

j

A

A

 (1) 

The loss coefficient K for an abrupt contraction is interpreted as below 

Aj+1/Aj 0.0 0.04 0.16 0.36 0.64 1.0 

Kcontractio

n 

0.5 045 0.38 0.28 0.14 0.0 

 

Once the loss coefficient K has been determined using the above equation and data, the 

pressure drop across the abrupt expansion or abrupt contraction may be calculated as follows. 

The flow at section j has a velocity Vj

 , while the flow upon reaching section j1 has a 

velocity Vj1 that is lower or higher than Vj

 because of the abrupt cross-section change. The 

change in pressure from points j

 to j1 which is caused by the abrupt area change yields 

  2 2 2

P

2

1

2 2

1

j j 1 1

 

   















      j j j

j j

V

K

V V

P P  

 (2) 
2-2 

The most external ring is the downcomer while the inner three rings are the core zone. There 

are 408 fuel bundles and 130 separators. The 408 fuel bundles are modeled by six CHANs, 

which are located in the axial level 4 and 5. The separators and dryers are simulated by the six 

SEPDs located from the axial level 7 to 10. The pressure distributions in the reactor vessel, 

including the core area, separators, dryers and the upper plenum, are also validated. The 

Chinshan TRACE model includes two recirculation loops with one recirculation pump and ten jet 

pumps in each, together with four steam lines - each with one safety relieve valve (SRV), one 

main steam isolation valve (MSIV), two turbine control valves (TCVs), and one turbine bypass 

valve (TBV). The SRVs and the TBVs function as reactor pressure control systems to prevent 

the reactor from being over-pressured. The pressure drops along main steam lines and the 

recirculation loops are calculated per the formulas above and validated through the plant data 

and the FSAR of Chinshan NPP [1]. 

Fuel bundles are modeled by six CHANs with point kinetics feedbacks of the delay neutron 

fraction, Doppler reactivity coefficient and void reactivity coefficient [2]. The water rods, partial 

length rods, full length rods, tie rods and the leakage paths between the rods are modeled by 

the CHAN component. The CHAN is divided into twelve axial sections with the last section as 

the fuel handling. The power shape in axial can be set in the POWER component. The core 

bypass flow between the fuel bundles is simulated by the VESSEL component. Level 3 is the 

fuel support that the bypass flow between the fuel bundles can be simulated and divided into six 

sections. The ratio of the bypass flow to the core flow is validated. The CHAN is divided into 

twelve axial sections with the last section as the fuel handling. The power shape in axial can be 

set in the POWER component. 

The specific TRACE components, like the JETPUMP, SPEDs and CHANs, are used in the 

TRACE model of Chinshan NPP. The SEPD component is used to simulate its GE mechanical 

separators and dryers. The 130 separators are modeled by six SPEDs and located in the level 7 

and 8. The SEPD is divided into four cells with three cells to be the main tube and one cell to be 

the side tube which discharges the separated liquid. The side tube simulates the discharging 

flow of the separated liquid that the radial mass flow from outsides of separators to the 

downcomer can be analyzed. The capability to simulate the radial mass flow is important. With 

the SEPD components used, more thermal-hydraulic behaviors in separators and dryers can be 

analyzed, including the void fraction, pressure propagation, fluid velocity in separators, and the 

carryover and carryunder quality. Since the arrangement can be modeled more practically in the 

stand pipe area, the separators area, and the intermediate area, the radial mass flow of the 

discharge liquid to downcomer can be simulated. The SEPD component is validated with 

steady-state baselines and the startup tests, 83% power 75% flow Turbine Trip test and 100% 

power 100% flow Load Rejection test [3]. 

The feedwater control system is a three-element control model by the three parameters of the 

reactor water level, the steam flow rate, and the feedwater flow rate. The normal operating 

water level is at level 7. For a more practical instrument measurement of the reactor water level, 

some specific functions are used for the narrow range water level. The RCIC and the HPCI with 

turbine driven pumps are built with their injection paths via the feedwater lines. The HPCI pump 

provides injection flow at 267.8 kg/sec when the reactor pressure is between 1.03 MPa and 

7.74 MPa. The RCIC pump provides injection flow at 25.2 kg/sec. The driving steam for the 

RCIC and HPCI pump is modeled to simulate the governor valve of a turbine pump. The 

Chinshan TRACE model has been benchmarked through several transient cases [2]-[4] with the 

Chinshan FSAR, the start-up data [5] and the transient results of the RETRAN data [6]-[8]. 

Besides, the initial condition and animation model of Chinshan NPP is shown in Figure 2. 
2-3 

2.3 Test Cases and Sensitivity Studies 

The most limiting LOCA of a BWR/4 reactor is a break on the recirculation suction line. The 

combined accident of the LOCA and extended SBO will result in an extreme event beyond the 

DBAs. In this paper, the extreme event of LOCA combined SBO is analyzed with different sizes 

of break, the 100%, 10% and 1% of the recirculation suction line. The power and the core flow 

analyzed are 1864 MWt and 18 Mkg/h, respectively (102% power and 75% core flow). The 

sensitivity studies on the first two scram signals are analyzed with the three different break 

areas. In addition, sensitivity studies on the two important parameters, the RCIC flow and the 

early opening of the HPCI injecting valve, are analyzed with three break sizes, 100%, 10% and 

1%, on the recirculation suction line. 

In this report, the alternate mitigation strategies adopting the turbine driven pumps, the high 

pressure injection systems, are analyzed to maintain an “enough” water level before the reactor 

pressure is released. An evaluation from the plant wide to the system level is given. The ADS 

and LPCI will be out of service as the high pressure injecting systems, the RCIC and the HPCI 

systems, with turbine driven pumps of are the only available systems. Because the RCIC flow is 

found not sufficient even in a small break LOCA, the two mitigation strategies, to increase the 

RCIC flow and to inject the HPCI flow earlier, are analyzed for a possible resolution in the 

extreme event of the LOCA and the extended SBO. 
2-4 

Figure 1 The TRACE model of Chinshan nuclear power plant 
2-5 

Figure 2 The initial condition and animation model of Chinshan nuclear power plant 
2-6 

 
3-1 

3. RESULTS AND DISCUSSION 

3.1 The Mitigation Strategies of ECCS System 

The ECCS provides a variety of systems for the mitigation strategies of defense in depth which 

is divided into two segments (see Figure 3). The high pressure system includes the HPCI 

system. Although the RCIC system is not a typical ECCS, the turbine driven RCIC pump is 

evaluated as a mitigation injection coolant in this paper. When LOCA, the RCIC system and the 

HPCI system are activated by the reactor low water level (Level 2) and provide high pressure 

injecting mass flow rates at 25.2 kg/sec and 267.8 kg/sec, respectively. In a small break, the 

RCIC system can be started and make up the break flow immediately; in a medium break, the 

HPCI system provides more injecting coolant into the vessel. If the two high pressure injecting 

systems still fail to make up the coolant and the reactor water keeps reducing to reactor water 

level one, the automatic depressurized system (ADS) will be activated to release the reactor 

pressure for the successful injection of the low pressure systems, the LPCI and the containment 

spraying (CS), which require AC power for their electrical pumps. The containment spraying 

(CS), including drywell spraying and wetwell spraying, provides an important pressure reduction 

method to keep the integrity of the containment. 

The RCIC pump, a turbine driven pump, reaches the rated flow in ten seconds and injects 

coolant directly into the reactor. The HPCI system includes the injecting valve which interlocks 

with the reactor pressure. If the reactor pressure is more than 3.44 MPa, the HPCI injecting 

valve will not be opened and the HPCI flow will not inject into the vessel although the HPCI 

pump has been operating. In the current strategy, the ADS system will be activated to release 

the reactor pressure for the injection of the LPCI system which utilizes electrical driven pumps. 

The LPCI system provides a much larger mass flow to make up the reactor coolant in a large 

break LOCA. 

But, in the SBO accident, the mitigation strategy will be voided with the failure of the LPCI 

system and the CS system. When the reactor pressure is released, the RCIC system and the 

HPCI system will lose their driving force and at the same time the reactor water level reduces 

sharply. Without AC power, the low pressure injecting system will be out of service and the 

ECCS systems, including the high pressure and the low pressure injecting systems, will fail to 

makeup reactor coolant after the reactor pressure has been released. Although trying to utilize 

the external injecting system, the reactor pressure is still too high to inject coolant and results in 

the fuel being uncovered and damaged. Thus, adopting an alternate mitigation strategy to utilize 

the best of the residual steam through the RCIC system and the HPCI system instead of 

releasing reactor pressure is necessary in the extreme event of the LOCA and extended SBO. 

3.2 The Extreme Event of LOCA Combined SBO 

The most limiting LOCA of a BWR/4 reactor is a break on the recirculation suction line. The 

combined accident of a LOCA and an extended SBO will result in an extreme event beyond the 

DBAs. All AC power is assumed to be failed, resulting in the out of service of the low pressure 

injecting system and the automatic depressurized system. The current mitigation strategies of 

the ECCS as shown in Figure 3 can be easily voided in an extended station blackout (SBO) 

event, where all the onsite and offsite AC electrical power is failed and results in an LPCI failure. 

The electrical power of the critical control systems can be recovered by portable electrical 

generators, but the electrical pumps are hardly recovered by any portable device. The only 

available driving force for the pumps in SBO is the steam generated by the residual heat. Only 

the turbine driven pumps of the RCIC and the HPCI are available. If the reactor pressure is 

depressurized, the high pressure injecting systems with the turbine pumps will fail and the low 

pressure injecting systems with the electrical pumps also fail. None of injection flow is available. 

In addition, the reactor pressure is still too high for the external injecting system. Without 
3-2 

sufficient coolant, the fuel will be uncovered and the temperature will peak high and cause 

serious fuel damage. A different strategy in the extreme accident with SBO is necessary. To 

make the best use of the residual steam as a driving force with the high pressure core injecting 

systems is a possible mitigation strategy in the SBO. 

At the beginning of LOCA, the reactor water level reduces quickly because of the break flow; the 

reactor is scrammed by the signal of the drywell high pressure which is triggered earlier than the 

reactor low water level and the reactor high pressure. The important events and setpoints of the 

Chinshan NPP are listed in the Table 1. The reactor scram can be initialed by the three signals, 

the drywell pressure high, the reactor low water level and the high reactor pressure. The drywell 

pressure increases and the reactor water level reduces because of the break discharge; the 

reactor pressure peaks up immediately after the closure of the MSIV valves. The closure of the 

MSIV valves is activated by the signal of reactor water level 2. The peaking up of the reactor 

pressure is more significant in a small break LOCA and the SRV is activated to maintain the 

pressure in the 1% break area. The SRVs will be activated several times to prevent the reactor 

pressure from exceeding the limit. The MSIVs are closed by the reactor low water level (Level 2) 

and the steam flow and the reactor water level reduce very quickly. In the main steam lines, only 

the SRVs flow and the driving steam flow of the RCIC pump and the HPCI pump remains 

(shown in Figure 6). 

The scram signal of the reactor high reactor pressure will not be triggered in a large break 

LOCA because the discharging flow is massive that the reactor pressure will not peak up when 

the MSIV closes. The RCIC system and the HPCI system are initiated at the signal of the 

reactor low water level (Level 2) and inject coolant into the vessel via the feedwater lines. But 

the HPCI flow will not be injected into the reactor because the HPCI injecting valve will not be 

opened unless the reactor pressure lowers than 3.44 MPa. Thus, the reactor pressure is found 

to be a dominant factor for the HPCI injection in this paper. In a small break, take 1% area 

break for example, the HPCI flow will not be injected into the vessel although the HPCI pump 

has started at the reactor water level two. The delay will cause the reactor water lower below 

the TAF. In any break area, the reactor water level is not recovered unless the HPCI flow injects 

into the vessel. 

3.3 The Various Break Area on Recirculation Suction Line 

In this report, the power and the core flow analyzed are 1864 MWt and 18 Mkg/h, respectively 

(102% power and 75% core flow). The extreme event of the LOCA and the SBO is analyzed 

with different sizes of break, the 100%, 10% and 1% area of the recirculation suction line. Most 

of the break mass flow discharges in the beginning and reduces sharply after the reactor water 

level goes down (shown in Figure 4 and 5). Because the onsite and offsite AC power is failed, 

the low pressure ECCS systems and the automatic depressurize system (ADS) are failed. The 

high pressure injecting systems, RCIC and HPCI, with the turbine pumps are the only available 

systems. The peak cladding temperature increases at the beginning of LOCA and decreases 

slightly after the reactor scram. Since the reactor water level keeps decreasing, the fuel 

cladding temperature will keep increasing until the recovery of the reactor water level. 

The three different break sizes on the recirculation line will result in different discharging flows, 

the declining curves of the reactor pressure, the SCRAM times and the reactor water levels. In 

the Figure 4, the discharging flows of the three are massive in the first fifty seconds then reduce 

after much of the coolant has been discharged and the water level is below the TAF. Take the 

time sequences in Table 2 for example; the reactor water levels reduce very fast below the TAF 

level in the 100% and 10% break area while the 1% break area keeps above the TAF level in 

the first 65 seconds. 

The reactor water level is found not recovered until the HPCI injection. The HPCI injecting valve 

does not open unless the reactor pressure to be lower than 3.44 MPa. In Figure 6, the reactor 
3-3 

pressure of the 1% break area keeps high and the HPCI injecting is inhibited longer than the 

other two cases although the break discharging is less. The duration below the TAF level of the 

1% break area is much longer than the other two break areas because of the HPCI flow being 

inhibited. It results in a higher peak cladding temperature. The peak cladding temperature keeps 

increasing when the reactor water level coasts down and reaches the highest temperature 

before the recovery of the reactor water level (shown in Figure 7 and 8). 

The peak cladding temperature of the 1% break area is as high as the PCT of the 100% break 

area because the reactor pressure of the 1% break area remains high and inhibits the HPCI 

injection. Thus, a higher setpoint of the HPCI injecting valve opening could be an effective 

mitigation strategy to maintain the temperature and will be discussed later (see Figure 5 and 12). 

To maintain the reactor water level above the TAF is more important in the beginning. If the 

reactor water level reduces below the TAF too early, it will result in a higher peak cladding 

temperature and a lower reactor water level. Although the reactor water level recovers about at 

the same time in the 100% and 10% break area, a short recovery of the reactor water level in 

the beginning of the 10% break area could result in a lower peak cladding temperature (shown 

in Figure 5 and 12) The peak cladding temperature of the 100% break area is almost double 

than the 10% break area. 

3.4 The Sensitivity of Scram Signals

A reactor scram can be initialed by the three signals, the drywell high pressure, the reactor low 

water level and the high reactor pressure. The drywell pressure increases and the reactor water 

level decreases because of the discharging flow, and the reactor pressure peaks up 

immediately after the MSIV valves close which is activated by the signal of reactor water level 2. 

The peaking up of the reactor pressure is more significant in the 1% break area and the SRV 

valve has to be activated for the pressure to be maintained. 

In Figure 9, the high drywell pressure is found earlier than the other two signals, the low reactor 

water level and the reactor high pressure. It is more significant in a small break LOCA. If the 

high drywell pressure signal fails, the reactor will be scrammed by the low reactor water level 

(Level 3), which is several seconds after the high drywell pressure signal. The scram signal of 

the reactor high pressure is the last one and will not be triggered in a large break LOCA. 

From the 100% and 10% break areas in Figure 10, the deviation of the reactor water level is not 

much different between the two scram signals, the high drywell pressure and the low reactor 

water level. In the 1% break area, the reactor water level with the scram signal of the high 

drywell pressure drops slower in the first 200 seconds and recovers earlier (about 49 seconds). 

The trends of the reactor water level of the three break areas can be used to explain the trends 

of the PCT. Take the 1% break area for example. The peak cladding temperature with the low 

reactor water level is lower than the one with the high drywell pressure because of an earlier 

water recovery (see Figure 10 and 11). 

The setpoint of the high drywell pressure is 13.8 kPa higher than the normal drywell pressure. 

The drywell pressure is found to be peaking up quickly even in a small break LOCA. The time 

sequences of the six cases, the three different break areas with the scram signals of the high 

drywell pressure and the low reactor water level, are listed in the Table 2 and Table 3. 

3.5 The Sensitivity to Increase RCIC Flow

In Figure 5, the designed RCIC mass flow, 25.2 kg/sec, is found to be insufficient for the 

discharging flow even in the 1% break area and will result in a lower reactor water level 

although its break flow is less. From Tables 2 and 3, the reactor water levels of the 1% break 

area are found to be below the TAF longer than the other two break areas. 
3-4 

In this paper, the different RCIC mass flows are analyzed with the double and the triple. To 

increase the RCIC mass flow is found to be effective for the reactor water recovery and the 

reactor pressure control, especially in a small break LOCA. Take the 1% break area in the 

Figure 14 for example, if the original injecting pressure of the HPCI flow is considered, 

increasing the RCIC flow provides an earlier recovery of the reactor water level. The recovery 

time of the reactor water level with the double RCIC flow is about 140 seconds earlier and about 

540 seconds with the triple RCIC flow. The decline of the reactor water level is slower with the 

increasing of the RCIC flow. With the triple RICI flow, the time below the TAF is delayed to 300 

seconds while the cases with the other two RCIC flows are less than 100 seconds. To increase 

the RCIC flow will reduce the duration time below the TAF level and is effective to decrease the 

peak cladding temperature (see Figure 15 and 17). 

The reactor pressure declines faster with the increased RCIC flow because the more injecting 

flow the better cooling effect (shown in Figure 16). The faster pressure decline is good for an 

earlier HPCI flow injection and in an earlier reactor water recovery (shown in Figure 14). An 

earlier HPCI injecting flow is found more effective for the reactor water level recovery and the 

lower peak cladding temperature. In addition, the greater reactor pressure decline will also 

reduce the discharging flow (shown in Figure 13). To increase the RCIC mass flow is one of the 

effective mitigation methods in the extreme event of the LOCA and the extended SBO. 

3.6 The Sensitivity on Earlier Injecting of HPCI Flow

The HPCI flow provides a larger injecting flow at 267.8 kg/sec whick is ten times that of the 

RCIC flow. The reactor water level is found to not recover until the successful injection of the 

HPCI flow in the three different break areas. Although the HPCI pump has started at the same 

time as the RCIC pump, the HPCI flow will not be injected into the reactor vessel unless the 

HPCI injecting valve is opened. Since the opening of the HPCI injecting valve is closely related 

to the reactor pressure, the HPCI injecting flow will be inhibited in a reactor pressure above 3.44 

MPa. That is the reason why the reactor water level of 1% break area suffers a longer duration 

than those of 10% and 100% break areas. If the setpoint of the HPCI injecting valve is 

increased from 3.44 MPa to 6.21 MPa, the HPCI flow will inject earlier and the impact is more 

significant than the increasing RCIC flow. The reactor water level is recovered earlier and the 

peak cladding temperature is reduced more. 

In the Figure 14, comparing the two setpoints, 3.44 MPa and 6.21 MPa, it is found that the 

earlier HPCI injection is more effective for the reactor water level recovery. The recovery time of 

the reactor water level is 650 seconds earlier while the double RCIC flow is only 150 seconds 

earlier. 

In the Figure 17, the peak cladding temperature reduces more with the earlier HPCI injection 

when compared with the double and triple RCIC flow. The combined effects of the increasing 

RCIC flow and the earlier HPCI injection are more significant. With the triple RCIC flow and the 

earlier HPCI injection in the 1% break area, the reactor water level could be maintained above 

the TAF level and the peak cladding temperature will not increase. 

From the sensitivity studies on the above proposed modifications, it is found that the impact is 

more significant with an earlier HPCI injection than with the increasing RCIC flow. The earlier 

injection of the HPCI flow is more effective to maintain the reactor water level and reduce the 

peak cladding temperature. From the plant owner’s point of view, to change the opening 

setpoint of the HPCI injecting valve for an earlier HPCI injection is easier and will cost less than 

to increase the capacity of the RCIC pump. 
3-5 

Figure 3 The ECCS system mitigation strategy of Chinshan NPP [1] 
3-6 

Figure 4 The break flow rate with various break area 
3-7 

Figure 5 The reactor water level with various break area 
3-8 

Figure 6 The dome pressure with various break area 
3-9 

Figure 7 The steam flow rate with various break area 
3-10 

Figure 8 The ECCS injecting flow rate with various break area 
3-11 

Figure 9 The SCRAM signals in 1% break area 
3-12 

Figure 10 The reactor water level in two SCRAM signals and various break area 
3-13 

Figure 11 The peak cladding temperature in two SCRAM signals and various break area 
3-14 

Figure 12 The peak cladding temperature with various break area 
3-15 

Figure 13 The total break flow with various RCIC flow 
3-16 

Figure 14 The reactor water level with various RCIC flow and HPCI injecting time 
3-17 

Figure 15 The ECCS injecting flow with various RCIC flow and HPCI injecting time 
3-18 

Figure 16 The dome pressure with various RCIC flow and HPCI injecting time 
3-19 

Figure 17 The peak cladding temperature with various RCIC flow and HPCI injecting 

 time 
3-20 

Table 1 Setpoints of Chinshan NPP 

Parameter/Event Value 

Thermal power (MW) 1828 

Steam flow (Mkg/h) 3.46 

Core flow (Mkg/h) 24 

Feedwater flow (Mkg/h) 3.46 

Vessel dome pressure (MPaG) 6.98 

Number of reactor internal pumps 20 

 

Reactor Scrams 

(1)Drywell Pressure High Original +13.8 kPa 

(2)Rx Low Water Level Level 3(TAF+455cm) 

(3)Dome High Pressure 7.54 MPa 

 

RCIC Pump initialed delay10 seconds 

(1) Drywell Pressure High Original +13.8 kPa 

(2)Rx Low Water Level Level 2(TAF+312cm) 

 

HPCI Pump Initialed delay 30 seconds 

(1) Drywell Pressure High Original +13.8 kPa 

(2)Rx Low Water Level Level 2(TAF+312cm) 

 

HPCI Injecting Valve Open Reactor Pressure<3.44 MPa 

Recirculation Pump Tripped Level 2 

MSIV Closed Level 2 
3-21 

Table 2 Event sequences (scrammed by high drywell pressure) 

Events 

Time Sequences (second) 

100%Area 10%Area 1%Area 

Break Begins 0 0 0 

Reactor Scrams(High Drywell Pressure) 0.54 2.2 6.2 

Dome Pressure High None 14.0 21.4 

Rx Level-3 0.04 4.7 8.1 

Rx Level-2 1.0 7.8 12.5 

Rx Level-1 3.0 11.0 19.6 

RCIC Injecting 11.0 17.8 22.5 

MSIV close 1.0 7.8 12.5 

Rx Level below TAF 4.4 12.5 65 

HPCI starts Injecting(P < 3.44 MPa) 35.8 161 1202 

Rx Level above TAF 246.8 194.1 1265 

PCT reached 242.4 181.6 1100 
3-22 

Table 3 Event sequences (scrammed by low reactor water level) 

Events 

Time Sequences (second) 

100%Area 10%Area 1%Area 

Break Begins 0 0 0 

Reactor Scrams(Rx Level-3) 1.0 6.0 13.0 

Dome Pressure High None 14.0 23.0 

Rx Level-3 0.5 5.6 12.5 

Rx Level-2 1.3 10.1 18.8 

Rx Level-1 3.5 13.9 26.0 

RCIC Injecting 11.3 20.1 28.8 

MSIV close 1.3 10.1 18.8 

HPCI starts Injecting 

(P < 3.44 MPa) 

36.0 163 1198 

Rx Level above TAF 214 194 1314 

PCT reached 209.8 179.3 1285 
4-1

4. CONCLUSIONS

(1) In the SBO event, the low pressure injecting systems with electrical pumps fail and the 

reactor pressure is still too high for the external low pressure injecting system. The 

electrical power of the critical control systems can be recovered by portable electrical 

generators after some improvements from the Fukusima experience, but the electrical 

pumps are difficult to recover by any portable device. The only possible driving force for 

the pumps in SBO is the steam generated by the residual heat. The current strategies in 

the extended SBO are mostly focus on the low pressure injection, but the reactor water 

level will decrease sharply while the reactor pressure is released that results in a higher 

PCT. To utilize the residual steam through the turbine driven RCIC system and HPCI 

system is an alternate mitigation strategy to makeup the reactor coolant than just to 

release the reactor pressure. 

(2) The current RCIC injecting flow is only sufficient for the SBO event without LOCA. With 

the current RCIC flow, the duration of the reactor water level below TAF in the 1% break 

area is longer than in the 100% and 10% break areas because its reactor pressure is 

higher and results in a later HPCI injection. To increase the RCIC flow is helpful for the 

recovery of the reactor water level and the reactor pressure decline, but the impact is 

significant only in a small break LOCA. The reactor pressure declines faster with the 

increasing RCIC flow because more RCIC injecting flow results in a better cooling effect. 

(3) The reactor water level is not recovered until the successful injection of HPCI in three 

break areas. Although the HPCI pump is started at the same time as the RCIC pump, 

the HPCI flow will not be injected into the reactor vessel until the open of the HPCI 

injecting valve. That is the reason why the reactor water level of 1% break area suffers a 

longer duration than those of 10% and 100% break areas. If the setpoint of the HPCI 

injecting valve is increased from 3.44 MPa to 6.21 MPa, the HPCI flow will inject earlier 

and the impact is more significant than the increasing RCIC flow. The reactor water level 

is recovered earlier and the peak cladding temperature is reduced more. 

(4) Two alternative mitigation strategies, the increase of the RCIC flow and the earlier 

injection of the HPCI flow, are effective for the reactor water recovery in the combined 

accident of LOCA and SBO. The earlier injection of the HPCI flow is more effective to 

maintain the reactor water level and to decrease the peak cladding temperature. The 

combined effects of the increasing RCIC flow and the earlier HPCI injection are more 

significant. With the triple RCIC flow and the earlier HPCI injection in the 1% break area, 

the reactor water level could be maintained above the TAF level and the PCT will not 

increase. From the plant owner’s point of view, to change the opening setpoint of the 

HPCI injecting valve for an earlier HPCI injection is easier and costs less than to 

increase the capacity of the RCIC pump. 

(5) The containment pressure is found to increase very quickly even in a small break. The 

containment spraying (CS) provides a spraying function to reduce the drywell and 

wetwell pressure. But the current containment spraying pump uses an electrical pump 

which will fail in an SBO event. The containment pressure will be out of control and will 

lose containment integrity in a combined accident of LOCA and SBO. A backup 

containment spraying with a turbine driven pump is necessary. The excess HPCI flow 

can be utilized as an alternate mitigation strategy for the containment spraying function 

because the HPCI flow is more than the break flow in a small break LOCA. 

In this report, the extreme event of a LOCA and an extended SBO is analyzed using the 

Chinshan TRACE model. Two alternative mitigation strategies with three different break areas 

are analyzed. From the experience of the Fukusima accident, although the reactor pressure is 

reduced, it is still too high for the alternative low pressure injecting system and will cause a 

serious low water level. Thus, an alternative mitigation strategy to utilize the best of the residual 
4-2

steam through the high pressure injecting systems, RCIC and HPCI, is necessary. The current 

design for RCIC flow is not sufficient even in a very small break LOCA. The reactor water level 

is not recovered until the successful injection of the HPCI flow in a LOCA. Two alternative 

mitigation strategies, increasing the RCIC flow and injecting the HPCI flow earlier, are effective 

for an earlier reactor water recovery and a lower peak cladding temperature in the extreme 

event of LOCA and SBO. The current strategies in the extended SBO are mostly focus on the 

low pressure injection, but the reactor water level will decrease sharply while the reactor 

pressure is released that results in a higher PCT. In this report, the alternate mitigation 

strategies adopting the turbine driven pumps, the high pressure injection systems, are analyzed 

to maintain an “enough” water level before the reactor pressure is released. As a result of this 

report, the recommended actions could be adopted for evaluations of the emergency 

operational procedures (EOPs) and severe accident mitigation guidelines (SAMGs).
5-1

5. REFERENCES

1. Taiwan Power Company, 2000. Final Safety Analysis Report, Chinshan Nuclear Power 

Station Unit 1&2 Taipower Report. 

2. Tang, J. R. et al., 2009. Inadvertent Startup of HPCI Transient Analysis for Chinshan BWR/4, 

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

3. Chen, C.Y. et al., 2011. Validation on the SEPD component of the TRACE Chinshan NPP 

Model with Startup Tests. ANS-2011-Winter Conference, October 30- November 3, 2011, 

Washington, DC, USA. 

4. Chen, C.Y. et al., 2010. TRACE modelling of Chinshan NPP benchmark test. 

ASME-ATI-UIT 2010 Conference on Thermal and Environmental Issues in Energy Systems, 

May 16-19, 2010, Sorrento, Italy. 

5. GEZ-6415, 1977. Chinshan Nuclear Power Station Units 1 & 2 Control Systems Design 

Report. 

6. Hsu, W. S., Tang, J. R., 1997. The calculation notes of the startup test benchmark analysis 

for the Chinshan RETRAN-02 model. INER report, Institute of Nuclear Energy Research 

Atomic Energy Council, R.O.C. 

7. Kao, L., Chiang, S.-C., 2005. Peach bottom turbine trip simulations with RETRAN using 

INER/TPC BWR transient analysis method. Nucl. Technol. 149, 265–280. 

8. McFadden, J.H., et al., 1981. RETRAN-02: A Program for Transient Thermal-Hydraulic 

Analysis of Complex Fluid Flow System. EPRI-NP-1850-CCM. 




UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA

-0440 The Alternate Mitigation Strategies Study of Chinshan BWR/4 by 

Using the LOCA and SBO Analysis of TRACE