NUREG/IA-0437 

International 

 Agreement Report

Sensitivity Study of the DEG LBLOCA Transient on the

Counter-Current Flow Limitation by Using 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: 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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infringe privately owned rights. 
NUREG/IA-0437 

International 

 Agreement Report

Sensitivity Study of the DEG LBLOCA Transient on the

Counter-Current Flow Limitation by Using 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: 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

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 a sensitivity study on the 

counter-current flow limitation (CCFL) model. The CCFL model plays a key role in any large 

break loss of coolant accident (LBLOCA) analysis since it affects the calculated discharge flow, 

reflooding time and peak cladding temperature (PCT). In this report, a sensitivity study on the 

CCFL model is performed, by modeling LBLOCA occurring at the Chinshan NPP. The scenario 

assumes 102% power and 75% core flow, with a double-ended guillotine (DEG) break on the 

recirculation loop, which is the most limiting LBLOCA for a BWR/4 reactor. Two break locations, 

i.e. on the suction and the discharge side of a recirculation pump, are evaluated, with high 

pressure core injecting (HPCI) and low pressure core spraying (LPCS) available whereas low 

pressure core injecting (LPCI) failed. The TRACE code is used for the analysis. The Chinshan 

TRACE model was benchmarked against steady-state and transient data contained in the plant 

FSAR report, as well as start-up data and the transient results using the RETRAN code. The 

thermal hydraulic phenomena in the lower plenum area and the jet pumps are also analyzed. 
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 

sensitivity study of the CCFL model of Chinshan NPP. 
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 Description of LBLOCA .............................................................................................. 2-1 

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

2.3 Pressure Drops of Reactor Vessel and Recirculation Loop ................................ 2-2 

2.4 The CCFL Model of the TRACE Code ....................................................................... 2-3 

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

3.1 The Break Flow of the Recirculation Loop .............................................................. 3-2 

3.2 The Transient Curves of Core Flow ........................................................................ 3-2 

3.3 The Curve of Peak Cladding Temperature ............................................................. 3-2 

3.4 The Impact of CCFL Effect ....................................................................................... 3-3 

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

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

 
ix

FIGURES

Page

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

Figure 2 Pressure distributions in recirculation loop ....................................................... 2-6 

Figure 3 Comparison of the total break flow ...................................................................... 3-4 

Figure 4 Comparison of reactor pressure .......................................................................... 3-4 

Figure 5 Comparison of the core flow ................................................................................ 3-5 

Figure 6 Comparison of peak cladding temperature ......................................................... 3-5 

Figure 7 Comparison of the jet pumps exit flow rate at the broken loop ....................... 3-6 

Figure 8 Comparison of the jet pumps suction flow rate at the broken loop ................. 3-6 

Figure 9 Comparison of the reactor water level ................................................................ 3-7 

Figure 10 Comparison of driven flow rate of jet pumps at the broken loop..................... 3-7 

Figure 11 Comparison of the steam flow in lower plenum................................................. 3-8 

Figure 12 Comparison of the water level in lower plenum ................................................. 3-8 

Figure 13 Comparison of the liquid flow of fuel support .................................................... 3-9 
x

 
xi

TABLES

Page

Table 1 The Initial conditions for the LBLOCA of Chinshan NPP ...................................... 3-9 

Table 2 Event Times for LBLOCA of Recirculation Loop of Chinshan NPP ................... 3-10 
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 a sensitivity 

study on the counter-current flow limitation (CCFL) model. The CCFL model plays a key role in 

any large break loss of coolant accident (LBLOCA) analysis since it affects the calculated 

discharge flow, reflooding time and peak cladding temperature (PCT). In this report, a sensitivity 

study on the CCFL model is performed, by modeling LBLOCA occurring at the Chinshan NPP. 

The scenario assumes 102% power and 75% core flow, with a double-ended guillotine (DEG) 

break on the recirculation loop, which is the most limiting LBLOCA for a BWR/4 reactor. Two 

break locations, i.e. on the suction and the discharge side of a recirculation pump, are evaluated, 

with high pressure core injecting (HPCI) and low pressure core spraying (LPCS) available 

whereas low pressure core injecting (LPCI) failed. The TRACE code is used for the analysis. 

The Chinshan TRACE model was benchmarked against steady-state and transient data 

contained in the plant FSAR report, as well as start-up data and the transient results using the 

RETRAN code. The thermal hydraulic phenomena in the lower plenum area and the jet pumps 

are also analyzed. 
xiv

 
xv

ABBREVIATIONS

ADS Automatic Depressurize System

CCFL Counter-Current Flow Limitation

CAMP Code Applications and Maintenance Program

CHAN A Component of TRACE to Simulate Fuel Bundles 

CS Containment Spraying 

DEG Double-Ended Guillotine

ECCS Emergency Core Cooling System

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

LPCI Low Pressure Core Injecting 

LPCS Low Pressure Core Spraying

MSIV Main Steam Isolation Valve

NPP Nuclear Power Plant

PCT Peak Cladding Temperature 

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. 

The most limiting LBLOCA for a BWR/4 reactor is the DEG break on the recirculation loop, 

which is 100% break flow on each side. In this paper, two break locations, i.e. on the suction and 

discharge side of a recirculation pump are evaluated, with HPCI and LPCS available whereas 

LPCI failed. The TRACE code is used for the analysis. The CCFL model plays a key role in any 

LBLOCA analysis since it affects the calculated discharge flow, reflooding time and peak 

cladding temperature (PCT). The current CCFL researches focus on a PWR reactor [1][2], a 

phenomenon review [3] or a code investigation [4]. In this report, a sensitivity study on the CCFL 

model is performed, by modelling a LBLOCA occurring at the Chinshan BWR/4 plant. The 

scenario assumes 102% power and 75% core flow, with a DEG break on the recirculation loop. 

10 CFR 50 Appendix K requires that any effect of fuel rod flow blockage calculated to occur 

during blowdown as a result of cladding swelling or rupture. In the events of DEG LOCAs, 

emergency core cooling system (ECCS) water is injected into the core through the upper plenum, 

lower plenum, or both. Of particular interest during the reflood period of a LBLOCA, is CCFL and 

pool formation above the upper core plate following core safety injection. Water that stagnates in 

the upper plenum and that is held there by CCFL cannot drain into the core and contribute to 

core cooling. When the CCFL occurs, the mass and heat transfer reduce between the gas and 

liquid phase, and a water pool forms above the top of fuels. This phenomenon reduces the 

injecting ECCS water in core area and results in a rapid increase of the fuel temperature. In 

Damerell’s report [5], analysis of the CCFL occurrence is essential to the safety of nuclear 

reactors. In Issa’s paper [3], the CCFL model is applied in the core area. 

Chinshan nuclear power plant, the first nuclear power plant in Taiwan, operates at a thermal 

power of 1828MWt after the measurement uncertainty recovery project in 2008. The rated steam 

flow is 3.5 Mkg/h with the rated core flow at 24.0 Mkg/h and the reactor pressure at 6.98 MPa. In 

order to validate the numerical model and to initialize the LOCA simulation, 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 the following parameters have been 

benchmarked with the steady state of FSAR [6], the start-up data [7], the transient results of the 

RETRAN data [8], and the vendor’s data [9][10]: 1. SEPD components adopted for separators 

and dryers and benchmarked with the 83% power 75% flow Turbine Trip test and 100% power 

100% flow Load Rejection test [11][12], 2. The CHAN components are adopted for fuel bundles 

with a reactivity feedback and benchmarked with the inadvertent startup of HPCI test. [13], 3. 

The behaviors of the break flow, core flow, reactor water level and reflood time [14][15].
1-2 

 
2-1 

2. METHODOLOGY AND MODELING 

2.1 Description of LBLOCA 

The LOCA is a hypothetical accident described in 10CFR50.46 that consists of a loss of reactor 

coolant resulting from breaks in reactor coolant pressure boundary piping including a DEG break 

on the largest pipe. The broken pipe is postulated to occur inside the primary containment and 

upstream of the first isolation valve, which is a general test criteria described in 10CFR50 

Appendix A. 

The LOCA is generally separated into three phases: the blowdown phase, the refill phase, and 

the reflood phase. The duration of each phase is strongly dependent upon the break size and 

location. In the blowdown phase, the reactor pressure and the discharging flow are high and 

then reduce very soon with loss of reactor water level. During the blowdown phase of a LOCA, 

the decrease of coolant inventory will result in an increase in fuel cladding temperature due to 

core flow degradation, and for the larger breaks, the core becomes fully or partially uncovered. 

The reactor pressure and water level decrease rapidly during the blowdown phase. The safety 

relieve valve (SRV) and automatic depressurize system (ADS) are not activated in LOCA 

because the primary coolant system is depressurized rapidly during the blowdown phase. During 

the early phase of the depressurization, the exiting coolant provides core cooling. The HPCI and 

LPCS injecting water also provide some heat removal. The end of blowdown is defined to occur 

when LPCS reaches rated flow. 

In the refill phase of a LOCA, the ECCS is functioning and there is a net increase of coolant 

inventory. During this phase, the core sprays provide core cooling and, along with LPCI, supply 

liquid to refill the lower portion of the reactor vessel. In this paper, the LPCI is assumed to be 

failed and only the HPCI and LPCS are available. During the refill phase, high steam pressure in 

the core region will prevent the penetration of ECCS liquid down into the reactor vessel. This 

phenomenon is called counter-current flow limitation (CCFL), and results in longer reflood times 

and consequently higher PCTs. In this stage, the heat removal is mainly by steam; thus the 

heat transfer from the core to coolant is less than the fuel decay heat rate, which results in 

continuous increase of fuel cladding temperature during the refill phase. 

The reflood phase begins when the core is being reflooded with ECCS water and the mixture 

level reenters the core region. During reflood, cooling above the mixture level is provided by 

entrained reflood liquid and below the mixture level by pool boiling. Large axial variations of wall 

temperature and of heat fluxes occur in a very narrow region close to the quench front, which 

results in steep axial temperature gradients. Steam generated mainly at the quench front 

entrains water slugs, fragments, and droplets upward. The cladding quenches, very rapidly 

cooling to the saturation temperature as the cladding surface becomes wetted. The cladding 

temperatures were observed to turn around very shortly after the onset of reflood [9][10][16]. 

2.2 Chinshan TRACE Model

The TRACE model of Chinshan Nuclear Power Plant (1) is developed, 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 (Figure 1). The 

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

azimuthally divided sectors. There are 408 fuel bundles and 130 separators. The 408 fuel 

bundles are modeled by the CHAN component, a specific TRACE component which is located in 

axial level 4 and 5 with point kinetics and dynamic reactivity feedback. The separators and 

dryers are simulated by the SEPD component which is located from axial level 7 to 10. The 

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

pumps in each, and four steam lines with one SRV, one main steam isolation valve (MSIV), one 

turbine control valve (TCV) and one turbine bypass valve (TBV) in each line. The HPCI injects 

coolant into the upper downcomer and flow through the lower plenum to the core area; the LPCS 
2-2 

water is injected in the upper plenum area. The HPCI provides the injection mass flow at 267.8 

kg/sec. The LPCS provides the coolant directly on the top of fuels with the mass flow at 472.2 

kg/sec per loop. 

The specific VESSEL component of TRACE code is able to model a more practical reactor 

vessel and its associated internals. The component is 1-, 2-, or 3D in a cartesian or cylindrical 

geometry and uses a six equation, two-fluid model to evaluate the flow through and around all 

internals so that more thermal-hydraulic phenomena can be analyzed. The internals of a reactor 

vessel can be modeled practically by the 3D VESSEL component including the downcomer, 

fuel-assembly, core area, core bypass flow, and upper and lower plenum. Modeling options and 

features incorporated into the VESSEL component are designed mainly for LOCA analysis, but 

the VESSEL component can be applied to other transient analyses as well. The energy 

transportation between the fluid and the reactor structure is important for LOCA analysis and is 

modeled by the heat-structure component of TRACE code. Structures that can exchange heat 

with the fluid in a reactor vessel include downcomer walls and the support plate and they are 

modeled as slabs and vessel rods. The convective heat transfer from any structure into the 

different fluid phases is modeled using Newton’s law of cooling to represent the energy 

exchange rate between the structure and the fluid phase [17]. The energy equations of the 

different fluid phases for various heat-transfer regimes are included in the TRACE code to 

provide better correlation of the heat transfer coefficient (HTC) between the wall and fluid. Heat 

transfer from the fuel rods and other structures is calculated using flow-regime-dependent HTC 

obtained from a generalized boiling curve based on a combination of local conditions and history 

effects. Inner- and/or outer surface convection heat-transfer and a tabular or point-reactor 

kinetics with reactivity feedback volumetric power source can be modeled. The coupling 

algorithm is implicit in terms of the wall temperature, liquid, vapor, and saturation temperatures 

and explicit in terms of the heat transfer coefficient (1). For each new-time step, the wall HTCs 

( h ) of a given structure are evaluated using the surface wall temperatures (T w ) and the fluid 

conditions obtained for the last time step. The new-time fluid-dynamics and conduction 

equations are solved using these HTCs, the new-time surface temperatures, and the new-time 

fluid and saturation temperatures where the total energy flux between the heat structure surface 

and the fluid cell can be written as 

( ) ( ) ( )

+1 +1 +1 +1 +1 +1 +1

= − + − + −

n

sat

n

w

n

sat

n

g

n

w

n

g

n

l

n

w

n

l

n Qtotal h T T h T T h T T

 (1) 

The term ( )

+1 +1

−

n

sat

n

w

n

hsat T T

 represents heat flux to the liquid that goes directly to boiling and 

( )

+1 +1

−

n

l

n

w

n

hl T T

 is the liquid heat flux that contributes to heating of the liquid. Since both the heat 

structure and fluid use the same heat flux term at each time step, energy is conserved for the 

coupling between the conduction and flow equations.

2.3 Pressure Drops of Reactor Vessel and Recirculation Loop

The flow resistances in the reactor vessel and the recirculation loop are crucial in the LOCA 

analysis and closely related to the break flow rate, the pressure transient and the reflood time 

[16]. The pressure distributions in the vessel are validated with plant data, including the lower 

plenum, fuel support plate, fuel zone, separators, dryer and dome area [12]. The pressure 

distribution in the recirculation loop including the recirculation pump is also validated. (Figure 2) 

The recirculation system provides forced flow through the reactor core for better cooling 

efficiency and to be one of the power modulating methods of a BWR reactor. There is one 

recirculation pump and ten jet pumps of each recirculation loop. The suction of the recirculation 

loop is connected to the lower downcomer and is pressurized by the recirculation pump as a 

driven flow of the jet pumps which is about 40% of the core flow. The recirculation pump adopts 

a TRACE built in homologous pump which is based on the standard homologous-curves 

approach. These curves represent the performance of the pump in a normalized format, giving 

the normalized pump head as a function of the normalized volumetric flow and normalized pump 
2-3 

speed. Homologous curves (one curve segment represents a family of curves) are used for this 

description because of their simplicity. These curves describe, in a compact manner, all 

operating states of the pump obtained by combining positive or negative pump-impeller angular 

velocities with positive or negative fluid volumetric flows. Each recirculation pump includes a 

motor-generator set which provides excess moment force and core flow to mitigate the impacts 

while trips. The recirculation pump is modeled by a centrifugal pump which calculates the 

pressure differential across the pump impeller and the pump impeller’s angular velocity as a 

function of the fluid flow rate and fluid properties including two-phase effect. In this paper, the 

homologous pump is used with a two phase flow with the pressure and flow rate validated. 

When reactor water level is below level-2, the recirculation pump will be tripped and cause the 

decreasing of core flow. After the recirculation pump trips, the pump speed will decrease with a 

ramp rate at 10~13 seconds. 

Since the breaks in this paper are located in the recirculation loop, the pressure distribution 

along the loop is important for analyzing the discharge flow. The flow resistance in the 

recirculation loop is critical in LOCA analysis and is closely related to the break flow rate and the 

pressure transient. The pressure distribution in the recirculation loop, including the recirculation 

pump, is validated with plant data as shown in Figure 2. The elevation and the flow resistance 

along the recirculation loop are built based on design data. The recirculation loop suction 

connects to the lower downcomer area. The upstream of recirculation pumps is the lowest point 

of the recirculation loop. The break flow at the recirculation pump suction is larger than the one 

at the discharge side because of the higher pressure and lower flow resistance. When coolant 

flows through the recirculation pump, the pressure is increased. The pressure peaks when the 

coolant flows through the nozzle of the jet pumps which have an abrupt cross area. 

2.4 The CCFL Model of the TRACE Code

10 CFR 50 Appendix K requires that any effect of fuel rod flow blockage calculated to occur 

during blowdown as a result of cladding swelling or rupture. The restrictions of Appendix K 

generally cause ECCS evaluation models to predict the PCT to occur during the reflood phase 

of the LBLOCA. Best estimate codes, however, have generally shown that the PCT occurs 

during the blowdown phase because of the improved understanding of the various reflood 

phenomena. In the event of a LOCA [16], ECCS water is injected into the core through the upper 

plenum, lower plenum, or both. Of particular interest during the reflood period of a LBLOCA, is 

CCFL and pool formation above the upper core plate following core safety injection. The ECCS 

water that is stagnated in the upper plenum and held there by CCFL cannot drain into the core 

and contribute to core cooling. Water flows downward from the upper plenum into the core under 

the influence of gravity. This downward flow is opposed by steam flow upward from the core. 

When the CCFL occurs, the mass and heat transfer reduce between the gas and liquid phase, 

and a water pool forms above the top of fuels. As ECC water enters the hot core, some of it will 

turn into steam. The increase in pressure tends to suppress the reflooding velocity in the core. 

This retarding effect of steam on the reflood rate is referred to as steam binding. This 

phenomenon reduces the injecting ECCS water in core area and results in a rapid increase of 

the fuel temperature [5]. Analysis of the CCFL occurrence is essential to the safety of nuclear 

reactors [3]. The CCFL behavior in BWRs is important because steam generation in the fuel 

bundles can restrict the emergency core cooling spray water from entering the top of the bundles. 

In LBLOCA, the reactor water level reduces rapidly because of its massive discharging flow, and 

the HPCI is not sufficient. In this paper, only the LPCS is assumed to be available and sprays 

coolant from the top of fuels. During the refill phase of a LOCA, the counter-current flow would 

occur at the top tie plate and the core support plate in both the BWR and PWR reactor vessels. 

When the LPCS sprays coolant into the top of fuels, the coolant forms a pool on the top of core 

because the pressure in the core area is high enough to prevent liquid flow down. This 

phenomenon is called the countercurrent flow limitation (CCFL), where only penetration liquids 

get into the core area. The cladding temperature increases rapidly when CCFL occurs because 

there is little mass and heat transfer between liquid and steam. When CCFL occurs, the ECCS 
2-4 

coolant would be prevented from penetrating down to the fuel zone which why this is called the 

ECCS bypass model. 

The CCFL is defined when the downward flowing liquid in a countercurrent flow is about to 

change its direction and flow upward. This condition can be obtained if the steam flow rate is 

increased while keeping the liquid flow rate constant in a countercurrent flow. Depending upon 

the geometry of the equipment used, the occurrence of the CCFL could vary. In reactor 

applications, CCFL can occur at flow area restrictions when liquid flows downward through rising 

gas or vapor. For example, in the tie-plate region during reflood, the upward flow of steam can 

prevent or limit the down flow of liquid. Since the TRACE code is well designed for the LOCA 

analysis, both categories of CCFL phenomena, the single pipe and the perforated plate, are 

adopted and validated. The TRACE code is capable of calculating the CCFL and predicting the 

CCFL by resolving the conservation equation of the CCFL model. The TRACE code models four 

principle regimes, the three bubbly/slug flow regimes and one annular/mist flow regime, on 

the vertical and horizontal flows. The CCFL correlation of the TRACE code can be applied in its 

3D component and 1D component. The CCFL correlations describe a superficial liquid mass flux 

in down-flow versus a superficial gas mass flux in up-flow which are represented as three types, 

the Wallis type (6) [18], the Kutateladze type (7) [19] and the Bankoff type (8) [20]. The CCFL 

model of the TRACE code adopts the three correlations and is verificated [4][21]. 

g W f Cw

j +m j =

* 2/1 * 2/1

 (6) 

Kg +mK Kf = CK

∗ 2/1 ∗ 2/1

 (7) 

Hg +mBH f = CB

2/1* 2/1*

 (8) 

Where m and C are constants determined by experiments, 

∗

k

j

, 

∗ Kk and ∗ Hk

 are 

dimensionless superficial velocity, k is liquid or gas phase. 

( )

2/1

*

















−

=

f g

k

k k

gd

j j

ρ ρ

ρ

 (9) 

( )

4/1

2

















−

=

∗

f g

k

k k

g

K j

σ ρ ρ

ρ

 (10) 

( )

2/1

*

















−

=

f g

k

k k

gw

H j

ρ ρ

ρ

 (11) 

Where k

j

and ρ k are the superficial velocity and the density of phase k , d is the hydraulic 

diameter, σ is the surface tension, and w is the interpolative length scale.
2-5 

Figure 1 The TRACE model of Chinshan nuclear power plant 
2-6 

Figure 2 Pressure distributions in recirculation loop
3-1

3. RESULTS AND DISCUSSION 

When DEG break on recirculation loop begins, reactor water level decreases fast because of 

massive discharging flow. In this report, Table 1 shows the initial conditions and the reactor is 

assumed to scram by the signal of reactor low water level (Level-3) that occurs in less than one 

second soon after the DEG break begins. In Figure 3, the break flow is large in the first twenty 

seconds and most of the coolant in the reactor inventory is discharged. In the first few seconds, 

the break flow at the recirculation pump discharge peaks higher than at the suction because the 

recirculation pump is still running, but it decreases rapidly after the recirculation pumps trip at 

reactor water level 2. After the closure of MSIVs, the different discharging flow can be noted 

from a peak of reactor pressure at the recirculation pump discharge (Figure 4). The reactor 

pressure at the recirculation pump suction does not increase because of its larger break flow. 

When the pressure in lower plenum decreases, a flash phenomenon occurs and results in a 

peak of core flow (Figure 5) and cladding temperature decrease (Figure 6). The flash 

phenomena occur earlier and more significant at the recirculation pump suction because of its 

larger discharging flow. The flash at the recirculation pump suction occurs earlier at 8.5 seconds 

while the one at the discharge occurs at 15.3 seconds (Table 2). From Figure 5 and 6, the core 

flow is found to be a main effect on the cladding temperature. When the DEG break occurs, the 

cladding temperature increases immediately because of the massive break flow, fast decreasing 

core flow and reactor water level. The sudden increase of core flow impacted by the flash 

phenomenon provides a better cooling effect and results in the decrease of cladding 

temperature (Figure 6). Although the cladding temperature decreases for a while, the massive 

discharging coolant will result in continuous loss of coolant inventory and consequently the 

increasing cladding temperature. 

When the DEG break occurs, the reverse flow is found in the path from the lower plenum to 

downcomer through jet pumps and becomes a part of break flow. The core flow decreases 

rapidly when the DEG breaks occurs and the reverse flow is observed as a negative value of the 

jet pumps exit flow in Figure 7. The jet pumps exit flow transfers from positive to negative, and 

then turns to zero after most of the coolant having been discharged. The jet pumps uncover and 

recirculation loop uncover occur in less than twelve seconds, and it is earlier at the recirculation 

pump suction. The coolant flashes in lower plenum soon after the recirculation pipes uncover. 

When the flash occurs, a sudden pressure increase in the lower plenum will drive out the coolant 

in the lower plenum through the core area, jet pumps and downcomer. In Figure 8, the jet pumps 

suction flow also peaks up when the flash occurs. 

In this report, LPCI and containment spraying (CS) are assumed to be failed. The MSIV closes 

and the recirculation pump trips at level-2. After the closure of MSIVs, the main steam flow stops. 

The HPCI is initiated at the signal of reactor water level-2 and injects coolant into the vessel via 

the feedwater line B. The LPCS is initiated at the signal of reactor water level-1 and injects 

coolant directly on the top of active fuels (TAF). The cladding temperature increases rapidly 

because of the fast decrease of core flow. Since the break flow is larger than the HPCI injecting 

flow, reactor water level keeps decreasing until the LPCS injected. At the beginning of LPCS 

injection, the reactor water level is still below the fuel support plate and the increasing cladding 

temperature keeps increasing. In the mean time, the steam fills the whole fuel zone with high 

pressure that prevents the LPCS injecting water from flowing into the fuels bundles and forms a 

pool on the top of fuels. The coolant will penetrate or dump into the fuel zone when the static 

pressure of the pool is greater than the steam pressure. The phenomenon is called CCFL. In 

this stage, the conductive heat transfer from liquid to steam is not efficient, resulting in a higher 

cladding temperature. With the more coolant penetrating or flowing to the fuels zone, more 

thermal heat is removed and the steam pressure is reduced, and the lower plenum begins to 

refill. But fuel temperature will keep increasing until reflood. 
3-2

3.1 The Break Flow of the Recirculation Loop 

The DEG breaks at the recirculation pump suction and the discharge are analyzed. The break 

flow at the recirculation pump discharge is less because the flow resistance of the recirculation 

pump reduces the break flow (Figure 3), which results in a slower decline of reactor pressure 

(Figure 4) and a later decline of reactor water level (Figure 9). The effect with a lower resistance 

is more significant in the first few seconds while the break flow is large. From Figure 9, the 

duration of the fuel uncover is longer at the recirculation pump suction and results in higher PCT 

(Figure 6). The break flows with and without the CCFL model are almost the same. Several data 

are about the same cures with and without the CCFL model in the first 40 seconds from Figure 3 

through Figure 13, which implies that the CCFL model only affect the reflood time and PCT. 

3.2 The Transient Curves of Core Flow 

In Figure 5, the core flow decreases rapidly soon after the DEG break begins. The fast 

decreasing core flow results in a rapid increase of cladding temperature. The coolant in the 

lower plenum becomes reverse-flow to the downcommer through jet pumps as the negative 

values of the jetpump exit flow and discharge flow indicated in Figure 7 and 8. In addition, the 

reverse-flow is also found in the driven flow of jet pumps at the broken loop, but it is smaller 

because of its higher elevation and flow resistance of nozzles. Because of the massive 

discharging flow, the suction end of jet pumps uncovers in 2.5 and 5.3 seconds at the 

recirculation and the discharge, respectively. The recirculation pump trips at the signal of reactor 

water level-2. Because the discharging flow on the recirculation loop is massive, the reactor 

water level decreases very fast. 

Because the break flow at the recirculation pump suction is much larger than at the discharge 

(Figure 3), the reverse-flow is more significant at the former one. In Figure 10, the reverse-flow 

at the driven nozzle of jet pumps is found smaller at the recirculation pump suction because its 

driven nozzle connects to the recirculation pump with flow resistance added . 

In Figure 5, the core flow peaks up in both the DEG breaks at the recirculation pump and the 

discharge because the coolant in the lower plenum flashes and results in a peak of pressure. 

The flash phenomenon can be evaluated through Figure 11 when the steam flow in lower 

plenum increases suddenly and then peaks up and down. The flash phenomenon occurs earlier 

and more significantly at the recirculation pump suction because its water level in the lower 

plenum reduces faster (Figure 12). 

3.3 The Curve of Peak Cladding Temperature 

The cladding temperature is found to be closely related to core flow at the beginning of LOCA. In 

Figure 6, the cladding temperature peaks up fast while the core flow decreases rapidly. The core 

flow reduces rapidly because of the massive break flow and the trip of recirculation pumps. In 

Figure 5, the core flow peaks at the recirculation pump suction while a rapid cladding 

temperature decrease is observed in Figure 6. The rapid temperature decrease is not observed 

at the recirculation pump discharge. It is the reason that the first peak of the cladding 

temperature at the recirculation pump discharge is higher than at the suction in Figure 3 

although its break flow is less? After the first peak of cladding temperature, the temperature 

decreases for a while because of the sudden peak of core flow caused by the flash phenomena 

in lower plenum (Figure 5). In Figure 6, the peak cladding temperature at the recirculation pump 

suction is lower and decreases earlier than at the discharge. This can be seen in Figure 5 with 

an earlier peak of core flow at the recirculation pump suction. The core flow peaks up because 

of the flash phenomena in lower plenum. In Figure 5, the core flow decreases fast soon after 

the break occurs and results in rapid increase of PCT in Figure 6. When the reactor pressure 

and water level decrease causing by massive break flow, the coolant flash occurs in the lower 

plenum and results in a peak of core flow (Figure 5) and the flash phenomenon is more 

significant at the recirculation pump suction. 
3-3

In addition, the HPCI is initiated at the signal of reactor water level-2 and it provides excess 

coolant to the vessel at 265.5 kg/sec via the feedwater line B which is also helpful to reduce the 

cladding temperature. Since the HPCI flow is still less than the break flow, the reactor water level 

will not recover before LPCS injection. In this stage, the transient curve of cladding temperature 

is found closely related to reactor water level after the sensitivity studies with the core flow and 

scram time. When comparing Figure 6 with Figure 9, the PCT is found to be closely related to 

the recovery of reactor water level. The higher break flow at the recirculation pump suction will 

result in a faster decline and lower reactor water level with faster increase and higher peak 

cladding temperature. The time to reflood at the recirculation pump suction is longer than at the 

discharge (Figure13). The later recovery of reactor water level will result in a higher PCT. The 

PCT at the recirculation pump discharge is the lowest one among the four tests and is the 

highest one at the recirculation pump suction with CCFL model. 

3.4 The Impact of CCFL Effect 

In Figure 6, the second peak of cladding temperature with CCFL effect is found to be higher than 

without CCFL effect because less ECCS water penetration provides less cooling in the fuel zone. 

Although the cladding temperature decreased slightly after the sudden increase of core flow, it 

keeps increasing because of the continuing loss of coolant inventory. In this paper, the CCFL 

model of the core zone is analyzed. The impact on the steam pressure of the fuel zone during 

the refill phase is found to be critical for the reflood time and PCT. The ECCS water will 

penetrate into the fuel zone only when the static pressure of the pool is higher than the steam 

pressure in the fuel zone. The more coolant penetrating into the core area results in a more heat 

removal and a faster temperature decline. In Figure 9 and 12, although the HPCI and LPCS are 

initiated almost at the same time in the four test cases, a later recovery of reactor water level is 

found with the CCFL model because the LPCS flow is blocked and forms a pool above the top of 

fuels. The curve of reactor water level decline at the beginning of LOCA is found not affected by 

the CCFL effect. The reason is that the reactor water level decline is mainly related to the break 

flow. 
3-4

Figure 3 Comparison of the total break flow 

Figure 4 Comparison of reactor pressure 
3-5

Figure 5 Comparison of the core flow 

Figure 6 Comparison of peak cladding temperature 
3-6

Figure 7 Comparison of the jet pumps exit flow rate at the broken loop 

Figure 8 Comparison of the jet pumps suction flow rate at the broken loop 
3-7

Figure 9 Comparison of the reactor water level 

Figure 10 Comparison of driven flow rate of jet pumps at the broken loop 
3-8

Figure 11 Comparison of the steam flow in lower plenum 

Figure 12 Comparison of the water level in lower plenum 
3-9

Figure 13 Comparison of the liquid flow of fuel support

Table 1 The Initial conditions for the LBLOCA of Chinshan NPP 

Parameters 102%POWER 

75% FLOW

Thermal Power (MW) 1864 

Steam Flow (Mkg/h) 3.54 

Core Flow ( Mkg/h ) 18 

Feedwater Flow (Mkg/h) 3.54 

Recirculation Loop Flow (Mkg/h) 3.95 

Dome Pressure (MPaG) 6.98 

Middle Core Pressure (MPaG) 7.09 
3-10

Table 2 Event Times for LBLOCA of Recirculation Loop of Chinshan NPP

Value(Seconds)

Event/Setpoint Discharge 

DEG

Suction 

DEG

Discharge 

DEG(CCFL)

Suction 

DEG(CCFL)

Initial break 0 0 0 0

Initiate Scram(reactor water level 3) 0.5 0.3 0.5 0.3 

Reactor water level 2 2.1 1.0 2.1 1.0

Steam flow stop(MSIV closed) 2.4 1.3 2.4 1.3

Reactor water level 1 4.1 3.0 4.1 3.0 

Jet pump suction uncover 5.3 2.5 5.3 2.5

Recirculation Pipe uncovers 11.4 6.0 11.4 5.8 

Lower Plenum Flashes 15.3 8.5 15.3 8.0

Recirculation Isolation Valve Begins to 

Close 

17.6 7.3 17.6 7.1 

HPCI Flow Starts 22 21 22 22

LPCS Flow Starts 54 53 54 53

Recirculation Isolation Valve Fully Close 64 54 64 54 

Start Reflooding 98 99 128 135

PCT Reached (Temperature, K) 17 (883K) 133 (1035K) 133 (950K) 155 (1081K)

Water Above BAF 144 223 204 233

Water Above TAF 246 267 252 227
4-1

4. CONCLUSIONS

With the CCFL effect considered, the impacts on the steam pressure in the fuel zone during the 

refill and reflood phase are critical for the time to reflood and PCT. Although the HPCI and LPCS 

are initiated almost at the same time, the time to reflood at the recirculation pump suction takes 

longer than at the recirculation pump discharge because the later with a flow resistance of 

recirculation pump, which results in a less break flow. The CCFL effect does not affect the break 

flow but closely related to the reflood time. The time to reflood with the CCFL model also takes 

longer than without the CCFL model because the ECCS water forms a pool above the fuels with 

less cooling, which results in a much higher PCT. 

The break flow at the recirculation pump suction is larger than at the recirculation pump 

discharge because of the flow resistance of recirculation pumps, which affect the break flow 

more at the beginning of DEG LOCA. The break flow rate is massive at beginning that the flow 

resistance of recirculation pumps plays an important role and results in a higher break flow at the 

recirculation pump suction. The break flow at the recirculation pump suction is larger, which 

results in a faster decline of reactor water level and pressure. The cladding temperature peaks 

immediately at the beginning of LOCA because of the fast decreasing core flow. The cladding 

temperature decreases for a while because the sudden peak of core flow, which is caused by 

the flash phenomena, provides a better cooling. But the reactor water level is found not to 

recover without LPCS injection that the cladding temperature will keep increasing before the 

core is reflooded. With the Chinshan TRACE model adopted, more LOCA events will be 

analyzed in the future. The sensitivity study on the pellet-cladding gap is the next planned 

analysis. 
4-2

 
5-1

5. REFERENCES

1. Takeuchi, K. et al., 1998, Analyses of subcooled CCFL tests for evaluation of 

WCOBRA/TRAC applicability, Nuclear Engineering and Design, p.p. 127-140. 

2. Jeong, Hae Yong, 2002, Prediction of counter-current flow limitation at hot leg pipe during a 

small-break LOCA, Annals of Nuclear Energy 29, p.p. 571–583. 

3. Issa, A. Al, Macian, R., 2011. A review of CCFL phenomenon, Annals of Nuclear Energy 38, 

p.p. 1795–1819. 

4. Cheng, Yi-Hsian et al., 2010. An investigation of steam-water countercurrent flow in TRACE, 

Annals of Nuclear Energy 37, p.p. 1378–1383. 

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NUREG/IA-0127. 

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Energy Council, R.O.C. 

9. AREVA, 2001. Realistic Large Break LOCA Methodology, AREVA Document 

EMF-2103(P), revision 0. 

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EXAM BWR-2000 ECCS Evaluation Model, AREVA Document EMF-3109 (P) , revision 0. 

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

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18. Wallis, G.B. 1969. One Dimentional Two Phase Flow, New York: McGraw-Hill. 
5-2

19. Takahashi, T., Akagi, Y., Kisimoto, T., 1979. Gas and liquid velocities at incipient liquid 

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countercurrent flow through multiple short path, International Journal of Heat and Mass 

Transfer 25(9), pp. 1301-1312. 



UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA

-0437 Sensitivity Study of the DEG LBLOCA Transient on the

 Counter-Current Flow Limitation by Using TRACE