NUREG/IA-0171 

International 

Agreement Report 

Simulation of LOCA 6" and 

LOCA 2" Transients 

in the RHR of a PWR 

Under Low Power Conditions 

Using RELAP5/MOD3.2 

Prepared by 

S. Martorell, S. Carlos, G. Verdfi, V Serradell, J. Rodenas, R. Mir6 ,C. Llopis* 

*C.N. Vandellos II 

Travessera de les Corts, 55, Lateral 

08028 Barcelona 

Department of Chemical and Nuclear Engineering 

Polytechnic University of Valencia 

Camino de Vera, 14 

Spain 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

February 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0171

International 

Agreement Report 

Simulation of LOCA 6" and 

LOCA 2" Transients 

in the RHR of a PWR 

Under Low Power Conditions 

Using RELAP5/MOD3.2 

Prepared by 

S. MartoreU, S. Carlos, G. Verdid, V Serradell, J. Rodenas, R. Mir6 ,C. Llopis* 

*C.N. Vandellos II 

Travessera de les Corts, 55, Lateral 

08028 Barcelona 

Department of Chemical and Nuclear Engineering 

Polytechnic University of Valencia 

Camino de Vera, 14 

Spain 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

February 2000 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
ABSTRACT

The present study consists of the simulation of two loss of coolant accidents, 

LOCA 6" and LOCA 2", in one of the residual heat removal system (RHR) lines outside 

the containment, using the thermal-hydraulic code RELAP5/MOD3.2. 

Both transients have been simulated on a typical three loop, Westinghouse 

design, pressurized water reactor plant working under shutdown conditions. 

The study was focused on the simulation of the most important thermal

hydraulic parameters in order to check the validity of the success criteria assumed in the 

plant probabilistic safety analysis (PSA) under shutdown conditions. Also to analyze the 

code capability for simulating shutdown conditions was of interest in this study. 

As a result of this study, it can be concluded that the main thermal-hydraulic 

plant features follow what is foreseen in the plant PSA, although it can not be assured 

that the values reached are the correct ones due to the lack of experimental data.

iii
TABLE OF CONTENTS 

A B STR A CT .............................................................. iii

LIST OF TABLES ............................ 

LIST OF FIGURES ............................ 

EXECUTIVE SUMMARY ...................... 

1. INTRODUCTION .......................... 

2. PLANT AND TRANSIENT DESCRIPTION ..... 

2.1. PLANT DESCRIPTION ................. 

2.2. TRANSIENT DESCRIPTION ............. 

3. CODE INPUT AND MODEL DESCRIPTION ... 

3.1. SYSTEM DESCRIPTION ............... 

3.2. INITIAL CONDITIONS ................. 

3.3. LOCA 6" SIMULATION ................ 

3.4. LOCA 2" SIMULATION ................ 

4. RESULTS ................................ 

4.1. LOCA 6" TRANSIENT RESULTS ......... 

4.2. LOCA 2" TRANSIENT RESULTS ......... 

4.3. SENSITIVITY STUDIES ................ 

5. RUN STATISTICS ....................... 

6. CONCLUSIONS AND RECOMMENDATIONS...

vii 

vii 

ix 

.1 

.3 

.3 

4 

7 

7 

9 

10 

11 

13 

13 

14 

16 

17 

19

7. REFERENCES ......................... .............................. 2 1

V
LIST OF TABLES: 

TABLE 1: MAIN PLANT CHARACTERISTICS ................................. 23 

TABLE 2: INITIAL VALUES ................................................ 24 

TABLE 3: RUN STATISTICS ................................................ 25 

LIST OF FIGURES: 

FIGURE 1: VANDELLOS II Nodalization for RELAP5/MOD3.2 Under

FIGURE 2: 

FIGURE 3: 

FIGURE 4: 

FIGURE 5: 

FIGURE 6: 

FIGURE 7: 

FIGURE 8: 1 

FIGURE 9: 

FIGURE 10: 

FIGURE 11: 

FIGURE 12: 

FIGURE 13: 

FIGURE 14: 

FIGURE 15: 

FIGURE 16: 

FIGURE 17: 

FIGURE 18: 

FIGURE 19: 

FIGURE 20: 

FIGURE 21: 

FIGURE 22:

Shutdown Conditions ............................... 

LOCA 6" Transient Simulation .......................... 

LOCA 2" Transient Simulation .......................... 

Core Level for LOCA 6". A-5 Loop No.2: Coolant Temperature 

Core Outlet and Clad Temperatures for LOCA 6" .......... 

Break and CVC System Mass Flow for LOCA 6" ........... 

Reactor Coolant System Pressure for LOCA 6" ............ 

Hot Legs Mass Flow Rates for LOCA 6" ................ 

Cold Legs Mass Flow Rates for LOCA 6" .................. 

Reactor Coolant System Mass for LOCA 6" ............... 

Core Level for LOCA 2" .............................. 

Core Outlet and Clad Temperatures for LOCA 2" ......... 

Reactor Coolant System Pressure for LOCA 2" ........... 

Steam Generators Pressures for LOCA 2" ............... 

Steam Generators Levels for LOCA 2" ................. 

Hot Legs Mass Flow Rates for LOCA 2" .................. 

Cold Legs Mass Flow Rates for LOCA 2" ............... 

Pressurizer Level for LOCA 2" ......................... 

CPU Time for LOCA 6" .............................. 

Time Steps for LOCA 6" .............................. 

CPU Time for LOCA 2" ..............................

Time Steps for LOCA 2" ......

.. 26 

.. 27 

.. 28 

.. 29 

.. 30 

.. 31 

.. 32 

.. 33 

.. 34 

.. 35 

.. 36 

.. 37 

•. 38 

.. 39

.... 40 

.... 41 

.... 42 

.... 43 

.... 45 

.... 45 

... 46

.................................. 47

vii
EXECUTIVE SUMMARY

In this paper the simulation of two loss of coolant accidents of 6" and 2" of 

diameter in one of the RHR lines of a three loop pressurized water reactor working 

under shutdown conditions using the RELAP5/3.2 thermal-hydraulic code are presented. 

Although this code was primarily developed to study loss-of-coolant (LOCA) accidents 

with the system working at full power, what means high pressure, this study is focused 

on studying those transients with the plant working under low power or shutdown 

conditions, what means low pressure. 

The residual heat removal system is part of the emergency core cooling system 

(ECCS) in a nuclear reactor, which is responsible for removing the residual heat during 

low power and shutdown conditions. As the computer code RELAP5/MOD3.2 has 

already been validated for LOCA transients at full power conditions, it seemed of 

interest to analyze the capability and limitations of the code to simulate low power and 

shutdown conditions. Particularly, in order to investigate if RELAP5/MOD3.2 could 

reproduce the physical phenomena involved in a LOCA transient in one of the RHR 

lines when the plant works at low power or shutdown conditions. 

The main objectives of this study are the verification of the success criteria 

foreseen in the plant probabilistic safety analysis (PSA), and the evaluation of the 

RELAP5/MOD3.2 code capability for simulating transients under the established 

conditions. And also, to investigate if RELAP5/MOD3.2 could reproduce the physical 

phenomena involved in a LOCA transient in one of the RHR lines when the plant works 

at low power or shutdown conditions.

ix
The plant chosen for simulating the transient was Vandellos 11, which is a three 

loop pressurized water reactor plant, designed by Westinghouse, of 2775 Mwt of 

nominal thermal power. The plant is initially working in Mode 5 with the main reactor 

coolant system (RCS) average temperature lower than 93.3°C, the main reactor coolant 

system (RCS) pressure close to 24 kg/cm , a pressurizer level of 25% and the steam 

generators in wet layout. This situation is known as cold standby. 

The first transient simulated is a LOCA 6" in one of the lines of the RHR system 

outside the containment. It is supposed that inventory recovery is successful, which 

takes place by the charging lines through the normal charging way, being the maximum 

available time for human action 600 sec. after the break is initiated. 

The second transient consists of a LOCA 2" in one of the lines of the RHR 

system. It is supposed that the inventory recovery by the charging pumps fails, both 

through the normal charging way and through the safety injection. Nevertheless, the 

break can be isolated; having 600 sec. available after the break is initiated to take this 

action. The residual heat is removed using a steam generator inventory as the final heat 

sink, without being necessary neither the main nor the auxiliary feedwater systems. 

Neither for LOCA 6" nor for LOCA 2" transient any damage to the core is 

supposed in the plant PSA. Nevertheless, for LOCA 6" inventory of the refueling water 

storage tank needs to be recovered in order to assure long-term injection. For LOCA 2" 

the plant is expected to tend to a stable situation. 

With regard to the initial conditions, in the original plant nodalization at full 

power conditions, some changes were realized to model the RHRS in order to allow 

recirculation mode, as it is demanded in the initial conditions.

X
From both transients developments we have obtained the evolutions of the most 

interesting thermal-hydraulic variables. The calculations were run in a CONVEX SPP 

1000 owned by the University Polytechnic of Valencia. The CPU time consumed in 

both transient simulations has been considerable, due to the time steps needed for 

achieving a correct transient simulation, and to the large periods of time that had to be 

simulated in order to appreciate changes in the evolutions of the most important 

thermal-hydraulic parameters in order to analyze the events that take place. 

Once the study was finished we concluded that the results of both transient 

simulations confirm, satisfactorily, the success criteria foreseen in the plant PSA, 

although experimental data are needed to assure the values obtained are the correct ones 

and whether the code works properly under these conditions.

xi
1. INTRODUCTION

The RELAP5 is a thermal-hydraulic code widely used for studying transients in 

pressurized water reactors (PWR). This code was primarily developed to study loss-of

coolant (LOCA) accidents with the system working at full power, what means high 

pressure. 

Here the study is focused on studying a loss-of-coolant accidents, but with 

different conditions from the above mentioned. In this case, the plant is working under 

low power or shutdown conditions, what means low pressure, with the residual heat 

removal (RHR) system working in recirculation mode. 

The residual heat removal system is part of the emergency core cooling system 

(ECCS) in a nuclear reactor, which is required for removing the residual heat during low 

power and shutdown conditions. 

It is of great interest to study those transients that cause a loss of the RHR system 

under these working conditions, and also to check the alternative ways foreseen to 

evacuate the residual heat generated. 

As the computer code RELAP5/MOD3.2 version has already been validated for 

LOCA transients at full power conditions, it seemed of interest to analyze the capability 

and limitations of the code to simulate low power and shutdown conditions. 

Particularly, in order to investigate if RELAP5/MOD3.2 could reproduce the physical 

phenomena involved in a LOCA transient in one of the RHR lines when the plant works 

at low power or shutdown conditions. 

The ultimate objective of this study is to validate, considering the results 

obtained and the limitations encountered, the success criteria established in the 

probabilistic safety analysis (PSA) of the plant, for such operational modes.

I
This study is part of the code applications and maintenance program (CAMP), 

and is focused on studying some transients described in the PSA for low power and low 

pressure. It has been conducted by the Chemical and Nuclear Engineering Department 

of the Polytechnic University of Valencia (UPV), in collaboration with the Consejo de 

Seguridad Nuclear (CSN) of Spain. 

This document is organized as follows. In chapter two a brief plant description is 

presented, highlighting its most important features. The descriptions of the two 

transients simulated, LOCA 6" and LOCA 2", are also presented in this chapter. The 

input file used for reaching the transient initial conditions is explained in chapter three, 

together with other specific modifications needed in each transient development. 

Chapter four contains the results of both simulations, showing the evolutions versus 

time of some thermal-hydraulic variables considered as important. It also contains the 

discussions of the additional sensibility studies performed, and some recommendations 

on the aspects that should be subject of a further study. Chapter five gives information 

about the time steps utilized in both simulations and the time consumed in the transient 

developments. The conclusions derived from the whole study and the most important 

recommendations for other users are discussed in chapter six. References needed in the 

elaboration of this study are detailed in chapter seven. Finally, the tables and figures 

referenced in the text are shown at the end of the report.

2
2. PLANT AND TRANSIENT DESCRIPTION

2.1. PLANT DESCRIPTION 

The plant chosen for simulating the transient was Vandellos II, which is a three 

loop pressurized water reactor plant, designed by Westinghouse, of 2775 Mwt of 

nominal thermal power. The plant is equipped with three U-tube steam generators 

without preheaters, and uses the seawater as final heat sink. The reactor vessel is cold 

head type. In table 1, the most important plant features are presented. 

Figure 1 shows the nodalization diagram used in RELAP5/MOD3.2 for the plant 

under low power and low-pressure conditions. In this diagram there are three loops, with 

a steam generator each, a reactor vessel, and a pressurizer. The RHRS is also modeled, 

which includes the high-pressure coolant injection system (HPCIS) and the low-pressure 

coolant injection system (LPCIS), adapted for recirculation. 

The RHR system consists of two lines, A and B, with their low pressure coolant 

injection part connected with loops 1 and 2 of the main reactor coolant system (RCS) 

and with the refueling water storage tank (RWST). When the plant is in Mode 5, the 

RHR system extracts water from the hot legs of loops 1 and 2, which is recirculated by 

the RHR pumps towards the heat exchangers, to the cold legs of the three loops. In 

Mode 5, under the conditions required for initiating the transients, the RHR system 

recirculates a total mass flow of 300 kg/sec., using both pumps.

3
2.2 TRANSIENTS DESCRIPTION

Both studies consist of simulating a break, no guillotine, of 6" and 2" diameters 

respectively, in one of the RHR system lines of a PWR outside the containment, with 

the plant working in Mode 5. This situation is known as cold standby, with the 

following characteristics: 

- Main reactor coolant system (RCS) average temperature lower than 93.3°C. 

- Main reactor coolant system (RCS) pressure close to 24 kg/cm2. 

- Pressurizer level 25%. 

- Steam generators in wet layout. 

In order to obtain the above mentioned values, it was necessary to modify the 

level and pressure controls, adjusting them to the new working conditions. The initial 

conditions for the most important variables are presented in table 2. 

The procedure applied for LOCA 6" and LOCA 2" transient scenarios is the 

ARG-2 Westinghouse Shutdown LOCA, which main actions are: 

- Symptom or entry condition: decrease in pressurizer level. 

- RHR pumps trip. 

- Chemical and volumetric control system (CVC) discharge isolation. 

- Increase of charging flow or realignment of safety injection. 

- Break isolation attempt. 

- Refueling water storage tank (RWST) recovery. 

The transient developments are different depending on the break produced, due 

to fact that the sequence of the events of the success criteria foreseen in the plant PSA 

under these conditions is different in each case.

4
Thus, when a LOCA 6" in the RHR system outside the containment is produced, 

it is supposed that inventory recovery is successful, which takes place by the charging 

lines through the normal charging way, being the maximum available time for human 

action 600 sec. after the break is initiated. It is also supposed that the break can not be 

isolated and that the refueling water storage tank needs to be recovered in order to 

assure long term injection. 

After a LOCA 2", it is supposed that the inventory recovery with the charging 

pumps fails, both through the normal charging way and through the safety injection. 

Nevertheless, the break can be isolated; having 600 sec. available after the break is 

initiated to take this action. The residual heat is removed using a steam generator 

inventory as the final heat sink, without being necessary neither the main nor the 

auxiliary feedwater systems.

5
3. CODE INPUT AND MODEL DESCRIPTION.

3.1. SYSTEM DESCRIPTION 

The simulations have been run on a typical three loop pressurized water reactor 

full scaled model, figure 1, in which the low and high pressure injection systems are also 

modeled, and the latter has been modified to allow recirculation. 

It has been necessary to adapt the plant nodalization using RELAP5/MOD3.2 

from initial available full power conditions, to low power conditions, as this was the 

previous state of the plant before starting the transients. 

An important nodalization change has been the inclusion of the RHR system 

modification, so it could work in recirculation mode, since it is demanded in the initial 

conditions. By this reason, some volumes have been added, simulating the pipes that 

connect the hot legs of loops 1 and 2 of the primary system with the lines A (vols. 301, 

303) and B (vols. 321, 323) of the low pressure injection system respectively. Among 

these added volumes are the RHR heat exchangers (vols. 308, 309, 328, 329), which are 

able to remove all the residual power generated, providing the appropriated temperature 

to the water being recirculated to the loops. 

From the initial input deck for full power conditions, the volumes that simulate 

the turbine, the steam dump, the Main Steam Relieve Valves (MSRV) and the 

accumulators have been eliminated, since they are not used in the transients simulated, 

nor in other possible transients in low power conditions. 

The most important change in the reactor coolant system (RCS) was made in the 

pumps (vols. 118, 148, 178), for adjusting them to the new working conditions, as the 

characteristic curve was not prepared for simulating them when they are stopped. The 

solution was found by modifying the suction and discharge loss coefficients adjusting 

them until the mass flow rate through the RCS was the same as the measured in plant.

7
At the secondary side of the steam generators the water injection through the 

main and auxiliary feedwater systems has been deactivated, although it has not been 

removed to allow that they could be required in the simulation of future transients under 

low power conditions. 

In order to obtain reliable results from the transient developments, it is needed to 

simulate the break as accurately as possible. It has been rather complicated in this case 

since experimental data were not available to compare with. The breaks have been 

modeled as a TRIPVALVE (vol. 340) that connects one of the RHR lines with a 

TMDPVOL (vol. 334), which simulates the atmosphere. The valve diameters that 

simulates the breaks corresponds to 6" or 2" break respectively in a 12" sch 40 pipe. 

One of the most important characteristics of the valve is to adopt the abrupt area 

change model, using as discharge coefficients 1, 1.8, and 0.8 for subcooled, two phases 

and superheated regimen respectively, as it is suggested in the literature [3]. These 

values are more conservative than the ones suggested in the RELAP5/MOD3.2 code 

manual [12] [13], what will suppose an increase in the mass flow through the break. In 

fact, the change of the discharge coefficients will provide a more realistic value for the 

mass flow rate since the code underpredicts it for low quality flow, just as it occurs in 

our transient. 

The choking model also presents several problems in simulating mass flows 

rates through breaks under low-pressure conditions. The root cause of this misfunction 

seems to be the inability of the code to unchoke a junction even though its velocity is 

below the critical one for the flow conditions [5]. This happens when nonhomogeneous 

flow is activated, since this model produces unrealistic low mass fluxes at low 

pressures. The current recommendation given in the RELAP5/MOD3.2 code manual 

[12] is to use homogeneous and choked flow for break junctions and other connections 

to the atmosphere, what will produce mass fluxes close to the homogeneous equilibrium 

critical flow model. For internal junctions it is suggested to invoke nonhomogeneous 

model with the choking model turned off.

8
Finally, we have to mention that the valve that simulates the break is governed 

by a trip. It is initially closed and opens when the trip is activated, in both transients time 

is the variable that activates the trip. 

For LOCA 6" transient it was necessary to model the safety injection through the 

normal charging way (vols. 461, 462). Before the break is produced, a constant water 

mass flow rate is injected through the CVC system (vols. 455, 460). Once the transient 

is initiated this valve closes and, when the control variable that governs the inventory 

recovery is activated, the water begins to be injected through the time dependent 

junction. The mass flow rate injected depends on the RCS pressure. 

3.2. INITIAL CONDITIONS. 

A new input deck with the above modifications was built to simulate the plant 

state previous to both transients. It consisted of simulating the plant in Mode 5 with the 

following features: 

- RCS average temperature lower than 93.3°C. 

- RCS pressure about 24 kg/cm2. 

- Steam generators in wet layout. 

Other data measured in plant were added to the requirements above exposed. 

Table 2 shows a list of the most important parameters calculated with 

RELAP5/MOD3.2 code and their comparison with the values required or measured in 

plant. 

Once the initial conditions were reached and having checked that the plant was 

in a stable state, two new input decks were built considering the values of the variables 

at this point to perform the transient simulations.

9
3.3 LOCA 6" SIMULATION

In the LOCA 6" transient, it was necessary to activate the countercurrent flow 

limitation model due to the low RCS mass flow that leads to a misfunction of the code. 

This model was invoked only in the internal junctions where it was likely to occur, 

following the suggestions of RELAP5/MOD3.2 code manual [12] [13]. In this case it 

was activated in the core and core by-pass, the downcomer vessel, steam generator U

tubes, hot legs and the entrance of the vessel. 

In all junctions the Wallis CCFL correlation was used by default. The values for 

the Wallis correlation parameters, m and c, were taken from the literature [4] [12], 

which have been obtained from the experimental data of other experiments with similar 

working conditions to our situation. Possibly, the correlation used is not the proper one, 

since it seems not to work properly at low-pressure situations [6] 

After all the changes were made, the LOCA 6" transient was simulated following 

the sequence showed in figure 2, which describes the scenario foreseen in the plant PSA 

under these conditions. 

Firstly, the plant has been maintained at the initial conditions for 50 sec., to 

assure that the plant was in the initial state. Then, the transient is initiated by opening 

the valve that simulates the break in the RHR outside the containment. Immediately 

after the break opens there is a fast decrease in the pressurizer level, which causes the 

activation of the trips that isolate the CVC system charge and discharge, stopping the 

RHR pumps isolating the RHR heat exchangers through their by-pass. Safety injection 

starts injecting water 600 sec after the break was produced, through the normal charging 

way. The break can not be isolated during the whole transient.

10
3.4. LOCA 2" SIMULATION.

LOCA 2" transient was simulated at indicated in figure 3, which agrees with the 

sequence of events foreseen in the plant PSA under such conditions. 

In this case, the initial conditions were the same as for LOCA 6" transient, 

simulating 50 sec. before producing a 2" break in one of the RHR lines, outside the 

containment. Then, the break opens and the pressurizer level starts to decrease, but not 

so fast as in the previous case. When the pressurizer reaches a certain level the isolation 

of the RHR heat exchangers and the isolation of the CVC system charge and discharge 

trips are activated, and the RHR pumps are stopped. For this transient, it is supposed 

that the residual power generated is removed by the inventory of one steam generator, so 

the other two were isolated. In addition, no safety injection is allowed at any time.

11
4. RESULTS

From the transients developments the evolution of plant most important thermal

hydraulic variables have been obtained. Some of the most interesting evolutions are 

presented in this chapter. 

4.1. LOCA 6" TRANSIENT RESULTS 

For LOCA 6" transient 7300 sec. were simulated. During all this transient the 

core remains covered, as it can be seen in figure 4. This means that no damage to the 

core should be expected, as it is foreseen in the plant PSA. 

The most critical temperatures are the core outlet and clad temperatures, shown 

in figure 5. In this figure it can observed that all the temperatures have a peak just when 

the break is produced, to decrease immediately after. Although safety injection starts at 

650 sec., it has no effect on the core and clad temperatures until 1000 sec. when they 

begin to decrease. At the end of the transient the core temperatures tend to a stable 

value, which is approximately the temperature of the water injected. This result is also 

foreseen in the plant PSA, where it is supposed that the system will tend to a stable 

situation as long as water reservoir in the reactor water storage tank has to be recovered 

in a long term. 

In figure 6, the mass flow rate through the break and the safety injection mass 

flow rate are presented. When the valve opens, there is a peak in the mass flow rate 

through the break that reaches 700 kg/sec. After this peak, there is a period of time in 

which the flow presents a smooth decrease. At about 900 sec. there is a sudden fall of 

the mass flow rate and it starts to fluctuate near the -safety injection value. The safety 

injection mass flow rate has constant value during all the transient, since it depends on 

the primary system pressure change and, as can be seen in figure 7, the pressure remains 

almost at a constant value during all the time simulated after the break.

13
In figures 8 and 9 the hot and cold leg mass flow rates for each loop are 

represented. In both figures it can be observed that there are important mass flow rate 

fluctuations, not only when the break is produced, but during all the simulation. 

In figure 10 the total mass inside the three primary loops is presented. It can be 

seen that the mass inside the loops present a sharp decrease at the beginning of the 

transient. At about 900 sec. the mass starts to recover due to the effect of the safety 

injection, which finally tend to keep stable. 

4.2. LOCA 2" TRANSIENT RESULTS. 

For LOCA 2" transient 300000 seconds were simulated. Under the conditions in 

which the plant is working the residual power is so low that a large period of time is 

needed to observe significant changes in the thermal-hydraulic variables. As it is 

foreseen in the plant PSA, no damage to the core should be expected, as it can be seen in 

figure 11 where the core level remains covered during all the transient, although there 

are some fluctuations. 

In figure 12 the core outlet and clad temperatures are represented. All the 

temperatures rise from the beginning of the simulation until about 200000 seconds. At 

this time they remain at a constant value until the end of the transient simulation. 

Pressures in the primary and secondary side of the steam generators, see figures 

13 and 14 respectively, present the same shape as the temperatures. In this case, when 

the break is produced the RCS pressure, measured in the pressurizer, rises up until 

200000 seconds. At this time it remains stable at 8.E+6 Pa, which is nearly the same 

pressure value as for the steam generator secondary side. It is also at this time when the 

level of the steam generator 2 starts to decrease, due to the relief valve in this steam 

generator opens, what means that this steam generator is removing the residual heat. 

Figure 15 presents the evolution of the level of the steam generator in which it can be 

seen that even at the end of the transient there is still liquid inside the secondary side.

14
As in the previous case, the hot and cold legs mass flow rates present high 

oscillations, figures 16 and 17, which only disappear when the time simulated is 

extremely high, and the plant tends to be stable. 

Finally, figure 18 shows the pressurizer level evolution. When the valve, which 

simulates the break, opens the pressurizer level drops, and after the break is isolated it 

begins to recover until at 160000 sec., when the pressurizer is completely full.

15
4.3 SENSITIVITY STUDIES

Several sensitivity studies have been performed. In both transients the break was 

modeled as it is suggested in the RELAP5/MOD3.2 code manual [13], using the abrupt 

area change model. But, as it is not well known the actual mass flow rate through the 

break, it has been impossible to determine the suitable values for the discharge 

coefficients. The suggested values were tried, but in this case the flow through the break 

was extremely high in both transients, which led up to incoherent results. 

The solution adopted was to model the break supposing crossflow in the valve 

with regard to the main flow direction in the RHR pipe. That is, the connection from the 

RHR line with the valve is supposed as a crossflow, and the connection between the 

valve and the time dependent volume that simulates the atmosphere is defined as a 

normal junction with an abrupt area change. In this case the results for the mass flow 

rate seemed to be more reliable than in the previous case. 

In LOCA 2" transient two more sensitivity studies were performed. The first 

consisted on changing the maximum time step defined in the input deck, since in some 

transients the influence of time steps used in the calculations provoke differences in the 

values reached by the thermal-hydraulic parameters. In our case a maximum time step of 

0.5 sec. was selected because the variables change slowly. A shorter simulation was run 

using 0.05 sec as maximum time step without observing any difference between both 

simulations. 

Also in LOCA 2" a calculation was run activating the CCFL model, as it was 

made for LOCA 6" simulation, but in this transient the results obtained did not differ 

from those calculated without the possibility of invoking the CCFL model.

16
5. RUN STATISTICS

The calculations have been made in a CONVEX SPP 1000 owned by the UPV, 

using SPP-UX-3.1 as operating system and For77-HP as compiler. Table 3 presents the 

information about time steps used in each simulation and the CPU time consumed. 

Figure 19 shows the CPU values versus time simulated for LOCA 6", and figure 

20 presents the time step versus time simulated for the same transient. 

As it was done for LOCA 6", figure 21 represents CPU time versus simulation 

time for LOCA 2" and in figure 22 the time steps versus simulation time.

17
6. CONCLUSIONS AND RECOMMENDATIONS

It has been not possible to compare the values of the results obtained in the 

simulation with experimental data, as no experiment is available under these conditions. 

Thus, although the thermal-hydraulic variable evolutions agree with the success criteria 

foreseen in the plant PSA, it can not be assured that the values they reach are the actual 

ones due to the lack of such experimental results. 

Regarding the models implemented in the code, it was noted the misfunction of 

the code in simulating noncondensables presence, it was specially important in the 

LOCA 6" simulation due to the amount of noncondensable species that comes into the 

primary system through the break. 

Thus, the presence of noncondensables produces an important increase of the 

time needed for calculations, because it was necessary. to reduce the maximum time step 

required for the code to obtain a coherent solution. The introduction of noncondensables 

also produces a decrease in the transmission heat coefficients, specially in low flow 

conditions, just as it occurs in these studies. 

Also, due to the lack of experimental data it can not be assured that the values 

used for the constants of the CCFL model are the most suitable ones, since the values 

encountered in others studies with similar working conditions were adopted. We can not 

either be completely sure that the Wallis correlation used was the most appropriated for 

our studies. In this case, a further study in which a comparison of the results obtained 

with different correlations seems to be necessary, and this study also requires 

experimental data in order to compare the results. 

In the abrupt area change model, the values of the discharge coefficients used for 

the subcooled, two phases and superheated regimens, were taken from the literature for 

similar plant conditions, and although the results obtained seem to be quite reliable, 

nothing else can be assured.

19
So, the most important impediment for concluding that the results obtained with 

RELAP5/MOD3.2 are completely corrects is the lack of experimental data needed to 

assess the code capability under low power and low pressure conditions. It would be 

interesting the performance of experiments for these conditions to develop a more 

detailed study on the code capability.

20,
7. REFERENCES

[D] Yassin A. Hassan and Sibashis S. Banerjee. "RELAP5/MOD3 simulation of the loss 

of the residual heat removal system during midloop operation experiment conducted at 

the ROSA-IV large scale test facility". Nuclear technology Vol. 108. November 1994. 

[2] Sandra M. Sloan and Yassin Hassan. "Study of RELAP5/MOD2 and 

RELAP5/MOD3 predictions of a small break loss-of-.coolant accident simulation 

conducted at the ROSA-IV large scale test facility". Nuclear technology vol. 100. March 

1995. 

[3] Peter P. Cebull and Yassin A. Hassan. "Simulation of the IAEA's problem exercise 

small break loss-of-coolant accident using RELAP5/MOD3.1". Nuclear technology vol. 

105. March 1994. 

[4] Alessandro Annunziato. "Secondary-side feed-and-bleed effectiveness in pressurized 

water reactors". Nuclear technology vol. 105. March 1994. 

[5] W.S. Yeung and J. Shirkov. "On the subcooled critical flow model in 

RELAP5/MOD3". Nuclear technology vol. 114. April 1997. 

[6] Yuh-Ming Ferng and Chien-Hsiung Lee. "Assessment of the simulation capability of 

RELAP5/MOD3 compared with IIST tests for loss of the residual heat removal system 

during midloop operation". Nuclear technology vol. 116. October 1996. 

[7] Gy6rgy Ezs6l, Atilla Guba et altri. "Simulation of a small cold-leg-break experiment 

at the PMK-2 test facility using the RELAP5 and ATHLET codes". Nuclear technology 

vol. 118. May 1997. 

[8] Yuh-Ming Ferng and Chien-Hsiung Lee. "A comparison of RELAP5/MOD3 code 

with the IIST natural circulation experiments". Nuclear technology vol. 111. July 1995.

21
[9] Borut Mavko, Stojan Petelin, Oton Gortnar. "RELAP5 modelling of the 

Westinghouse model D4 steam generator". Nuclear technology vol. 101. February 1993. 

[10] C Llopis, R Mendizdbal, J. P6rez. "Assessment of RELAP5/MOD2 against a load 

rejection from 100% to 50% power in the Vandell6s II nuclear power plant". ICAP 

assessment report. October 1991. 

[11] Julidn Rozal6n, Carlos Llopis. "Modelaci6n del primario y sistemas asociados de la 

C.N. Vandell6s II con el c6digo RELAP5/2.5". Julio 1994. 

[12] "RELAP5/MOD3 code manual vol. 2: User's guide and input requirements". 

NUREG/CR-5535. June 1995. 

[13] "RELAP5/MOD3 code manual vol. 5: User's guidelines". NUREG/CR-5535. June 

1995. 

[14] "RELAP5/MOD3 code manual vol. 4: Models and correlations". NUREG/CR

5535. June 1995.

22
Table 1: Main plant characteristics.

Thermal Reactor Power (Mwt) 2775. 

Electrical Power (Mwe) 992. 

Fuel U02 

Number of assemblies 157 

Number of coolant loops 3 

Cladding tube material Zircaloy 4 

Absorber material B4C + Ag-In-Cd 

Reactor Operating Pressure (MPa) 15.4 

Coolant Average Temperature Zero load ('K) 564.8 

Coolant Average Temperature 100% load ('K) 582.3 

Stem Generator Westinghouse type F 

Number of tubes in SG 5626 

Total tube length (m) 98759. 

Inner diameter tubes (m) 0.0156 

Tube Material Inconel 

Pumps type Westinghouse D 100 

Discharge head of pumps (bar) 18.8 

Design flow rate (m3/sec)) 6.156 

Speed of pumps (rad/sec) 155. 

Primary volume (m3) 106.19 

Pressurizer Volume (M3) 39.65 

Heating Power of the heaters rods (KW) 1400. 

Maximum spray flow (kg/sec) 44.2 

Steam mass flow rate at 100% (kg/sec) 1515.

23
Table 2: Initial values

Reference parameters Problem Obtained 

Data RELAP-5/3.2 

Nuclear Power (%) - Residual Heat 0.05 (i) 0.05 (i) 

RCS Pressure (MPa) 23 23 

Pressurizer level (%) 25 25 

RCS average temperature (*K) - loops 1/2/3 < 366.46 365.43 / 365.38 / 364.91 

RCS hot leg temperature (*K) - loops 1/2/3 -- 366.01 / 365.98 / 364.94 

RCS clod leg temperature ('K) - loops 1/2/3 -- 364.85 / 364.88 / 364.89 

Core outlet temperature (°K) - loops 1/2/3 -- 366.02 / 366.02 / 364.94 

Clad temperature (°K) -- 366.37 

Core level (%) -- 6.6556 

RCS mass flow (Kg/s) - loops 1/2/3 -- 94.26 / 95.43 / 94.30 

Primary GVs mass flow (Kg/s) - loops 1/2/3 -- -2.27 / -4.93 / -4.69 

Secondary steam generator pressure (Mpa)- loops 1/2/3 -- 0.081 / 0.0785 / 0.0802 

Secondary GVs temperature (°K) - loops 1/2/3 -- 366.15 / 366.15 / 366.15 

Steam generator level (%) - loops 1/2/3 -- 82.02 / 82.02 / 82.35 

GVs Relief valves mass flow (Kg/s) - loops 1/2/3 -- 0.0 / 0.0 / 0.0 

RHR outlet pressure (MPa) - lines A/B -- 2.977 / 2.979 

RHR inlet pressure (MPa) - lines A/B -- 2.50/2.50 

RHR outlet temperature (°K) - lines A/B -- 364.87 / 364.84 

RHR inlet temperature (°K) - lines A/B -- 366.11 / 366.00 

RHR mass flow (Kg/s) - lines A/B -- 148.59 / 146.85 

CVCS charge mass flow (Kg/s) -- 2.6 

CVCS mass flow discharge (Kg/s) -- 2.6 

Break mass flow (Kg/s) -- 0.0 

Secondary RHR heat exchangers mass flow (Kg/s)-lines A/B -- 40.002 / 40.002

(i) Nominal power for 100% load is 2775.0 MW (1.3875 MW for 0.05%).

24
Table 3: Run statistics

Transient time (sec.) 

Execution time (sec.) 

Maximum time step (sec.) 

Total number of volumes 

Total number of time steps 

Grind time (msec.) GT=(CPU* 103/(C*DT))

RT CPU TS CPU/RT C DT GT 

4000. 13737.08 0.05 3.4343 252 119322 0.4605 

LOCA 6" 7000. 370944.41 0.001 52.9920 252 3149372 0.4674 

7200. 417045.81 0.0005 57.9230 252 3543058 0.4671 

LOCA 2" 309000. 77400. 0.5 0.25 281 665346 0.4139

RT: 

CPU: 

.TS: 

C: 

DT: 

GT:
Figurel: Vandellos H nodalization diagram for RELAP5JMOD3.2 under low power conditions.

REACTOR VESSEL

FR UL USLEG 

."CS OtSO.Ct

LPI SYSTEM

RHR LINE A

RHR LINE B 

LOOP 2
Figure 2: LOCA6" transient simulation.

PRZ low level signal 

RHR heat exchangers isolation 

RHR pumps trip 

n CVCS charge isolation 

CVCS discharge isolation

V

6" break in RHR

t = 50 seg

CVCS charge and discharge constant 

RHR mass flow constant 

No level control program 

No pressure control program

TRCS Safety injection 

HPCI pump 

No break isolation 

t = 650 seg

Simulatior 

start 

t =0
Figure 3: LOCA 2" transient simulation

PRZ low level signal

RHR heat exchangers isolation 

RHR pumps trip 

CVCS charge isolation 

CVCS discharge isolation 

SG #1 y #3 isolation (GV #2 active, 

no auxiliary water feed)

\i7

2" break in RHR 

t = 50 seg

CVCS charge and discharge constant 

RHR mass flow constant 

No level control program 

No pressure control program

Break isolation 

No RCS safety injection 

t = 650 seg

Simulation 

start 

t =0
level

29

0 le+03 2e+03 3e+03 4e+03 5e+03 6e+03 7e+03 

Time (seconds)
Figure 5: Core outlet and clad temperatures for LOCA 6".

0 le+03 2e+03 3e+03 4e+03 

Time (seconds

x- .x 

loop 2 

loop 1 

Clad 

5e+03 6e+03 7e+03

30

484

loop 3
0 

0 

I.

CD

Mass flow (kgfsec)

I.

.4. Q

0 0

0

(D 

0 

0 

WI4 

Cb 

0 

U.) 

0

CD)

*< 

C1x ) 

Cn r.n s

o -l cI

(t

II

0

r----- ----- -x: -- - u - ---- '-- - -- - - -- --- - -

I

U. 

U-... 0 

- - a 

*-0 

Q 

b A 

S. - - -

o - - - - - a ---.-u



- -U - - - -

-z: a 

.5-- I

0 C 

a -a 

0 a 

-

O - - - -

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

I - - - - - - - I -
Figure 7: Reactor coolant system pressure for LOCA 6".

p 40405 

0 le+03 2e+03 3e+03 4e+03 5e+03 6e+03 7e+03 

Time (seconds)

32

2.3e+06
Figure 8: Hot legs mass flow rates for LOCA 6".

loop 3 

X- x 

loop 2 

o..o a loop I

0 le+03 2e+03 3e+03 

Time

4e+03 5e+03 6e+03 

(seconds)

33

7e+03
Figure 9: Cold legs mass flow rates for LOCA 6"

943

C

-550

0 le+03 2e+03 3e+03 

Time

4e+03 5e+03 

(seconds)

loop 3 

)f- -x 

loop 2 

loop I 

7 

6e+03 7e+03

34
Figure 10: Reactor coolant system mass for LOCA 6".

1.19e+05

le+05

0 le+03 2e+03 3e+03 4e+03 5e+03 

Time (seconds)

6e+03 7e+03

35

water 

*- C. 

vapor 

total
Figure 11: Core level for LOCA 2"

le+05 

Time

2e+05 

(seconds)

36

l-v 

level

6.66 

6.6

6.5

6.42

0 3e+05
Figure 12: Core outlet and clad temperatures for LOCA 2".

le+05 

Time

2e+05 

(seconds)

37

571

loop 3 

X- .X 

loop 2 

loop I 

*- -- a 

Clad

0 3e+05
Figure 13: Reactor coolant system pressure for LOCA 2".

p- 4 040 5 

0 le+05 2e+05 3e+05 

Time (seconds)

38
Figure 14: Steam generators pressures for LOCA 2".

SG3

)E- X 

t SG 2 

I 

7e+06- , .. 

t SG I 

I 

6e+06. 

A 

I 

5e+06

I 

4e+06I 

2e+06- : 

2, I 

J I 

4 / 

3e+06. 

,• / 

S.../ 

/ 2e+06-/' 

/ 

// 

le+06

7.68e+04 

0 le+05 2e+05 3e+05 

Time (seconds)

39
Figure 15: Steam generators levels for LOCA 2".

107 

100

5.75

40

SG3 

x- •\ - --.x 

SG 2 

\SG I 

\ 

le+05 2e+05 3e+05 

Time (seconds)

0
Figure 16: Hot legs mass flow rates for LOCA 2".

0 le+05 2e+05 

Time (seconds)

41

loop 3 

x- -x 

loop 2 

loop 1

369

0•

-95.9

3e+05
Figure 17: Cold legs mass flow rates for LOCA 2".

0 le+05 

Time

*- .x 

loop 2 

loop 1 

X-- * *- . -- --X--X , 

/ 

1X . X-- X-- X, X0 , 

0' 

2e+05 3e+05 

(seconds)

42

465

loop 3
Figure 18: Pressurizer level for LOCA 2"

le+05 2e+05 

Time (seconds)

43

level

110

C.) 

C.)

-9.37

0 3e+05
Figure 19: CPU time for LOCA 6".

4.17e+05

ri

1

C 

'-' 2e+05 

Q

0 le+03 2e+03 3e+03 4e+03 5e+03 

Time (seconds)

6e+03 7e+03

44

Cpu
Figure 20: Time steps for LOCA 6".

U, 

C 0.04

i 0.03-

0 le+03 2e+03 3e+03 4e+03 5e+03 6e+03 7e+03 

Time (seconds)

45
Figure 21: CPU time for LOCA 2".

7.74e+04

0 

C.? 

'I, 

U

0 le+05 2e+05 

Time (seconds)

46

Cpu

3e+05
Figure 22: Time steps for LOCA 2"

0 le+05 2e+05 

Time (seconds)

47

1

oNfw

3e+05
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (2-W) 

NRCM (Assigned by NRC, Add Vol., Supp., Rev., 110Z B DATA SHEET and Addendum Numbers, If any.) 3201,3202 DlDIBUCIG PIC DAT SHEE 

(See insb'uckns on m&e evown) 

2. TITLE AND SUBTITLE NUREGAA-0171 

Simulation of LOCA 6' and LOCA 2" Transients in the RHR of a PWR 

Under Low Power Conditions Using RELAP5/MOD3.2 3. DATE REPORT PUBLISHED 

MONTH YEAR 

Februa 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

S. Martorell, S. Carlos, G. Verdu, V. Serradell, J. Rodenas, R. Miro, C. Uopis* Technical 

7. PERIOD COVERED (Inclush, Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS 7fYNRC, provide Diviskon, Oef or Regio. U.S. Nudear RegulAty Comrission, and mailirg add*"s, i contacio 

provide no, and ma•ti address) 

C. Vandellos II Department of Chemical and Nuclear Engineering 

Travessera de les Corts, 55, Lateral Polytechnic University of Valencia 

08028 Barcelona Camino de Vera, 14 

Spain

9. SPONSORING ORGANIZATION - NAME AND ADDRESS ( NRCe "Same above, if conkctor, provide ARC Division, Offle orRegion, U.S. NucleeRegulaoray Commission, and mailir addres.)

Division of System Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research, 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

11. ABSTRACT w20o words or ess) 

The present study consists of the simulation of two loss of coolant accidents, LOCA 6' and LOCA 2", in one of the residual heat 

removal system (RHR) lines outside the containment using the thermal-hydraulic code RELAP5/MOD3.2. Both transients have been simulated on a typical three loop, Westinghouse design, pressurized water reactor plant working under shutdown conditions. The study was focused on the simulation of the most important thermal-hydraulic parameters in order to check the validity of the 

success criteria assumed in the plant probabilistic safety analysis (PSA) under shutdown conditions. Also to analyze the code capability for simulating shutdown conditions was of interest in this study. As a result of this study, it can be concluded that the main thermal-hydraulic plant features follow what is foreseen in the plant PSA, although it can not be assured that the values reached are the correct ones due to the lack of experimental data.

12. KEY WORDSI/DESCRIPTORS (List words orplmees met wm ,.iassw&treaess mi ocai0& meraprt) 13. AVAJLABUTY STATEMENT 

RELAP5/MOD3.2, LOCA 6", LOCA 2", PSA, low pressure, RHR unlimited 

14. SECURITY CLASSIFICATION 

unclassified 

(This Report 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

-~. ý,.• q •OW) W 

IV. •vrrlwrallli;ll iPlll•l nl•/ll;•} t
Federal Recycling Program
NUREG/IA- 0171

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

FEBRUARY 2000

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

JSNRC 

PERMIT NO. G-67

SIMULATION OF LOCA 6" AND LOCA 2" TRANSIENTS IN THE RHR OF A PWR 

UNDER LOW POWER CONDITIONS USING RELAP5/MOD3.2