c- -o -o g-2

NUREG/IA-0173 

International 

Agreement Report 

Simulation of a 

Station Black-Out in a PWR 

Under Midloop Conditions 

Using RELAP5/MOD3.2 

Prepared by 

S. Martorell, S. Carlos, G. Verd6i, 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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owned rights.
NUREG/IA-0173

International 

Agreement Report 

Simulation of a 

Station Black-Out in a PWR 

Under Midloop 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
EXECUTIVE SUMMARY

In this paper the simulation of the loss of the residual heat removal system 

(RHRS) due to a station black-out in mid-loop conditions is presented, using the 

RELAP5/MOD3.2 thermal-hydraulic code. Although this code was primarily developed 

to perform LOCA transients at full power, the use of the code to simulate transients with 

the plant in other operational modes is being also considered. In this case, the interest is 

focused on studying all the phenomena involved in a loss of the residual heat removal 

system (RHRS) due to an external station black-out (SBO) during mid-loop operation. 

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

(ECCS) in a nuclear reactor, which is responsible of the residual heat removal under low 

power and shutdown conditions. Sometimes, the RHR system is required to work with 

the primary inventory level reduced to the height of the primary loop, and the upper part 

of the reactor coolant system (RCS) filled with air. This mode of operation is called 

mid-loop operation. 

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

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

low pressure conditions. It was 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. 

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

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

RELAP5/MOD3.2 code capability for simulating transients under mid-loop operation.

I
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 initially in mid-loop with the pressurizer manway 

opened. 

Under these conditions the plant is stable in Mode 5 and, when the station black 

out is produced, it is supposed that there is no possibility of recovering the external 

electric power supply, and that the only way for removing the residual heat produced is 

by gravity feed from the refueling water storage tank through the connection with the 

low pressure injection system of the residual heat removal system, being 2820 seconds 

the maximum time available for taking this action. 

In the plant PSA it is supposed that when. the SBO transient is produced, 

following the actions above explained, the plant will tend to a stable situation, and no 

damage to the core is expected to occur. 

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. 

From the transient development we have obtained the evolutions of the plant 

most important thermal-hydraulic variables. The calculations were run in a CONVEX 

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

transient simulation 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. Such a large time being simulated results in a high cost of CPU time needed 

for this transient

2
It seems from the simulation results that there is a misfunction of the code in the 

presence of noncondensable. It was specially important in this simulation due to the 

amount of noncondensable species present in the primary system, that produces an 

important increase of the time needed for calculations, since the code requires a lower 

maximum time step to obtain a coherent solution. The introduction of noncondensables 

also seems to produce an underestimation of the heat transmission coefficients. 

It can also be concluded that the simulation results obtained with 

RELAP5/MOD3.2 reproduce, satisfactorily, the success criteria foreseen in the PSA, 

although experimental data would be necessary to verify that they are the correct values 

and also to validate the code proper performance.

3
ABSTRACT

The present study consists of the simulation of a station black-out (SBO) 

transient, when the plant is in mid-loop conditions, using the thermal-hydraulic code 

RELAP5/MOD3.2. 

This transient has been simulated on a typical three loop, Westinghouse design, 

pressurized water reactor (PWR) model working under mid-loop conditions. 

The study was focused on obtaining the most important thermal-hydraulic 

variables in order to check the validity of the success criteria described in the plant 

probabilistic safety analysis (PSA), and also to analyze the code capability to simulate 

such conditions. 

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

plant features follow what it is foreseen in the plant PSA, although the values that are 

reached can not be completely taken as the correct ones due to the lack of experimental 

data to validate RELAP5 under these conditions.

4
TABLE OF CONTENTS 

LIST OF TABLES ........................................ ................. 6 

LIST OF FIGURES .......................................................... 7 

1. INTRODUCTION ........................................................ 8 

2. PLANT AND TRANSIENT DESCRIPTION .................................. 10 

2.1. PLANT DESCRIPTION .............................................. 10 

2.2. TRANSIENT DESCRIPTION ......................................... 11 

3. CODE INPUT AND MODEL DESCRIPTION ............................... 12 

3.1. SYSTEM DESCRIPTION ........................................... 12 

3.2. INITIAL CONDITIONS ............................................. 14 

3.3. SBO SIM ULATION ................................................. 14 

4. SBO TRANSIENT RESULTS ............................................. 16 

5. RUN STATISTICS .................................................... 17 

6. CONCLUSIONS AND RECOMMENDATIONS ................................ 18 

7. REFEREN CES ......................................................... 19

5
LIST OF TABLES: 

TABLE 1: MAIN PLANT CHARACTERISTICS ................................. 21 

TABLE 2: INITIAL VALUES ................................................ 22 

TABLE 3: RUN STATISTICS ................................................ 23

6
LIST OF FIGURES:

FIGURE 1: VANDELLOS H NODALIZATION FOR RELAP5/MOD3.2 

UNDER MIDLOOP CONDITIONS .................. 

FIGURE 2: SBO TRANSIENT SIMULATION .................... 

FIGURE 3: CORE LEVEL .................................... 

FIGURE 4: CORE OUTLET AND CLAD TEMPERATURES ......... 

FIGURE 5: GRAVITY INJECTED MASS ........................ 

FIGURE 6: MASS INSIDE THE PRIMARY LOOPS ................ 

FIGURE 7: PRESSURIZER LIQUID LEVEL ...................... 

FIGURE 8: HOT LEG MASS FLOW RATE ....................... 

FIGURE 9: COLD LEG MASS FLOW RATE ...................... 

FIGURE 10: CPU TIME ...................................... 

FIGURE 11: TIME STEPS ....................................

...... . 24 

.............. 25 

.............. 26 

.............. 27 

.............. 28 

.............. 29 

.............. 30 

.............. 3 1 

.............. 32 

.............. 33 

.............. 34

7
1. INTRODUCTION

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

pressurized water reactor (PWR) plants. This code was primarily developed to perform 

loss-of-coolant (LOCA) accidents when the system is working at full power. 

Nevertheless, the use of the code to simulate transients with the plant in other 

operational modes is being considered. In our case, we are interested in studying all the 

phenomena involved in a loss of the residual heat removal system (RHRS) due to an 

external station black-out during mid-loop operation. 

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

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

low power and shutdown conditions. 

Sometimes, the RHR system is required to work with the primary inventory level 

reduced to the height of the primary loop, and the upper part of the reactor coolant 

system (RCS) filled with air. This mode of operation is called mid-loop operation. 

Under this working conditions it is of great interest to study those transients that 

cause the loss of the RHR system, in order to check the capability of the alternative 

ways foreseen in the PSA to evacuate the residual heat. 

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

obtained and the limitations encountered, the success conditions established in a 

probabilistic safety analysis (PSA) for a PWR plant in these operational conditions. 

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 conditions. 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.

8
This document is organized as follows. In chapter two, a brief plant description 

is presented, highlighting its most important features. The description of the transient 

being simulated, SBO, is also presented in this chapter. The input file used for reaching 

initial conditions of the transient is explained in chapter three, together with specific 

nodalization modifications needed to simulate the transient. Chapter four contains the 

results of the simulation, showing the time-dependent evolutions of some thermal

hydraulic variables considered as important. It also contains the discussions of the 

additional sensitivity studies performed and some recommendations on the aspects that 

should be subject of a future study. Chapter five gives information about the time steps 

utilized along the transient simulation and the time consumed in the transient 

development. The conclusions derived from the whole study, and the most important 

recommendations for other users, are discussed in chapter six. All the 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 this report.

9
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, with the primary level in mid-loop only one of the RHR pumps recirculates 

water to the three loops with a total mass flow rate of 90 kg/sec.

10
2.2 TRANSIENT DESCRIPTION

The experiment consists of simulating the loss of the residual heat removal 

system due to an external station black-out (SBO) with the plant working in mid-loop 

conditions in Mode 5, according to the following characteristics: 

- Pressurizer manway opened. 

- Primary level of the RCS hot legs reduced up to 10± 3 cm above the vessel 

entry. 

Under these conditions the reactor coolant system (RCS) average temperature is 

close to 60 °C and there is only one pump of the RHR system working, which injects in 

the three loops a total mass flow rate of 90 kg/sec. Then, the plant is stable in Mode 5, 

when the station black out occurs. We suppose that there is no possibility of recovering 

the external electric power supply, and that the only way of removing the residual heat 

produced is by gravity feed from the refueling water storage tank through the connection 

with the low pressure injection system of the residual heat removal system, being 2820 

seconds the maximum time available for taking this action manually.

11
3. CODE INPUT AND MODEL DESCRIPTION.

3.1. SYSTEM DESCRIPTION 

The simulation has 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. 

At the beginning of this study, we had the plant nodalization for full power 

conditions, and we needed to adapt the model to work under mid-loop conditions, as 

this was the previous state of the plant for starting the transient. 

An important nodalization change was the RHR system modification to work in 

recirculation mode, as 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 I and 

2 of the primary system with the lines A (vols.300, 301, 302, 303) and B (vols. 320, 

321, 322, 323) of the low pressure injection system. Among these added volumes we 

can find the RHR heat exchangers (vols. 308, 309 328, 329), which are able to remove 

all the residual power generated. 

From the input deck, the volumes that simulate the turbine, the steam dump, the 

MSRV's and the accumulators have been eliminated, since they are not used in the 

transients simulated. 

The most important change in the 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 

modifying the suction and discharge loss coefficients until the mass flow rate through 

the RCS was the same as the one measured in plant.

12
At the secondary of the steam generators the water injection through the main 

and auxiliary feedwater systems have been deactivated, although have not been removed 

since they could be required in future transients under low power conditions. 

At this point, the plant model was adapted for working in low power conditions, 

but for simulating mid-loop operation is necessary to add the volumes that simulate the 

pressurizer manway. The top of the pressurizer is connected with a time dependent 

volume (vol. 347) that simulates the atmosphere through a trip controlled valve (vol. 

346) which has as main characteristics the utilization of the abrupt area change model 

and the choking model. 

To avoid the problems the choking model may present when simulating 

shutdown conditions, the current recommendation given in the RELAP5/MOD3.2 code 

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

connections to the atmosphere. That will produce mass fluxes close to the homogeneous 

equilibrium critical flow model. Instead, it is suggested to invoke nonhomogeneous 

model with the choking model turned off for internal junctions 

As we know, in the transient development the gravity injection system is 

required. In this way, the refueling water storage tank is modeled by four branches, two 

of them connected with the low pressure injection system (vols. 852 and 854). The pipes 

between the tank and the low pressure injection lines are simulated by two branches 

(vols. 835 and 840) that connect it with the R-IR through two trip controlled valves 

(vols. 304 and 324), which open when the gravity injection is required.

13
3.2. INITIAL CONDITIONS.

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

state previous to the transient in mid-loop operation with the following features: 

- Pressurizer manway opened. 

- Primary level of the hot legs reduced up to 10± 3 cm above the vessel entry. 

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

Table 2 shows a list of the most important variables calculated with RELAP5/MOD3.2 

code, and their comparison with the values required or measured in plant, for the initial 

state. 

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 simulation. 

3.3 TRANSIENT SIMULATION 

In the SBO transient, due to the low RCS mass flow which leads to a 

misfunction of the code, it was necessary to activate the countercurrent flow limitation 

model. This model was invoked only in the internals junctions where it was likely to 

occur, following the suggestions of RELAP5/MOD3.2 code manual [12] [13]. In our 

case, it was activated in the core and in the core by-pass, in the vessel downcomer, in 

the steam generators U-tubes, in the hot legs, and in the entrance to the vessel. 

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

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

These values have been obtained from the experimental data of other experiments with 

similar working conditions to our situation.

14
After all the changes were made, the SBO transient was simulated following the 

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

under these conditions. 

In the transient simulation we wait for 50 seconds before taking any action to 

assure the initial conditions were reached. After that, the RHR pump A that is working 

in mid-loop operation is stopped. No other action is taken until 2820 seconds after the 

pump stops, when the gravity injection from the reactor water storage tank is manually 

activated.

15
4. RESULTS.

From the transient development we have obtained the evolutions of plant most 

important thermal-hydraulic variables. Some of the most interesting evolutions are 

presented in this chapter. 

About 25000 sec. of the station black-out transient were simulated to find 

significant results in the thermal-hydraulic variables. 

As we can see in figure 3, the reactor core level has no important change during 

the simulation. In fact it remains with no change until 14000 sec., at this time it starts to 

decrease, but no core damage occurs since the core is never uncovered. 

The behavior of the most significant temperatures is presented in figure 4. In this 

figure we can observe how the core outlet and clad temperatures rise before the 

injection. When the water from the RWST is injected there is a decrease in all the 

temperatures, but once the injection has finished they start to rise again, but not so fast 

as before the injection. 

The mass injected from the RWST is presented in figure 5. In this figure we can 

observe that all the water is injected around 2870 sec. after the SBO, and no other 

injection is produced in the simulation. A certain quantity of the water injected fills the 

RCS loops, figure 6, which where partially filled with air at the beginning of the 

simulation. The water injected is also utilized to fill the pressurizer, which was 

completely empty before the injection. This can be seen in figure 7, where the 

pressurizer level is presented. 

Finally, figures 8 and 9 represent the mass flow rates through the hot and clod 

legs respectively. In both evolutions the most important feature is the oscillating 

behavior of the mass fluxes, which is specially important in the first period of the 

simulation for the hot leg mass flow rates, when the RCS loops were partially filled with 

air.

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 times consumed. 

Figure 10 shows the CPU values versus time simulated for the SBO, and in 

figure 11 the time step versus time simulated for the same transient is represented.

17
6. CONCLUSIONS AND RECOMMENDATIONS

Although the thermal-hydraulic parameters evolution agrees 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 experimental data to validate them. 

The misfunction of the code when simulating noncondensable presence was 

noted. It was specially important in the simulation due to the amount of noncondensable 

species present in the primary system. The presence of noncondensables produces an 

important increase of the time needed for calculations, since the code requires a lower 

maximum time step to obtain a coherent solution. The introduction of noncondensables 

also produces an underestimation of the transmission heat coefficients. 

It can not be assured that the constants used in the CCFL model are the most 

suitable ones, because the values encountered in others studies with only similar 

working conditions were adopted. In addition, it can not be assured that the correlation 

used was the most adequate for our studies. Probably, the correlation used is not the 

proper one, as it seems not to work properly at low pressure situations [6]. In this case, it 

should be necessary a further study in which a comparison of the results obtained with 

different correlations should be done. This study also requires experimental data in order 

to validate the results. 

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 conditions. It would be interesting the 

performance of experiments in these conditions to develop a more detailed study on the 

code models.

18
7. REFERENCES

[1] 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 acccident 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 HIST tests for loss of the residual heat removal system 

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

[7] Gybrgy tzsrl, 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 Femg and Chien-Hsiung Lee. "A comparison of RELAP5/MOD3 code 

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

19
[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. Perez. "Assessment of RELAP5/MOD2 against a load 

rejection from 100% to 50% power in the Vandell6s I1 nuclear power plant". ICAP 

assessment report. October 1991. 

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

C.N. Vandell6s I1 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. 

[15] Yassin A. Hassan and Laxminarayan L. Raja. "Simulation of loss of RHR during 

midloop operations and the role of steam generators in decay heat removal using the 

RELAP5/MOD3 code". Nuclear technology Vol. 103. September 1994.

20
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 (

0K) 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.

21
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) 0.1 0.1 

Pressurizer level (%) 

RCS average temperature (*K) - loops 1/2/3 333. 335.3 

RCS hot leg temperature ('K) - loops 1/2/3 -- 338.8 

RCS clod leg temperature (*K) - loops 1/2/3 - 328.4 

Core outlet temperature (*K) - loops 1/2/3 -- 336.33 

Clad temperature (

0K) -- 343.8 

Core level (%) -- 6.6556 

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

Steam generator level (%) (Average) -- 95.46 

Total RHR mass flow (Kg/s) 90 89.924 

Hot legs average level (m) 0.4683 0.4975 

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

22
Table 3: Run statistics.

It - - I - I

CPU TS CPU/RT C DT GT

2.9 1E6 0.001 11.6867 24901414 04

Transient time 

Total execution time. 

Maximum time step. 

Total number of volumes in the model. 

Total number of time steps 

Grind time (msec.) GT=(CPU* 1000)/(C*DT)

RT I 24900

N") 

w,,

C

284 1

DT

RT: 

CPU: 

TS: 

C: 

DT: 

GT:

24901414

01T

I I I 0.4115 11
Figure 1: Vandellos II nodalization for RELAP5/MOD3.2 under mid-loop conditions 

IOU 24636 ~ 664 H $30 R T VS6 6 

cqx rqxREACTOR VESSEL

CeCs VISCHARC 

ESSEL 

LPI SYSTEM

620

FROW

RHR LINE A 

- LOOP 1 

RHR LINE B 

- LOOP 2

414
Figure 2: SBO transient simulation.

Simulation 

start 

t=0

"B lack Out", 

RHR pump trip 

t 50 seg

RCS inventory recovery from the RWST 

by gravity feed 

t = 2870 sec. (47 min.)
Figure 3: Core level.

le+04 

Time (seconds)

2e+04

level 

2.49e+04

26

6.66

0
Figure 4: Core outlet and clad temperatures.

400

" "• loop 3 

,?-41, loop 2 / .. r"

390" / -" f .o" ploop I 

Clad / ar 

380" 

iri / A '. 

S/ 9 

,370 . 

* V 

S360

350- . 

340--. 

333 .. 

0 le+04 2e+04 2.49e+04

Time (seconds)

27
Figure 5: Gravity injected mass.

6.7e+04 

6e+04

5e+04 

4e+04. 

3e+04. 

2e+04

le+04-

1 e+04 

Tiempo (segundos)

2e+04 2.49e+04

28

Line A 

Line B 

Total

I

0 I

0

.o . . .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. . ... .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. .. ...... . ...........

0
Figure 6: Mass inside the primary loops.

7.75e+04

le+04 

Time (seconds)

2e+04 2.49e+04

29

Water 

Steam 

Total

0
Figure 7: Pressurizer level.

Level

-9.68 1',.-I....... ................... 

0 le+04 2e+04 2.49e+04

'lime (seconds)

30
Figure 8: Hot leg mass flow rates.

237

loop 3 

200 *- x 

loop 2 

loop I 

100

: t 

-100 

6o 

-173......... . ..... 

0 le+04 2e+04 2.49e+04

lime (seconds)

31
Figure 9: Cold leg mass flow rates.

le+04 

Time (seconds)

2e+04

loop 3 

*- -x 

loop 2 

a --- a 

loop 1 

'-4ei 

2.49e+04

32

51.4

-64.5

0
Figure 10: CPU time.

2.91e+06

le+04 

Time (seconds)

2e+04 2.49e+04

33

0-0 

CPU

0
Figure 11: Time steps.

0.002 

0.001 

0

le+04 

Time (seconds)

2e+04

time step 

2.49e+04

34

a p * p p p p p p p * 2 P

0

I a I
NRC FORM 338 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-8W (Assigned by NRC, Add VoL. SUpp. Rev., BIBUOGRAPHIC DATA SHEET and Addendum Numbers, If any.) 

(Smef rn~wc/ th e ryewrs) 

2. TTLE AND SUBTITLE NUREG/IA-0173 

Simulation of a Station Black-Out in a PWR Under Midloop Conditions 

Using RELAP5/MOD3.2 3. DTE R PBIH MONTH I YEAR 

February 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

S. Martorell, S. Carlos, G. Verdu, V. Serradell, Technical 

J. Rodenas, R. Miro, C. Uopis* 

7. PERIOD COVERED (lncdwive Dow) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, p vldeDiviison Oflwoe rRegion, U.S. NucelwReglafty C ,mmus and mailig adde•s ifonbac, 

pmide ne, eWnd mailing adrsew) 

*C.N. Vandellos Il 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 (/f NRC, type 'Same as above, if con&tcr, prvide NRC Diior4 Offime orRegion, U.S. NudewRegulatry Comrnasion, andmaig ad&asj

Division of System Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington. DC 20555-0001

IV./, SJl'rr'iYr= I I•I% 1 1 1-C

11. ABSTRACT (200 wads cris) 

The present study consists of the simulation of a station black-out (SBO) transient, when the plant is in mid-loop conditions, using the thermal-hydraulic code RELAP5/MOD3.2. This transient has been simulated on a typical three loop, Westinghouse design, pressurized water reactor (PWR) model working under mid-loop conditions. The study was focused on obtaining the most 

important thermal-hydraulic variables in order to check the validity of the success criteria described in the plant probabilistic safety analysis (PSA), and also to analyze the code capability to simulate such conditions. As a result of this study, it can be concluded that the main thermal-hydraulic plant features follow what it has foreseen in the plant PSA, although the values that are reached 

can not be completely taken as the correct ones due to the lack of experimental data to validate RELAP5 under these condirtions.

12. KEY WORDS/DESCRIPTORS (List words orphraswe that wdassstrseerchws in J w f trepot.) 13. AVAILABILITY STATEMENT 

RELAP5 unlimited 

LOCA 14. SECURrIY C.ASSIRCATION 

RHR (Tis Pag) 

unclassified 

(This Repo"l 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

MW FORM 400 (2-89)

4A e"
Federal Recycling Program
SIMULATION OF A STATION BLACK-OUT IN A PWR FEBRUARY 2000

UNDER MIDLOOP CONDITIONS USING RELAP5/MOD3.2

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67

NUREG/IA- 0173