NUREG/IA-0459 

International 

 Agreement Report

EPR Medium Break LOCA Benchmarking 

Exercise Using RELAP5 and CATHARE

 

 

Prepared by: 

Sebastian Gurgacz, Michał Pawluczyk, Piotr Mazgaj, 

Piotr Darnowski, Kacper Samul, Maciej Skrzypek

Warsaw University of Technology, 

Institute of Heat Engineering, 

Nowowiejska 21/25, 

00-665 Warsaw, Poland

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: July 2015

Date Published: December 2015

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0459 

International 

 Agreement Report

EPR Medium Break LOCA Benchmarking 

Exercise Using RELAP5 and CATHARE

 

 

Prepared by: 

Sebastian Gurgacz, Michał Pawluczyk, Piotr Mazgaj, 

Piotr Darnowski, Kacper Samul, Maciej Skrzypek

Warsaw University of Technology, 

Institute of Heat Engineering, 

Nowowiejska 21/25, 

00-665 Warsaw, Poland

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: July 2015

Date Published: December 2015

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

Thermal-hydraulic analyses are a key part in support of regulatory work for new and existing 

nuclear power plant design and operation. This paper describes the approach to model the Loss 

of Coolant Accident (LOCA) in a Light Water Reactor as part of the “Safety Analysis Report in 

Warsaw University of Technology” (SARWUT) project and the framework of the “Familiarization 

with the calculation codes application” program. 

The RELAP5 model of the European Pressurized Reactor (EPR) has been developed on the 

basis of an available CATHARE-2 input. Both thermal-hydraulic codes, RELAP5 and 

CATHARE-2, are used for the safety analysis of the NPP. The purpose of this report is to 

present the intermediate (6-inch) cold leg break calculations performed within the benchmark 

exercise using both RELAP5 and CATHARE-2 codes.

The results received are satisfactory, however as presented, the calculations performed with the 

use of both computer codes, at the early-stage give low cladding temperature but differ in the 

transient characteristics. The discrepancies in the values of chosen safety related parameters 

are analyzed in detail for understanding and future work.

v

CONTENTS

Page

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

FIGURES..................................................................................................................................vii

TABLES ....................................................................................................................................ix

EXECUTIVE SUMMARY ...........................................................................................................xi

ACKNOWLEDGMENTS.......................................................................................................... xiii

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

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

2. THE EPR GENERAL DESIGN............................................................................................. 3

3. THE BENCHMARK EXERCISE PROCEDURE ................................................................... 5

4. THE RELAP5 THERMAL-HYDRAULIC CODE MODEL...................................................... 7

 4.1 RELAP5 Nodalization................................................................................................. 7

 4.2 Safety Systems .........................................................................................................11

5. CATHARE ANALYTICAL METHOD...................................................................................13

 5.1 CATHARE-2 Thermal-Hydraulic Code.......................................................................13

 5.2 CATHARE Nodalization.............................................................................................13

 5.3 Critical Flow Model in CATHARE ..............................................................................16

6. CALCULATIONS................................................................................................................17

 6.1 Scenario Description .................................................................................................17

 6.2 Steady-State Results.................................................................................................17

 6.3 Transient Results ......................................................................................................18

7. DISCUSSION......................................................................................................................27

8. RUN STATISTICS...............................................................................................................29

9. CONCLUSIONS..................................................................................................................31

10. REFERENCES....................................................................................................................33

vii

FIGURES

Page

Figure 1 RPV Nodalization Scheme. ...................................................................................... 7

Figure 2 Core Heating Regions With Zone Coloring............................................................... 8

Figure 3 Core Nodalization Scheme With Zone Coloring........................................................ 8

Figure 4 Axial Power Shape as a Function of the Normalized Core Height. ........................... 9

Figure 5 Primary Side Single Loop Nodalization Scheme......................................................10

Figure 6 Steam Generator Secondary Side Nodalization Scheme. .......................................11

Figure 7 MHSI And LHSI Mass Flow Rate Curves. ...............................................................12

Figure 8 Reactor Pressure Vessel Nodalization Scheme. .....................................................14

Figure 9 Primary Loop Nodalization Scheme. .......................................................................15

Figure 10 Nodalization Scheme of the Secondary Side...........................................................15

Figure 11 Pressure in Core. ....................................................................................................18

Figure 12 Pressurizer Pressure...............................................................................................19

Figure 13 Water Level in the Pressurizer. ...............................................................................19

Figure 14 Water Level in the Pressurizer (First 100 S)............................................................20

Figure 15 Surge Line Mass Flow. ............................................................................................20

Figure 16 Steam Generator Pressure. ....................................................................................21

Figure 17 Steam Generator Water Level. ................................................................................21

Figure 18 Pump Speed. ..........................................................................................................22

Figure 19 MHSI Mass Flow Rate.............................................................................................22

Figure 20 LPSI Mass Flow Rate..............................................................................................23

Figure 21 Reactor Core Water Level.......................................................................................23

Figure 22 Cladding Temperature.............................................................................................24

Figure 23 Void Fraction at the Break.......................................................................................24

Figure 24 Integrated Break Mass Flow Rate. ..........................................................................25

Figure 25 Break Mass Flow Rate. ...........................................................................................25

Figure 26 Accumulator Mass Flow Rate..................................................................................26

Figure 27 Integrated Surge Line Mass Flow Rate. ..................................................................26

ix

TABLES

Page

Table 1 General Design Parameters for the EPR..................................................................... 4

Table 2 Steady-State Results Calculated in RELAP5 and CATHARE. ....................................17

Table 3 Sequence of Events. ..................................................................................................18

Table 4 Run Statistics .............................................................................................................29

xi

EXECUTIVE SUMMARY

There is a number of thermal-hydraulic computer codes used for safety analysis in nuclear 

installations. Such codes may vary within the scope of applicability, empirical correlations (e.g. 

for heat exchange under certain conditions or critical flow models) programmed in those codes

and modeling approach. Warsaw University of Technology has been provided with several 

thermal-hydraulic codes. This benchmarking exercise has been carried out in order to better 

understand the performance of two of them: RELAP5 and CATHARE. 

The Intermediate Break LOCA in the European Pressurized Reactor has been selected for 

modeling as a benchmarking exercise. The first step was to reach an acceptable steady-state. 

The second step was to perform calculations of the selected transient in order to find and 

compare predictions of both codes. A number of varying parameters have been plotted versus 

time. Finally, the differences between the results obtained by RELAP5 and CATHARE have 

been discussed.

In chapter 2 a brief description of the EPR design and main parameters is provided. It 

comprises general data with regard to nominal and design parameters, thermal power and 

electric output and core design data. The thermal-hydraulics data provided consists of nominal

temperatures in cold and hot legs, coolant mass flow and heat fluxes. There is also a 

description of containment parameters, such as total volume, design pressure and temperature.

Chapters 4 and 5 cover a brief description of both CATHARE and RELAP5 codes. The main 

differences in applied correlations and modeling philosophy have been outlined. Nodalization 

schemes, together with detailed description of the core region modeling and safety systems

have been also presented.

The analyzed scenario description has been provided in chapter 6. The steady-state results 

received in both RELAP5 and CATHARE have been compared to nominal values stated in the 

Pre-Construction Safety Report of the EPR [3]. The comparison between selected calculated

parameters in both codes and those provided in the PCSR is satisfactory. Finally, transient 

calculations have been performed in accordance with the provided scenario. 

A discussion of the obtained results, run statistics and conclusions are provided at the end of

the report.

xiii

ACKNOWLEDGMENTS 

The authors acknowledge the support from the Polish National Atomic Energy Agency (PAA) 

within the SARWUT project performed within the National Centre for Research and 

Development (NCBiR) strategic project “Technologies for the development of safe nuclear

energy”, Research Task No. 9 entitled: “Development and implementation of safety analysis 

methods in nuclear reactors during disturbances in heat removal and severe accident 

conditions.”

xv

ABBREVIATIONS

ASCII American Standard Code for Information Interchange

CAMP Code Application and Maintenance Program

CATHARE The Code for Analysis of Thermalhydraulics during an Accident of Reactor and 

safety Evaluation

CEA Commissariat à l'energie atomique et aux énergies alternatives (Atomic Energy 

and Alternative Energies Commission) 

EDF Électricité de France (Electricity of France) 

EPR European Pressurized Reactor

IET Integral Effects Test

LHSI Low Head Safety Injection

LOBI Loop Blowdown Investigation

LOCA Loss of Coolant Accident

LOFT Loss of Fluid Test

LSTF Large Scale Test Faciltiy

MHSI Medium Head Safety Injection

MOX Mixed Oxide Fuel

NCBiR Narodowe Centrum Badań i Rozwoju (The National Centre for Research and 

Development) 

NPP Nuclear Power Plant

NRC Nuclear Regulatory Commission

PAA Państwowa Agencja Atomistyki (Polish Atomic Agency)

PACTEL Parallel Channel Test Loop

PCSR Pre-Construction Safety Report

PKL Primärkreislauf (German Large Scale Test Facility)

PMK Scaled-down model of the Paks Nuclear Power Plant

PWR Pressurized Water Reactor

RCS Reactor Cooling System

RELAP Reactor Excursion and Leak Analysis Program

RPV Reactor Pressure Vessel

SET Separate Effects Test

SNAP Symbolic Nuclear Analysis Package

SPES Simulatore Pressurizzato per Esperienze di Sicurezza (PWR Test Facility)

TRACE TRAC/RELAP Advanced Computational Engine

US NRC United States Nuclear Regulatory Commission

1 

1. INTRODUCTION

There is a strong need to understand the characteristics of the Loss Of Coolant Accidents 

(LOCA) as such as the automatic countermeasures of the protection and safeguard systems 

designed to provide the withstand of the NPP for any break size and break location in the 

primary circuit. The NPP’s supplier is obliged to prove that the plant parameters during LOCA 

should not violate the acceptance criteria [1]. For this objective numerous thermal-hydraulic 

codes have been developed in different countries. Among the most popular computer codes 

used for the safety analysis of the NPP are the RELAP5 (USA) and the CATHARE-2 (France). 

Warsaw University of Technology is in the process of becoming in the future a TSO has started 

developing knowledge related to accident simulation working in cooperation with the Polish 

National Atomic Energy Agency. This work is a representation of some of the efforts from 

learning and using thermal-hydraulic codes developed by the NRC and French supplied codes.

3 

2. THE EPR GENERAL DESIGN

European Pressurized Reactor (EPR in Europe) or Evolutionary Pressurized Reactor (EPR in 

US) is a Generation III large PWR design by Framatome – currently owned by AREVA. It was 

developed on the basis of French and German experiences obtained during development of the 

N4 and Konvoi PWR reactors.

The reactor is characterized by a robust design based on the proven defense in-depth concepts. 

It has a high level of redundancy – with four safety divisions and independent emergency core 

cooling lines which provide proper protection against single failure and robust cooling capability. 

Most important systems of the four safeguard divisions contain a passive accumulator and two 

active systems with low and medium pressure head injections (LHSI & MHSI) per one loop.

Plant design is characterized by a diversity of safety systems and emergency electric power and 

water supplies to strongly reduce the probability of a common cause failure. In the design there 

is a high level of complementarity in order to provide proper mix of both active and passive 

systems. 

Among many safety related means, a crucial example is a large double walled containment with 

the outer shield building made of reinforced concrete and inner containment with steel liner and 

pre-stressed concrete. It forms the final barrier between the public, the environment and 

potential radioisotopes released during reactor or spent fuel pool accidents and severe 

accidents caused by both extreme internal and external hazards.

In order to withstand severe accidents, the plant utilizes an ex-vessel retention concept with a 

dedicated core-catcher to contain corium outside the vessel. The core catcher and the reactor 

cavity are initially dry to avoid a highly energetic interaction of corium with water and avoid a 

steam explosion. In order to prevent late containment failure due to the pressurization or the

basemat melt-through, the core catcher system utilizes both passive and active water cooling 

systems with heat exchangers which are able to provide ultimate cooling and stabilization of the 

corium.

The containment is equipped with a spraying system with recirculation dedicated to reduce 

containment pressure in the case of containment pressurization. Moreover, the plant is 

equipped with a reliable RCS depressurization system which severely decreases the probability 

of high pressure melt ejection and potential early containment failure due to the Direct 

Containment Heating phenomena. Additionally, the containment is equipped with a set of 

passive autocatalytic re-combiners (PARs) forming part of a combustible gas control system 

designed to remove hydrogen gas and suppress potential deflagration or detonation [2], [3], [4], 

[5], [6].
4 

Table 1 General Design Parameters for the EPR.

General data

Core thermal power 4500 MWth

Electric power (net) 1600-1650 MWe

Number of cooling loops 4 -

Nominal primary system pressure 155 bar

RPV design pressure 176 bar

Nominal secondary system 

pressure

78 bar

Secondary side design pressure 100 bar

Fuel array 17x17 -

Number of fuel assemblies 241 -

Fuel rods per assembly 265 -

Number of control rod clusters 89 -

Basic fuel UO2 (up to 5% %wt) 

or MOX (core up to 

30%)

-

Plant lifetime 60 yr

Average discharge burn-up 55-65 GWd/MTU

Basic 

thermalhydraulics

Thermal design flow rate per one 

loop

27185 m3/h

Core bypass flow rate 5.5 %

Nominal core inlet temperature 295.6 °C

Nominal core outlet temperature 331.6 °C

Core heat transfer surface 8005 m2

Average core heat flux 0.547 MW/m2

Maximum core heat flux during 

normal operation

1.573 MW/m2

Average linear heat flux 16.34 kW/m

Maximum linear heat flux during 

normal operation

47.0 kW/m

Power density 94.6 MW/m3

Containment

Overall form Spherical cap and 

cylindrical

-

Containment volume 80 000 m3

Design pressure 5.5 Bar

Design temperature 170 °C

Design leak rate 0.3 % volume 

per day
5 

3. THE BENCHMARK EXERCISE PROCEDURE

A benchmark calculation between two thermal-hydraulic codes requires maximum 

understanding in the phenomenology, methodology, code’s specifications and experience in the 

creation of a plant model. The analytical procedure should involve the following steps:

• Step 1: Preparation of equivalent inputs

Remove the inconsistencies in the nodalization scheme, initial conditions and model’s

options as far as possible

• Step 2: Performing the steady state and transient calculations

Discrepancies in results are discussed in details to avoid the impact of unequal models in

input (break simulation and options, local pressure losses, main coolant pump characteristic,

decay power history)

• Step 3: Definition of a new basic input deck

Checking full consistencies of the new inputs what has not been done in the Step 1.

Performing new calculations

• Step 4: Comparison of results and qualifying the capability of the code based on the

simulation of accident

Comparison and assessment of results accordingly to the code’s governing equations,

specific models and features.

7 

4. THE RELAP5 THERMAL-HYDRAULIC CODE MODEL

The RELAP5/MOD3.3 code has been developed for best-estimate transient simulations of light 

water reactor coolant systems during postulated accidents. The code may be used to model the 

coupled behavior of the reactor coolant system and the core region during loss-of-coolant 

accidents and operational transients such as anticipated transients without SCRAM, loss of 

offsite power and loss of feedwater flow. A generic modeling approach is used that permits 

simulating a variety of thermal hydraulic systems. Control system and secondary system 

components are included to allow modeling of plant controls, turbines, condensers, and 

secondary feedwater systems [7], [8],[9].

4.1 RELAP5 Nodalization

Nodalization of the reactor pressure vessel is presented in Figure 1. The reactor core is 

modelled by two pipe components no. 535 &536, with 9 axial meshes, where the meshes 1 

through 9 represent the heated length of the core. The division was based on core enrichment 

differences and differentiating between the inner and outer zones. Pipe 536 has two heat 

structures modeling the purple and blue region and pipe 535 has three heat structures modeling 

the green, yellow and red regions in Figure 2. These regions represent zones of different heat 

generation rates in the reactor core.

Figure 1 RPV Nodalization Scheme. 
8 

Figure 2 Core Heating Regions With Zone Coloring.

Figure 3 Core Nodalization Scheme With Zone Coloring.

The downcomer is modeled using two annulus components: 515, 516 , branches 505 and 506 

and multiple junctions 207,208. Core bypass is modeled by a pipe component 520 and branch 

541. Lower plenum of the vessel is represented in the model by the two branches 525, 530. 

Guide tubes are represented by a one branch 500. Upper plenum of the vessel is modeled by 

two pipes 550, 556 and branches 545, 555. Each component used for core modelling is 

connected thermally via a heat structure. The lower plenum is modeled by branches 525, 530.
9 

Figure 4 Axial Power Shape As a Function of the Normalized Core Height. 

The power axial distribution shown in Figure 4 is a version used for calculations in RELAP5. It is 

a normalized linear power average versus normalized core height.

0

0.5

1

1.5

2

2.5

0.06 0.17 0.28 0.39 0.50 0.61 0.72 0.83 0.94

normalized power

normalized height

axial power shape
10

Figure 5 Primary Side Single Loop Nodalization Scheme. 

Nodalization of the primary side is presented in Figure 5. The model consists of 4 separate 

loops. Loop no. 1 is presented in Figure 3. Loops 2 and 3 and 4 are modelled identically with the 

sole difference that the pressurizer is connected to loop no. 1. Finally, loop no. 3 holds the trip 

valve (element no. 600 in Figure 1.), which has an area of 0.184m2

, in order to simulate the 

postulated accident.

The horizontal part of the hot leg connected is modeled by two elements - 100, 102. The 

pressurizer is modeled by pipe 150 consisting of 8 nodes.
11

Figure 6 Steam Generator Secondary Side Nodalization Scheme. 

The secondary side nodalization scheme is presented in Figure 6. The inlet plenum for the 

steam generator is modeled with a single volume 106. U-tube bundles are modeled with pipe 

108 also having 8 nodes, the outlet plenum – single volume 110.

The water supply for the steam generator secondary side is flowing through the time dependent 

volume 182 and junction 181. Branch 174 mixes water returning from the steam separator 171 

and is also connected to the downcomer 176 modeled with an annulus. The downcomer has 5 

nodes. The mixture is then transported to the riser pipe 170. The riser has a heat structure 4 out 

of 6 nodes connecting the primary and secondary sides.

The most outer part of the steam separator is a single volume 172 and branches 178, 180 and 

single volume 120 is used to model the dryer and steam dome.

The rest of the secondary side is a steam line modeled with pipe 122. Valve 185 simulates the 

cutoff valve to the steam collector. The steam collector pipe 902 as well as the time dependent 

volumes for the turbine 904 and the condenser 906 close out the steam circuit. Additionally the 

model also has an isolation valve for the turbine 901 and steam bypass valve 903 from which 

steam goes directly to the condenser.

The break is modeled with two valves which are located on the second loop, the first one just 

after the reactor coolant pump and the second one at the cold leg just before the reactor vessel. 

The containment is modeled by a single volume component.

4.2 Safety Systems

The primary and secondary side are equipped in emergency cooling systems which can be 

used during abnormal work conditions. Those systems are:

On the primary side:

• Lower head safety injection (LHSI) – 2 injection systems (on broken loop and pressurizer

loop)

• Medium head safety injection (MHSI) – 2 injection systems (on broken loop and

pressurizer loop)

• Accumulators injection – four acculumators each having 31.74 m3 of liquid opening at

pressure 45 bar.
12

On the secondary side:

• Emergency steam generator feedwater system.

The injection schemes are shown in Figure 7 for the medium and low head injection systems. 

The mass flow for the steam generator injection is constant and is equal to 25 kg/s.

Figure 7 MHSI and LHSI Mass Flow Rate Curves. 
13

5. CATHARE ANALYTICAL METHOD

5.1 CATHARE-2 Thermal-Hydraulic Code

The CATHARE second generation thermal hydraulic code has been developed jointly by the 

CEA, EDF and AREVA NP to carry out safety analyses. CATHARE is a modular, two fluids 

code, capable of modeling mock-ups as well as entire Pressurized Water Reactors.

The approach adopted for the physical validation of CATHARE can be broken down into two 

tasks, which are: 

1. Qualification in analytical tests or separate effects tests (SETs)

2. Verification in global experiments or integral effects tests (IETs).

The matrix of SETs, which is used for qualification, brings together about 300 tests chosen from 

experiments regarding critical flow, determination of flow diagrams, depressurization of

adiabatic or hot test sections in various geometries, reflooding, filling the downcomer, phase

separation at Tee junctions, counter-current for complex geometries, the response of steam

generators and reactor coolant pumps and the thermo mechanics of the fuel rod.

The matrix of IETs, which is used for verification, is made up of 27 tests carried out on BETHSY, 

LOBI, LSTF, PACTEL, PMK, LOFT, PKL and SPES mock-ups [10-12].

5.2 CATHARE Nodalization

The EPR model in CATHARE has been developed by experts from AREVA and has been 

provided to the SARWUT project as a reference model. The model consists of four loops which 

are modeled separately. Safety systems, the medium head injection system and the low head 

injection system are modeled with the use of gadget components, however during the transient 

they operate only in the broken loop and pressurizer. Four accumulators are also included. All 

safety systems are connected to the cold legs at the distance of around 5 m from the reactor 

pressure vessel (RPV). The break is modeled with the use of boundary conditions which are

blind during steady state. The break is located 5 m from the RPV. The pressurizer is a single

volume element which is attached to the hot leg by the surge line (one axial element) at the 

distance of 5.5 m from the RPV. 

There are three boundary conditions associated with the pressurizer in order to appropriately

model the safety valves functioning. The pressurizer can be approximated by one element, due 

to the fact that it gets empty at the early stages of transient calculations and therefore it has no 

impact during the latter stage of transient. 

The RPV is modeled with the use of 9 components, five of which are volume components and 4 

of them are axial components. The coolant flows into the inlet plenum and is divided into 2 

streams. During steady state calculations it is assumed that less than 1 % of the mass flow

enters the upper head of the RPV. The water from the downcomer, which is modeled as an 

axial component, enters the lower volume. The lower volume models the lower plenum of the 

RPV and the free volume of the lower core support structure. About 95% of the total flow of the 

downcomer flow enters the reactor core during the steady state calculations and 5 % bypasses 

the core through the two axial components. 
14

The core is divided into 59 segments but only 55 of them model the active part of the core. The 

reactor power is set as a function of time. The characteristics of the mentioned 55 fuel segments 

are the same except for the axial peaking factor which differs at each elevation, simulating in 

this way the proper power distribution in the core. 

The coolant accumulates in the outlet plenum modelled by one volume. The free volume of the 

guide tubes is also modelled by one volume located above. The connection with the steam 

dome is possible only via the guide tubes and a connection with the inlet plenum. The steam 

generator is modeled using 6 components. The downcomer and the riser part below the U-tube 

bend is divided into 2 parallel axial components. One simulates the co-current part of the U-tube 

heat exchanger and the other simulates the countercurrent part of the U-tube heat exchanger. 

This simulates the economizer in the steam generator. The mixture is accumulated in a very 

small volume and is distributed to the axial element which simulates a U-tube bending and a 

riser above bending. 

The steam separator model is set at the junction which connects the riser with a volume

simulating the steam dome and the free volume of the separator. The steam flows through an 

axial pipe and is accumulated in the volume with boundary conditions set to simulate the 

turbine. The nodalization scheme of the RPV and the primary loop are shown on Figures 6. and 

7. respectively. The nodalization of the secondary side is presented in Figure 8.

Figure 8 Reactor Pressure Vessel Nodalization Scheme. 
15

Figure 9 Primary Loop Nodalization Scheme. 

Figure 10 Nodalization Scheme of the Secondary Side. 
16

5.3 Critical Flow Model in CATHARE

The implemented flow models in CATHARE are able to calculate precisely the two-phase flow 

situations such as stratification flow, counter current flow and the critical flow.

The equation to determine the critical mass flow rate consists of several parameters such as the 

mixture density, the void fraction, the ratio between length and the hydraulic diameter, the 

pressure losses and the difference between the pressure of the liquid and the saturation 

pressure at given temperature. This difference can be approximated by the correlations which 

depend on the liquid temperature. If the liquid temperature is lower than the saturated 

temperature at given pressure (single phase liquid), then the difference is taken from the 

equation base on the given pressure and the temperature in equilibrium condition.

The Critical model implemented in CATHARE has been validated against several experiments 

like Super Moby Dick, Bethsy, Marviken and Rebeca, in a wide range of pressure and different 

ratios between length and hydraulic diameter of a discharge pipe for subcooled and saturated 

fluid and mixture of water, steam and air [11], [12].
17

6. CALCULATIONS

6.1 Scenario Description

The scenario taken into consideration assumes a 6-inch in diameter break on the cold leg at the 

distance of about 5 m from the RPV.

Only two MHSI and two LHSI are available and work without delays. All available accumulators 

are available. Decay heat tables are used for core power calculation after reactor trip and 

constant power is assumed beforehand. Pump coastdown is delayed based on a function 

comparing primary side pressure as obtained from the CATHARE input deck. On the secondary 

side a partial cooldown procedure is modeled which simulates cooling down at a rate of 250K/h.

6.2 Steady-State Results

A summary of nominal conditions, taken from [2] is presented in Table 2. It includes also the 

results obtained in the steady-state simulations. Both RELAP5 and CATHARE predictions of 

steady-state working conditions are similar and compared to nominal conditions give good 

agreement. The pressurizer pressure and water level are slightly overestimated in CATHARE. 

On the other hand, the average temperature values in the primary side as well as the secondary 

pressure calculated by RELAP5 are a little higher than nominal. These results were used as 

initial conditions for LOCA simulations.

Table 2 Steady-State Results Calculated in RELAP5 and CATHARE.

Nominal RELAP5 CATHARE

Primary 

side

Pressurizer pressure MPa 15.5 15.49 15.77

Saturation temperature in 

pressurizer

0

C 345 344.77 346

Pressurizer water level % 56 55.5 61.8

RPV inlet temperature 0

C 295.6 301.6 297

RPV outlet temperature 0

C 329.8 334 330.7

Average temperature 0

C 312.7 317.8 313.9

Total coolant flow rate kg/s 22 235 22 177 21 931

Bypass flow rate % 5.5 5.6 5.1

Secondary 

side

Feedwater temperature 0

C 230 230 230

Steam generator narrow 

water level m 49 47.54 48.7

Steam pressure MPa 7.71 7.77 7.72

Total main steam flow 

rate kg/s 2 552.4 2 531.22 2 555
18

6.3 Transient Results 

Table 3. presents a sequence of events occurring during the simulation. There is a good 

comparison between RELAP5 and CATHARE predictions. The reactor is tripped after 16 s from 

the beginning of the simulated transient. Immediately after that, the turbine is tripped and the 

main steam bypass valves are opened. Approximately after 30 s, the pressure in the steam 

generator’s steam dome reaches 9.6 MPa, the partial cooldown is initiated and lasts for about 

450 s. Next, the pressurizer becomes empty and the medium head injection (MHSI) pump starts

working. Injection from the accumulator tanks starts at 670 s in RELAP5 and 595 s in 

CATHARE. This is the most significant difference in the transient calculations. The main coolant 

pumps are tripped after 79 s. Calculations are finished at 1 000 s.

Table 3 Sequence of Events.

RELAP5 CATHARE Event

0 0 Leak opening in cold leg

16.5 16.72 Reactor trip

18.7 18.82 Turbine trip

18.7 18.82 Opening of main steam bypass valves

27 31.41 SG 9.6 MPa, Safety Valve open,

Partial CoolDown Start, 250 0C/h

74 45 Pressurizer is empty

78 79.6 Main coolant pumps trip

188 200 Beginning of MHSI injection

485 500 End of partial cooldown

668 595 Beginning of accumulator injection

1 000 1 000 End of calculation

Figures 11. to 27. represent the results of the benchmark calculation between the two codes. 

The CATHARE results are illustrated with a blue color and RELAP5 with a red color.

Figure 11 Pressure in Core.
19

Figure 12 Pressurizer Pressure.

Figure 13 Water Level in the Pressurizer.
20

Figure 14 Water Level in the Pressurizer (First 100 s).

Figure 15 Surge Line Mass Flow.
21

Figure 16 Steam Generator Pressure.

Figure 17 Steam Generator Water Level.
22

Figure 18 Pump Speed.

Figure 19 MHSI Mass Flow Rate. 
23

Figure 20 LPSI Mass Flow Rate. 

Figure 21 Reactor Core Water Level. 
24

Figure 22 Cladding Temperature.

Figure 23 Void Fraction at the Break. 
25

Figure 24 Integrated Break Mass Flow Rate. 

Figure 25 Break Mass Flow Rate. 
26

Figure 26 Accumulator Mass Flow Rate. 

Figure 27 Integrated Surge Line Mass Flow Rate. 
27

7. DISCUSSION

The results shown are generally in good comparison. Most differences can be assigned to a 

different model for critical two-phase flow in both codes as well as different pump degradation 

curves and multipliers. As for the critical two-phase flow, CATHARE uses a six equation model 

where in RELAP5 the standard Henry-Fauske model is used. 

Figure 11. shows that pressure in the core has a sharper downward spike in RELAP5 about 50 

seconds into the transient but both pressures align after the 100th second and RELAP5 predicts 

higher pressure after 400 seconds till the end of the calculation. Figure 23. showing void fraction 

shows that until the 200th second RELAP5 predicts a constant rise of void fraction at the break 

where in CATHARE between the 50th and 100th second steam content is dropping even though, 

at that time, no additional water source (e.g. safety injection) is active. Additionally, CATHARE 

predicts a slightly higher integrated break mass flow rate than RELAP5. 

The pressurizer empties 30 seconds later in RELAP5 mainly due to the last 5% of inventory 

being cleared slowly. Figure 15. shows an interesting behavior of the mass flow rate through the 

surge line where though the flows are comparable, the two peaks occurring in the flow have a 

reversed order (high-to-low in CATHARE, low-to-high in RELAP5). Figure 16. shows the steam 

generator pressure at the secondary side which is controlled directly in CATHARE by a routine 

that sets steam generator pressure (after breaching 9.6 MPa) using a function representing a 

partial cooldown procedure of 250 0

C/h. 

An additional time-dependent-volume (per each loop) connected to the steam generator has 

been added in RELAP5 to mimic the same behavior as in CATHARE. The difference visible in 

Figure 17. concerning the steam generator water level can be attributed to the difference in the 

representation of geometry in both codes. The methodology for steam generator modelling is 

also different and therefore this result might not be comparable. 

The pump speeds shown in Figure 18. are in good alignment up to the 200th second where the 

slope in RELAP5 is much steeper. The MHSI in Figure 19. starts at almost the same time but 

because of being driven by pressure the flow rate in RELAP5 is a bit smaller; LHSI does not 

start in this transient calculation. Figure 21. presenting the reactor core water level is very 

closely predicted up to the 200th second and though there is a difference of about 0.5 meter, the 

trend shows increasing temperature in RELAP in the last part of the transient.

Pump data was an issue due to an inability to access CATHARE’s built in curves and two-phase 

multipliers. An issue was also observed with RELAP5 accumulator discharge flow where a 

short, sharp spike of mass flow can be seen in Figure 26. Figure 16. shows an implementation 

of the partial cooldown which is slightly different for RELAP5 and CATHARE due to a difference 

in logic controllers in CATHARE resulting from direct forcing of secondary pressure behavior.

29

8. RUN STATISTICS

The calculations were performed using Intel® CoreTM i5 M 560 @ 2.67 GHz processor. The 

operating system was Windows 7 Professional. 

Table 4. shows the run statistics for the codes RELAP5/MOD3.3 Patch 0.4 and CATHARE 2.5

calculations. Both times are comparable.

Table 4 Run Statistics

Code Transient Time

(s)

CPU Time

(s)

CPU/Transient 

Time

Number of Time 

Steps

CATHARE

1 000.00 986.40 0.9864

Time steps:6006/ 

Iterations:22380

RELAP5/MOD3.3 

Patch 4 1 000.01 1 062.74 1.063 109364

31

9. CONCLUSIONS

This benchmark activity is very valuable in terms of user experience in system modeling as both 

codes, though very similar, enforce different modelling of some critical components – e.g. the 

steam generator or safety system logic – e.g. partial cooldown. Achieving comparable results is 

an iterative process which demands in-depth understanding of both codes: RELAP5 and 

CATHARE.

Though there are differences in the results obtained from CATHARE and RELAP5, they can be 

attributed to factors related to lack of precise data such as pump data. Another factor can be 

related to differences in physical phenomena modelling such as critical two-phase flow in the 

codes. Finally, different representation of geometry of some components (e.g. the pressurizer in 

CATHARE has a trapezoid base compared to a rectangular RELAP5 shape), might also lead to 

discrepancies in the results of both codes. 

33

10. REFERENCES

1. Pressurized water reactor (PWR) systems, Reactor Concepts Manual, NRC.

2. Areva and EdF, UK EPR Pre-construction Safety Analysis Report, 2013

3. http://www.epr-reactor.co.uk/

4. Areva, US EPR Final Safety Analysis Report, 2013

5. http://www.nrc.gov/reactors/new-reactors/design-cert/epr.html

6. Computational validation of the EPR combustible gas control system, H. Dimmelmeier,

et. al, Nuclear Engineering and Design, 249 (2012), 118-124.

7. M. Fischer, The severe accident mitigation concept and the design measures for core

melt retention of the European Pressurized Reactor (EPR), Nuclear Engineering and

Design, 230 (2004), 169-180.

8. EPR safety in post-Fukushima context, Areva, March 2012, Presentation.

9. H. Ransom, et.al., RELAP5/MOD3 Code manual volume 1: Code structure, system

models, and solution method.,1994.

10. RELAP5 MOD3.3 Code manual volume 4: models and correlations, October 2010,

Information Systems Laboratories, Inc..

11. S. Shieh V. H. Ransom, RELAP5/MOD3 Code manual volume 6: Validation of numerical

techniques in RELAP5/MOD3.0,1994.

12. Eymard, C. (2005). CATHARE2 V2. 5 User manual No. SSTH/LDAS/EM/2004-040).

Grenoble, France: CEA.

13. Lavialle, G. L. (2005.). cathare2 v2.5_1:User guidelines No. SSTH/LDAS/EM/2004-067).

Grenoble, France, CEA.

14. Darona, J. (2006). cathare2v2.5_2mod8.1: Dictionary of directives and operators, CSSI.





NUREG/IA

-0459 EPR Medium Break LOCA Benchmarking Exercise 

Using RELAP5 and CATHARE