NUREG/IA-0250 

 International 

 Agreement Report

 

Simulation of the F2.1 Experiment at 

PKL Facility using RELAP5/MOD3

Prepared by: 

S. Carlos, J.F. Villanueva, S. Martorell, 

V. Serradell 

Universidad Politécnica de Valencia 

Cami de Vera s/n 

46022 Valencia, SPAIN 

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

February 2011 

 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

Under the International Code Assessment and Maintenance Program (CAMP) 

 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0250

 

Simulation of the F2.1 Experiment at PKL Facility

using RELAP5/Mod3 

 

 

 

 

 

Manuscript Completed: September 2010

Date Published: February 2011

Prepared by:

S. Carlos, J.F. Villanueva, S. Martorell, V. Serradell

Universidad Politécnica de Valencia

Camí de Vera s/n

46022 Valencia, SPAIN

Antony Calvo, NRC Project Manager

Prepared for:

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

iii

ABSTRACT

When a nuclear power plant is in shutdown conditions for refuelling, the reactor coolant system 

water level is reduced. This situation is known as mid-loop operation, and the residual heat removal 

(RHR) system is used to remove the decay power heat generated in the reactor core.

In mid-loop conditions, some accidental situations may occur with not a negligible contribution to 

the plant risk, and all involve the loss of the RHR system. Thus, to better understand the thermalhydraulic processes following the loss of the RHR during shutdown, transients of this kind have 

been simulated using best-estimate codes, comparing their results against experimental data taken 

from different integral test facilities. This paper focuses on the simulation, using the best estimate 

code RELAP5/Mod 3.3, of the experiment F2.1 conducted at the PKL facility, within the OECD/PKL 

project. This experiment consists of the loss of the RHR system when the plant is in mid-loop 

conditions for refuelling and with the primary circuit closed. In the experimental series F2.1the 

physical phenomena to investigate are the mechanisms of heat removal in presence of nitrogen 

and the deboration in critical parts of the primary system.

Two experiments belonging to this experimental series have been performed. The simulations 

present differences in the initial plant coolant inventory and temperature in the pressurizer, 

F2.1RUN1 and F2.1RUN2, to asses the influence of these differences in the transient evolution. 
iv
v

FOREWORD 

Extensive knowledge and techniques have been produced and made available in the field of thermal-hydraulic 

responses during reactor transients and accidents, and major system computer codes have achieved a high

degree of maturity through extensive qualification, assessment and validation processes. Best-estimate 

analysis methods are increasingly used in licensing, replacing the traditional conservative approaches. Such 

methods include an assessment of the uncertainty of their results that must be taken into account when the 

safety acceptance criteria for the licensing analysis are verified. 

Traditional agreements between the Nuclear Regulatory Commission of the United States of America 

(USNRC) and the Consejo de Seguridad Nuclear of Spain (CSN) in the area of nuclear safety research have 

given access to CSN to the NRC-developed best estimate thermalhydraulic codes RELAP5, TRAC-P, TRACB, and currently TRACE. These complex tools, suitable state-of-the-art application of current two-phase flow 

fluid mechanics techniques to light water nuclear power plants, allow a realistic representation and simulation 

of thermalhydraulic phenomena at normal and incidental operation of NPP. Owe to the huge required 

resources, qualification of these codes have been performed through international cooperation programs. 

USNRC CAMP program (Code Applications and Maintenance Program) represents the international 

framework for verification and validation of NRC TH codes, allowing to:

• Share experience on code errors and inadequacies, cooperating in resolution of deficiencies and 

maintaining a single, internationally recognized code version.

• Share user experience on code scaling, applicability, and uncertainty studies.

• Share a well documented code assessment data base.

• Share experience on full scale power plant safety-related analyses performed with codes (analyses of 

operating reactors, advanced light water reactors, transients, risk-dominant sequences, and accident 

management and operator procedures-related studies).

• Maintain and improve user expertise and guidelines for code applications.

Since 1984, when the first LOFT agreement was settled down, CSN has been promoting coordinated joint 

efforts with Spanish organizations, such as UNESA (the association of Spanish electric energy industry) as 

well as universities and engineering companies, in the aim of assimilating, applying, improving and helping the 

international community in the validation of these TH simulation codes1, within different periods of the 

associated national programs (e.g., CAMP-España). As a result of these actions, there is currently in Spain a 

good collection of productive plant models as well as a good selection of national experts in the application of 

TH simulation tools, with adequate TH knowledge and suitable experience on their use.

Many experimental facilities have contributed to the today’s availability of a large thermal-hydraulic database 

(both separated and integral effect tests). However there is continued need for additional experimental work 

and code development and verification, in areas where no emphasis have been made along the past. On the 

basis of the SESAR/FAP2 reports “Nuclear Safety Research in OECD Countries:Major Facilities and 

Programmes at Risk” (SESAR/FAP, 2001) and its 2007 updated version “Support Facilities for Existing and 

Advanced Reactors (SFEAR) NEA/CSNI/R(2007)6”, CSNI is promoting since 2001 several collaborative 

international actions in the area of experimental TH research. These reports presented some findings and 

recommendations to the CSNI, to sustain an adequate level of research, identifying a number of experimental 

facilities and programmes of potential interest for present or future international collaboration within the safety 

community during the coming decade. 

CSN, as Spanish representative in CSNI, is involved in some of these research activities, helping in this 

international support of facilities and in the establishment of a large network of international collaborations. In 

 1 It’s worth to note the emphasis made in the application to actual NPP incidents.

2 SESAR/FAP is the Senior Group of Experts on Nuclear Safety Research Facilities and Programmes of NEA Committee on the 

Safety of Nuclear Installations (CSNI).
vi

the TH framework, most of these actions are either covering not enough investigated safety issues and 

phenomena (e.g., boron dilution, low power and shutdown conditions), or enlarging code validation and 

qualification data bases incorporating new information (e.g., multi-dimensional aspects, non-condensable gas 

effects). In particular, CSN is currently participating in the PKL and ROSA programmes.

The PKL is an important integral test facility operated by of AREVA-NP in Erlangen (Germany), and designed 

to investigate thermal-hydraulic response of a four-loop Siemens designed PWR. Experiments performed 

during the PKL/OECD program have been focused on the issues:

• Boron dilution events after small-break loss of coolant accidents.

• Loss of residual heat removal during mid-loop operation (both with closed and open reactor coolant 

system.

ROSA/LSTF of Japan Atomic Energy Research Institute (JAERI) is an integral test facility designed to simulate 

a 1100 MWe four-loop Westinghouse-type PWR, by two loops at full-height and 1/48 volumetric scaling to 

better simulate thermal-hydraulic responses in large-scale components. The ROSA/OECD project has 

investigated issues in thermal-hydraulics analyses relevant to water reactor safety, focusing on the verification 

of models and simulation methods for complex phenomena that can occur during reactor transients and 

accidents such as:

• Temperature stratification and coolant mixing during ECCS coolant injection

• Water hammer-like phenomena

• ATWS

• Natural circulation with super-heated steam

• Primary cooling through SG depressurization

• Pressure vessel upper-head and bottom break LOCA

This overall CSN involvement in different international TH programmes has outlined the scope of the new 

period of CAMP-España activities focused on:

• Analysis, simulation and investigation of specific safety aspects of PKL/OECD and ROSA/OECD 

experiments.

• Analysis of applicability and/or extension of the results and knowledge acquired in these projects to 

the safety, operation or availability of the Spanish nuclear power plants.

Both objectives are carried out by simulating experiments and plant application with the last available versions 

of NRC TH codes (RELAP5 and TRACE). A CAMP in-kind contribution is aimed as end result of both types of 

studies.

Development of these activities, technically and financially supported by CSN, is being carried out by 5 

different national research groups (Technical Universities of Madrid, Valencia and Cataluña). On the whole, 

CSN is seeking to assure and to maintain the capability of the national groups with experience in the thermal 

hydraulics analysis of accidents of the Spanish nuclear power plants.

________________________________

Francisco Fernández Moreno, Commissioner

Consejo de Seguridad Nuclear (CSN)
vii

CONTENTS

Page

Abstract .................................................................................................................................iii

Foreword..................................................................................................................................v

Executive Summary...............................................................................................................xi

Acknowledgements ............................................................................................................. xiii

Abbreviations ........................................................................................................................xv

1. Introduction ...................................................................................................................1-1

2. PKL Facility description ...............................................................................................2-1

3. F2.1 Transient description ...........................................................................................3-1

4. RELAP-5 Model of PKL facility ....................................................................................4-1

5. F2.1 simulation results .................................................................................................5-1

5.1. F2.1 Run 1 simulation results ..................................................................................5-1

5.2. F2.1 Run 2 simulation results ..................................................................................5-8

5.3. Comparison of F2.1 Run 1 and F2.1 Run 2 plant evolution simulation.................. 5-14

6. Run Statistics ................................................................................................................6-1

7. Conclusions...................................................................................................................7-1

8. References.....................................................................................................................9-1
viii

Figures

Page

Figure 1. PKL Facility 2-1

Figure 2. Experimental evolution of pressures experiment F2.1 RUN 1 3-2

Figure 3. Experimental evolution of pressures experiment F2.1 RUN 2................................. 3-3

Figure 4. RELAP5 nodalization of PKL facility 4-2

Figure 5. Water level inlet side of SG1 U-tubes 5-2

Figure 6. Water level in the U-tubes of steam generator 2 5-3

Figure 7. Calculated heat exchange through the steam generators 5-3

Figure 8. Secondary side pressure in steam generator 1 5-4

Figure 9. Secondary side pressure in steam generator 2 5-5

Figure 10. Primary pressure 5-5

Figure 11. Reactor vessel level 5-6

Figure 12. Pressurizer level 5-7

Figure 13. Boron concentration in loop seal one 5-7

Figure 14. Inlet side of SG1 U-tubes 5-9

Figure 15. Inlet side of SG2 U-tubes 5-9

Figure 16. Pressure in the SG1 secondary side ................................................................... 5-10

Figure 17: Pressure secondary side SG2 ............................................................................. 5-11

Figure 18. Calculated heat exchange through the steam generators ................................... 5-11

Figure 19. Primary pressure evolution .................................................................................. 5-12

Figure 20 Primary level ........................................................................................................ 5-13

Figure 21. Pressurizer level .................................................................................................. 5-13

Figure 22. Boron concentration at the loop seal 1 ................................................................ 5-14

Figure 23. F2.1Run1 and Run2 primary pressure calculation .............................................. 5-15

Figure 24. F2.1Run1 and Run2 pressurizer level calculation ............................................... 5-16

Figure 25. F2.1Run1 and Run2 vessel level calculation....................................................... 5-16

Figure 26. F2.1Run1 and Run2 boron concentration calculation.......................................... 5-17
ix

Tables

Page

Table 1. F2.1 Run1 and Run2 initial conditions 3-1

Table 2: Experimental and calculated initial conditions for F2.1 RUN1. 5-1

Table 3. F2.1 RUN2 initial conditions 5-8

Table 4: Run Statistics F2.1 RUN 1 6-1

Table 5: Run Statistics F2.1 RUN 2 6-1
x
xi

EXECUTIVE SUMMARY

When a Pressurized Water Reactor is in shutdown conditions for refuelling, and to perform steam 

generator U-tubes and reactor coolant pump maintenance activities, the reactor coolant system 

water level is reduced to a height lower than the top of the hot leg pipe. Under these conditions, it is 

said that the plant is in mid-loop operation. In this mode of operation, the residual heat removal

system is used to remove the decay power heat generated in the reactor core. Some accidental 

situations may occur in mid-loop conditions that have a significant contribution to the plant risk, and 

all involve the loss of the RHR system. In fact, the loss of RHR has been experienced several times 

in pressurized water reactor plants. For these reasons, the study of transients in mid-loop operation 

is of great interest to analyze the plant safety.

To better understand the thermal-hydraulic processes following the loss of the RHR during 

shutdown, transients of this kind have been simulated using best-estimate codes such as RELAP5 

or CATHARE etc. Such codes have initially been developed to simulate full power operation 

conditions, which are different physical conditions from the ones faced in mid-loop operation mode. 

Thus, to assess the capability of best estimate codes in simulating the physical phenomena under 

mid-loop conditions it is necessary to compare the code calculations with data obtained from 

experiments simulating such type of conditions.

The work presented in this paper is focused on the simulation, using the best estimate code 

RELAP5/Mod 3.3, of the experiment F2.1 conducted at the PKL facility. The experiment F2.1 

belongs to an experimental series established in the OECD/PKL program devoted to the study of 

boron dilution sequences and the effect of the primary coolant inventory in shutdown conditions, 

when the RHR system is lost and the plant is in mid-loop conditions for refuelling and with the 

primary circuit closed. F2.1 experiment is composed by three runs which differences in the initial 

plant conditions. In this work two runs have been simulated that differ in the amount of water inside 

the reactor coolant circuit and in the pressurizer temperature. Thus, in experiment F2.1 Run1 the 

primary level is initially fixed to 3⁄4 loop and the pressurizer temperature ranges into 150-170oC, 

while in experiment F2.1 Run2, the coolant inventory is reduced to the lower edge of the primary 

circuit and the pressurizer temperature ranges between 25-45 oC.

From the results obtained in both simulations it can be concluded that, in general, the calculations 

reproduce the physical phenomena observed in the experimental data, although some differences 

are observed, especially in the reactor coolant circuit mass distribution. In addition when comparing

the calculations obtained in F2.1 Run1 and F2.1 Run2, it is observed that small differences in the 

plant initial conditions influences the plant behaviour after the RHR is lost, in fact for both cases the 

plant reaches a stable situation but, as it happens in the experiment, the plant stabilization values 

are different in each case. 
xii
xiii

ACKNOWLEDGEMENTS

This work contains findings that were produced within the OECD-NEA/PKL Project. The authors are 

grateful to the Management Board of the PKL Project for their consent to this publication, and thank 

the Spanish Nuclear Regulatory Body (CSN) for the technical and financial support under the 

agreement STN/1388/05/748.
xiv
xv

ABBREVIATIONS

ACC Accumulator

ATWS Anticipated transient without scram

C Total number of volumes

CAMP Code Assessment and Management Program

CL Cold Leg

CPU Execution time (s)

CSN Consejo de Seguridad Nuclear (Spanish nuclear regulatory commission)

DT Total number of time steps

ECCS Emergency Core Cooling System

HL Hot Leg 

HPIS High Pressure injection system

JAERI Japan Atomic Energy Research Institute

kg kilogram(s)

kg/s kilograms per second

LBLOCA Large Break Loss of Coolant Analysis

LPIS Low Pressure injection system

m meter(s)

Pa Pascal

oC degrees Celsius

oK degrees Kelvin

MWe megawatt(s) electric

MWt megawatt(s) thermal

NEA Nuclear Energy Agency

NPP nuclear power plant

NRC U.S. Nuclear Regulatory Commission

OECD Organisation for Economic Cooperation and Development

ppm parts per million

PWR Pressurized Water Reactor

RCL Reactor Coolant Line

RCS Reactor Coolant System

RELAP Reactor Excursion and Leak Analysis Program

RHR Residual Heat Removal System

RT Transient time (s)

s second(s)

SBLOCA Small Break Loss of Coolant Analysis

SG Steam Generator

TH Thermal-hydraulic

TS Maximum time step (s)

UNESA Asociación Española de la Industria Eléctrica
xvi
1-1

1. INTRODUCTION

When a Pressurized Water Reactor (PWR) is in shutdown conditions for refuelling, and to perform 

steam generator (SG) U-tubes and reactor coolant pump maintenance activities, the reactor coolant 

system water level is reduced to a height lower than the top of the hot leg pipe [1]. Under these 

conditions, it is said that the plant is in mid-loop operation. In this mode of operation, the residual 

heat removal (RHR) system is used to remove the decay power heat generated in the reactor core.

Some accidental situations may occur in mid-loop conditions that have a significant contribution to 

the plant risk, and all involve the loss of the RHR system. The three major causes of a loss of the 

RHR system are: A loss of reactor coolant system (RCS) inventory, a loss of RHR flow and a loss of 

support systems [2]. Moreover, the loss of RHR has been experienced several times in pressurized 

water reactor (PWR) plants as, for example, in Diablo Canyon Unit 2 [3] and in Voltge Unit 1 [4]. 

The causes of the loss of the RHR in those plants were a failure in the RHR pump and a loss of offsite power, respectively.

For the reasons above exposed, the study of transients in mid-loop operation has become of great 

interest during the last decades [1, 5, 6, 7, 8, 9, 10, 11].

To better understand the thermal-hydraulic processes following the loss of the RHR during 

shutdown, transients of this kind have been simulated using best-estimate codes such as RELAP-5 

[5, 6; 9; 10] or CATHARE [7] etc. Such codes have initially been developed to simulate full power 

operation conditions, which have different physical conditions from the ones one faces in mid-loop 

operation mode. Thus, to assess the capability of best estimate codes in simulating the physical 

phenomena under mid-loop conditions it is necessary to compare the code calculation with data 

obtained from experiments simulating such type of conditions. 

Different experiments simulating transients in shutdown conditions have been conducted at different 

integral facilities, such as ROSA, BETHSY and PKL. This paper focuses on the simulation, using 

the best estimate code RELAP5/Mod 3.3 [12], of the experiment F2.1 conducted at the PKL facility. 

The experiment F2.1 belongs to an experimental series established in the OECD/PKL program 

devoted to the study of boron dilution sequences and the effect of the primary coolant inventory in 

shutdown conditions [11], when the RHR system is lost and the plant is in mid-loop conditions for 

refuelling and with the primary circuit closed.

The aim of the study presented in this paper is to determine the physical phenomena observed in 

experiment F2.1 that are well predicted by RELAP5-Mod3.3 code, as well as to identify those 

processes that cannot be reproduced with the used code and model.

In particular, in this experiment the physical phenomena to investigate are the mechanisms of heat 

removal in presence of nitrogen and the deboration in certain parts of the primary system.

The rest of the paper is organized as follows: The PKL facility is briefly described in Section 2. 

Section 3 is devoted to introduce both experiments F2.1RUN1 and F2.1RUN2, section 4 is devoted 

to describe the RELAP5/Mod 3.3 model for the PKL facility used to simulate the experiment. In 

section 5, the main results obtained from both simulations are presented and compared with the 

experimental data. Finally, the main conclusions of this study are summarized in section 6. 
1-2
2-1

2. PKL FACILITY DESCRIPTION

The PKL test facility represents a typical 1300MWe Siemens/KWu design PWR with a volume and 

power scale of 1:145, while all the components height on the primary and secondary side 

correspond to real plant dimensions. It models the entire primary system and the relevant parts of 

the secondary side. In order to investigate the influence of non symmetrical boundary conditions on 

the system behaviour, PKL facility is equipped with four primary loops symmetrically arranged 

around the reactor pressurized vessel. Each loop contains a reactor coolant pump and a steam 

generator [11]. 

The facility also models all the important safety and auxiliary systems as eight accumulators, one in 

each of the hot legs and one in each of the cold legs, four independent injections from the high and 

low pressure injection system, the residual heat removal system and the pressure control in the 

presssurizer. Figure 1 shows an overview of PKL test facility.

Three experimental programs have been conducted at PKL facility. Programs PKL I and PKL II, 

focused on the study of Large Break Loss of Coolant Accidents (LBLOCAs) and Small Break Loss 

of Coolant Accidents (SBLOCAs) with the objective of best estimate codes test and validation. PKL 

III program started in 1986 [11], with the objective of studying different transients with and without 

LOCAs. The PKL tests results have also been used for preparation and verification of procedures 

described in the operating manuals and for answering questions posed by regulatory bodies. 

Figure 1. PKL facility.
2-2

In particular, PKL III F series include investigations on the inadvertent boron dilution events and on 

the effect of the primary coolant inventory in transients under shutdown conditions. The importance

of the boron dilution events lies on the possibility that low borated water may enter into the reactor 

vessel, and this may lead to a local reactivity insertion event in the reactor core.

Moreover, the initial situation of the plant may affect the plant evolution towards a safe situation 

when the RHR is lost in mid-loop conditions. Thus, series F2.1 consist of three experiments with 

differences in the reactor coolant inventory and in the pressurizer initial conditions. In this work two 

of these experiments are simulated with different amount of water in the primary and different 

temperature in the pressurizer.
3-1

3. F2.1 TRANSIENT DESCRIPTION

In shutdown conditions, one of the most important initiating events is the loss of the RHR system. 

When this happens, other alternatives for heat removal should be available to cooldown the reactor 

core. One possibility consists of using the steam generators as heat sink, by means of using the 

cooling capacity of the reflux condensation. This phenomenon takes place in the U-tubes of the 

steam generators with the secondary side full of water. The vapour generated in the core comes 

into the U-tubes and transfers the heat to the steam generator secondary side. This vapour 

condenses in the U-tubes and comes back again into the reactor vessel cooling the system.

The experimental series F2.1 conducted in the PKL facility consists of the loss of the RHR system 

when the plant has been shut down for refuelling, with the primary circuit closed and partially filled. 

Two of the steam generators (SG1 and SG2) have their secondary sides full of water and the other 

two (SG3 and SG4) are full of nitrogen. This paper focuses on the simulation of experiment F2.1 

Run1 and F2.1 Run2. Both experiments have differences in the plant initial conditions. In 

experiment F2.1 Run1 the primary level is initially fixed to 3⁄4 loop and the pressurizer wall 

temperature ranges into 136-156oC (hot pressurizer) in the N2 filled region, while in experiment F2.1 

Run2, the coolant inventory is reduced to the lower edge of the primary circuit and the pressurizer 

wall temperature ranges between 30-39 oC (cold pressurizer) in the N2 filled region. In table 1, the 

initial conditions of experiments F2.1 Run1 and F2.1 Run2 are shown.

Table 1. F2.1 Run1 and Run2 initial conditions.

Test (RUN) PKL III F OECD-PKL

F2.1 RUN1 F2.1 RUN 2

Secondary side boundary 

conditions

2 SG filled with water, one ready for operation and 

activated to keep pressure at 2 bar

Primary side:

 Upper head by-pass

0.5 0.5

 Level (hot legs) 3⁄4 loop RCL lower edge

 Temperature at core outlet (oC) 61 65

Pressurizer wall temperature

(water-filled region) (oC) 65 46

Pressurizer wall temperature

(N2-filled region)(oC) 136 – 156 30 – 39

 Power (kW) 225 224

Under the conditions shown in table 1, both experiments start when the RHR fails. Due to the 

residual heat generated in the core there is a rise in the core temperature and the void formation in 

the core starts, with an associated increase in the primary pressure. Then, primary coolant comes 

out from the vessel towards the steam generator U-tubes, which act as heat sink. The heat transfer 

into the steam generators causes their secondary-sides to heat up leading to void formation in the 

water-filled secondary sides of steam generators 1 (SG1) and 2 (SG2), and the associated rise in 

the secondary side pressure.

When the pressure of SG1 secondary side reaches 2·105 Pa, the main steam control system is 

activated, and the pressure is maintained constant at this value during the rest of the transient. Also 

in SG1 the secondary side level is controlled and maintained at 12.2 m. In SG2, neither the 
3-2

pressure nor the level controls are activated, so the pressure rises following the evolution of the 

primary system pressure. There is no change in the pressures of SG3 and SG4 as these steam 

generators are empty and no heat is transferred through them, so in both cases the pressure 

remain constant during all the transient (see Figure 2 and 3). After the secondary side pressure 

control activation, most of the residual heat generated in the core is removed through SG1, and the 

plant reaches a stable condition.

Since this moment, the actions taken to mitigate the accident are different for experiment F2.1 

RUN1 and RUN2. In experiment F2.1 RUN1, after the plant stabilization, water is injected into the 

primary system through the accumulators in the following sequence: Injection from accumulator in 

hot leg 1 (ACC HL1), injection from accumulator in cold leg 2 (ACC CL2), injection from 

accumulator in cold leg 3 (ACC CL3), and finally injection from accumulator in hot leg 4 (ACC HL4). 

In Figure 2 all the injections from the accumulators are represented.

After the injections, a depressurization in secondary side of the steam generator full of water but not 

available for operation, SG2, takes place, and the pressure in that zone is lowered to 105 Pa. 

Following the secondary side depressurization there is a primary side depressurization that lowers 

the pressure in the primary circuit to 6·105 Pa. This action allows the injection of the low pressure 

injection system (LPIS) and after that the RHR pumps are restarted.

Figure 2 shows the experimental evolution of the primary and secondary side pressures, with the 

main actions taking place during the transient, as the SG1 control activation, the injections from the 

accumulators, the secondary and primary side depressurization, the injection from the LPIS and the 

recovery of the RHR system.

Figure 2. Experimental evolution of pressures experiment F2.1 RUN 1.

The first part of experiment F2.1 RUN2 is similar to F2.1 RUN1, there is an increase in the core 

0

3

6

9

12

15

18

21

24

27

30

-5000 0 5000 10000 15000 20000 25000 30000 35000 40000 45000

SG 1 main steam pressure control activated

SG 1 secondary side feed activated

ACC hot leg 1

ACC cold leg 2

ACC cold leg 3

ACC hot leg 4

Primary side 

depressurization

LPSI cold leg 4

RHR system loop 4

Time in s

Pressure in bar Loops 1-4

Secondary side 

depressurization

Primary

SGs 3/4 secondary SG 1 secondary 

SG 2 secondary 
3-3

temperature and void formation in the core, then a circulation of primary coolant towards the steam 

generators U-tubes is established and the heat is evacuated through the steam generators and the 

plant reaches at lower pressure than in Run1. In order to achieve the plant conditions to recover the 

RHR system, after the plant stabilization there is an injection from the low pressure injection system

(LPIS) in loop 4, followed by the four injections through the accumulators in the same sequence as 

in the previous case. Once the injections from the accumulators have finished, there is a 

depressurization of the secondary side of SG2 followed by a depressurization of the primary circuit. 

Those actions are needed to allow a second injection from the LPIS in loop 4, followed by an 

injection from HPIS also in loop 4, before recovering the RHR system. Figure 3 shows the 

experimental evolution of the primary and secondary pressures and the different actions taken 

during the transient. 

0

3

6

9

12

15

18

21

24

27

30

-5000 0 5000 10000 15000 20000 25000 30000 35000 40000 45000

SG 1 main steam pressure control activated

SG 1 secondary side feed activated

LPSI cold leg 4

ACC hot leg 1

ACC cold leg 2

ACC cold leg 3

ACC hot leg 4

Primary side 

depressurization

LPSI cold leg 4

HPSI cold leg 4

Secondary side depressurization

Pressure in bar

Time in s

Loops 1-4

Primary

SG 2 secondary

SG 1 secondary

SGs 3/4 secondary

RHS system loop 4

Figure 3. Experimental evolution of pressures experiment F2.1 RUN 2.
3-4
4-1

4. RELAP-5 MODEL OF PKL FACILITY

The transient simulation has been performed using RELAP5-Mod 3.3 code [12]. The RELAP-5 

model used consists of 600 hydraulic volumes, 622 junctions and 512 heat structures. This model 

has been adapted to simulate shutdown conditions, from the PKL model provided by the facility 

[13]. Figure 4 outlines the nodalization used in the simulation.

The core is simulated using a pipe component of eight volumes. Six of these volumes contain the 

fuel rods, which are simulated using a heat structure component that generates the residual heat 

power established in Table 1. The vessel of the PKL facility has two external downcomers (see 

Figure 1) simulated in the RELAP model by means of two external pipes. The cold legs of all four 

loops are simulated using pipe and branch components, which are connected to two branches, 

volumes 232 and 234, which in turn are connected to the downcomer. The facility has four bypasses in the vessel upper head which have been collapsed in this model into two branch 

components, 223 and 225. 

The four primary loops are modelled with a pump and a steam generator in each loop using pipe, 

pump and branch components. The U-tubes of the steam generators are lumped into three pipe

components of different heights. The heat transfer between the primary and secondary systems is 

simulated using three heat structures one for each of the three pipes that simulate the steam 

generators U-tubes.

The different injections from the accumulators performed at the end of the transient (see Figure 2 

and 3) have been simulated using an accum component connected to the loop by a valve. The 

locations of the injections from the accumulators are shown in Figure 4.

As the RHR, LPIS and HPIS inject in the same location of the facility, one injection in each clod leg 

(see figure 4), the model of these injections has been simplified by using time dependent volumes

connected with time dependent junctions which establish the amount of water to be injected. 

The failure of the RHR system is simulated in this model by the on set of the power in the core heat 

structure, which generates the residual heat power included in Table 1. In the same way, the 

restoration of the RHR system at the end of the transient is simulated by disconnecting the power in 

the core heat structure. A sensitivity analysis has been performed using TMDPJUN and TMPDVOL 

to simulate the RHR system operation, but no difference was observed in the results.
4-2

Figure 4. RELAP5 nodalization of PKL facility.

Loop 3

Feedwater

HPIS injection

From 124

To 234 538

500

510

514

550

552

554 562

564

566

568

570

572

522

524

526

528 530

532

Feedwater

Loop 2

To 234

HPIS injection

From 122 438

400

810

805

800 410

414

450

424

428 430

422

468

470

472

452

454

432

466

464

462

426

Loop 1

Feedwater

From 122

To 232

HPIS injection

300

310

314 322

350

352

354 362

364 366

368

370

372

324

326

328

330

332

Loop 4

Feedwater

HPIS injection

From 124

To 232

600

610

614 622

650

652

654

662

664 666

668

670

672

624

626

628 630

632

220

130

124 122

112

102

42 40

22

20

256

246

223

230

234 232

638

338

742

722 712

732

801

815

225

242

To 300

To 400

To 500

To 600

From 438

From 538

From 338

From 638

ACC CL2 ACC CL1

ACC CL3 ACC CL4

ACC HL4
5-1

5. F2.1 SIMULATION RESULTS

This section is divided into two subsections, each one devoted to present the results obtained from 

the simulation using RELAP-5 of experiment F2.1 Run1 and Run2, respectively.

5.1. F2.1 Run1 simulation results 

The first step in the simulation of F2.1 Run1 experiment is the calculation of the transient initial 

conditions shown in Table 1. The amount of water required in the primary circuit has been obtained 

by controlling the water level in the primary loops to 3⁄4 of hot leg. In this phase, the RHR injections 

have been used to extract or supply the amount of water necessary to reach this level. The rest of 

the primary circuit is initialized as full of nitrogen.

The secondary side water levels of the steam generators in loops 1 and 2 are achieved using the 

secondary side feedwater injections (see Figure 4) to reach the initial value. The secondary sides of 

steam generator 3 and 4 are initialized full of nitrogen, as in the transient specification they are 

empty. Table 2 shows the values for the initial conditions obtained in the RELAP-5 calculations as 

compared with experimental data.

Table 2. Experimental and calculated initial conditions for F2.1 RUN1.

Experiment 

specification

RELAP-5 initial condition 

RUN1

Primary: closed and filled with borated water up to 3⁄4-loop, above filled with N2

Coolant inventory 3⁄4 loop 7.71 m in vessel (3⁄4loop)

Boron concentration 2200 ppm 2200 ppm

Pressure 0.93 bar 0.93 bar

Fluid temperature at core outlet 61oC 64oC

Pressurizer wall temperature

(water-filled region) 65oC 65oC

Pressurizer wall temperature

(N2-filled region) 136oC – 156oC 150oC

Loops flow conditions No flow 0.0 kg/s

Pressurizer level 1.0 m. 1.0 m

Secondary: SG1-2 full of water, SG1 ready for operation, SG2 isolated, SG3-4 full of air 

and isolated

Steam generators secondary pressure aprox. 1 bar 1.14 / 1.12 / 1.05 / 0.99

SG1 and SG2 secondary temperature 58oC 58oC

SG1 and SG2 level 12.1 m. 12.1 m
5-2

Once the initial conditions are reached, the transient is initiated by the failure of the RHR system, 

which is simulated by connecting the heat structure in the core that generates the residual heat 

showed in Table 1.

Thus, when the RHR system fails the primary circuit starts to heat up until it reaches saturation 

conditions and void formation in the core begins. The two steam generators with the secondary 

filled with water, SG1 and SG2, act as heat sink. 

Figure 5 shows the comparison between experimental data and RELAP-5 calculations of the water 

level in inlet side of SG1 U-tubes. The calculations performed by RELAP-5 provide a good 

prediction of the time at which water starts entering the U-tubes, although, in some periods of the 

transient the level calculated differs from the experimental data, e.g. around 15000 s.

Figure 5. Water level inlet side of SG1 U-tubes.

Figure 6, shows the water level in the inlet side U-tubes of steam generator 2, which is the steam 

generator with the secondary side full of water but not available, i.e. neither pressure nor level 

control is activated. It can be observed that heat is transferred through this steam generator during 

the first 14000 seconds. At this moment pressure control in SG1 is activated and there is a water 

displacement from steam generator SG2, which empties, towards SG1, where the level increases 

(see Figure 5). At the end of the transient, safety injections produce a new increase in the water 

level in SG2 tubes.
5-3

Figure 6. Water level in the U-tubes of steam generator 2.

The presence of water in the steam generators U-tubes that indicates there is heat transferred from 

the primary to the secondary side of the steam generators. Figure 7 represents the heat, calculated 

by RELAP5, which is transferred through both steam generators, SG1 and SG2, with the secondary 

side full of water. In this figure it can be observed that, when pressure and level control in SG1 is 

activated, the heat exchanged in SG2 decreases and SG1 becomes the only heat sink.

Heat Transfer

0,00E+00

2,00E+05

4,00E+05

6,00E+05

8,00E+05

1,00E+06

1,20E+06

1,40E+06

1,60E+06

0 5000 10000 15000 20000 25000 30000 35000 40000

Time (s)

Heat (W)

SG1 SG2

Figure 7. Calculated heat exchange through the steam generators.
5-4

The control pressure activation in the secondary side of SG1 is observed in Figure 8. This figure 

represents the evolution of the pressure calculated using RELAP5 and the experimental evolution 

for the SG1 secondary side pressure. It can be observed that, RELAP-5 calculation predicts a slight 

delay in reaching 2 105 Pa, which is the set point for pressure control activation, as compared with 

experimental data.

Pressure SG1 Secondary side

0.00E+00

5.00E+04

1.00E+05

1.50E+05

2.00E+05

2.50E+05

0 5000 10000 15000 20000 25000 30000 35000 40000

Time (s)

Pressure (Pa)

RELAP UPV PKL

Figure 8. Secondary side pressure in steam generator 1.

The delay in the RELAP-5 pressure calculations is also observed in the evolution of the secondary 

side pressure of SG2, as shown in Figure 9. In this steam generator, there is no pressure control, 

so an increase in the secondary side pressure until the plant reaches a stable condition about 

1.0·106 Pa is observed. The pressure starts to increase again when injections are activated, and in 

the experiment, the plant reaches another stabilization point around 1.2·106 Pa. However, in the 

calculations performed with RELAP5 the pressure does not reaches a stable value after the 

injections, but presents an increase until the secondary side depressurization, see figure 9.
5-5

Figure 9. Secondary side pressure in steam generator 2.

In the primary circuit, the evolution of the pressure calculated by RELAP5 is quite similar to the 

experimental data, as it can be observed in Figure 10. This figure shows that it is possible to reach 

a stable situation of the plant using one steam generator as final heat sink, with a primary pressure 

around 7·105 Pa, which is quite close to the experimental value. The injections make the pressure 

rise until 1.00·106 Pa and finally the pressure of the system is lowered until 5·105 Pa when the RHR 

is recovered. 

Figure 10. Primary pressure.

 Stabilization Depressurization
5-6

Regarding other important variables in the primary circuit, a difference between the mass inventory 

distribution calculated by RELAP5 and the experiment data is observed. Thus, figure 11 shows the 

reactor vessel level. This figure indicates that the mass inside the reactor vessel is similar for 

experimental and calculated data, although the calculations predict a lower level inside the reactor 

level, what means that more water is displaced from the core towards the reactor coolant circuit. 

Figure 11. Reactor vessel level.

The water comes out from the vessel towards the steam generators U-tubes, see figures 5 and 6, 

and towards the pressurizer. Figure 12 shows the water level inside the pressurizer and great 

differences between the experimental data and RELAP-5 calculation can be observed. In both 

cases, water starts entering the pressurizer, however, after about 2500 seconds the level in the 

pressurizer calculated by RELAP-5 presents a larger increase than the experimental measurement. 

So, more water is displaced towards this volume in the calculation with respect to the experimental 

measure. This situation is maintained for about 17000 seconds after which the level is stabilized, 

but the value reached differs in 7.4 meters between RELAP-5 simulation and experimental data.
5-7

Figure 12. Pressurizer level.

To analyse the boron dilution process during this transient, the boron concentration in the loop seal 

one, that is, in the lowest part of the loop one, has been followed. Figure 13 shows the evolution of 

the boron concentration in this part of the facility. In this figure it can be observed that the 

calculation predicts a delay in the decrease of the boron concentration. 

Figure 13. Boron concentration in loop seal one.

This drop in the boron concentration is due to the mass flow coming from the inlet to the outlet of 

the steam generators U-tubes. The vapour that reaches the top of the U-tubes condenses in the 

outlet side of the tubes. As this condensate is boron free, it causes the drop in the boron 
5-8

concentration when it reaches the loop seal, see Figure 13. 

5.2. F2.1 Run 2 simulation results

As in the previous case, the first step in the F2.1 Run 2 simulation is the calculation of the transient 

initial conditions. In this case the primary circuit is full of water up to the lower edge of the reactor 

coolant line (RCL), and the wall temperature in the pressurizer ranges in the interval 45-65oC. Table 

3 shows the initial conditions for F2.1 Run 2 transient obtained with RELAP-5 calculation as 

compared with experimental data. 

Table 3. F2.1 RUN2 initial conditions.

Experiment 

specification

RELAP-5 initial 

condition RUN 2

Primary closed and filled with borated water up to lower edge of RCL, above filled with N2

Coolant inventory Lower edge of hot legs 7.70 m in vessel.

Boron concentration 2200 ppm 2200 ppm

Pressure 0.93 bar 0.93 bar

Fluid temperature at core outlet 61oC 67oC

Pressurizer wall temperature

(water-filled region) 46oC 45oC

Pressurizer wall temperature

(N2-filled region) 30oC – 39oC 39oC

Pressurizer level 1.0 m. 1.0 m

Secondary: SG1-2 full of water, GV1 ready for operation, SG2 isolated, SG3-4 full of air 

and isolated

Steam generators secondary 

pressure aprox. 1 bar 1.14/1.12/1.05/0.99 bar

SG1 and SG2 secondary 

temperature 48oC 48oC

SG1 and SG2 level 12.1 m. 12.1 m

Once the initial conditions are reached, the transient is initiated by the failure of the RHR system, 

which is simulated by connecting the heat structure in the core that generates the residual heat 

established in Table 1. Thus, when the RHR system fails the primary circuit starts to heat up until it 

reaches saturation conditions and void formation in the core begins. The two steam generators with 

the secondary filled with water, SG1 and SG2, act as heat sink. 
5-9

Figure 14 and figure 15 show the comparison between experimental data and RELAP-5 

calculations of the water level in inlet side of SG1 and SG2 U-tubes, respectively. 

Figure 14. Inlet side of SG1 U-tubes

Figure 15. Inlet side of SG2 U-tubes

One can realize that great differences exist between the results obtained with RELAP-5 and

experimental data. As shown in Figure 14, RELAP-5 predicts that water enters into the inlet side of 

SG1 at 2000 s. after the transient is initiated, while the experimental data indicates there is no flow 

inside the SG1 U-tubes until around 9000 s. The same occurs in SG2, see figure 15, while RELAP-
5-10

5 calculations predict that 2000 s after the RHR is lost there is an amount of water in the SG2 Utubes, while in the experiment there is no water level in this part until 4000 s. In fact, differences 

exist not only in the time when water starts SG1 and SG2 U-tubes, but also in the amount of water 

present in the U-tubes, which would affect heat transfer between primary and secondary sides, and 

consequently primary system pressurization.

It seems there is an inconsistency between the experimental data supplied by the organizers and 

the plant behaviour. For example, observing the evolution of the secondary side pressure of SG1 

and SG2, see figures 16 and 17, it can be seen that the pressure on the secondary side rises 

similarly in both simulation and experimental data, due to the heat exchanged between the steam 

generators primary and the secondary sides, and some water is necessary to produce this 

exchange. Moreover, figure 16 shows the activation of the pressure control in SG1, which 

demonstrates with regard to this physical variable, that the calculations predict quite accurately the 

experimental data. So to achieve the appropriate conditions of pressure in the secondary side there 

must be a heat exchange as predicted by RELAP-5. Note also that activation of the pressure 

control in SG1 starts about 8000 s (Figure 16) while water level in the SG1 U-tubes starts about 

9000 s. (figure 14).

Pressure SG1 secondary side

0.00E+00

5.00E+04

1.00E+05

1.50E+05

2.00E+05

2.50E+05

0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000 24000 26000 28000 30000 32000 34000 36000 38000 40000

Time (s)

Pressure (Pa)

RELAP UPV PKL

Figure 16. Pressure in the SG1 secondary side
5-11

Figure 17: Pressure secondary side SG2

Regarding the heat exchanged, Figure 18 shows the amount of heat exchanged in SG1 and SG2 

calculated by RELAP-5. The heat exchanged explains the pressurization of the steam generators 

secondary sides, as it happens in experiment F2.1 Run1.

Also in this experiment when the pressure control in SG1 is activated, see figure 16, all the residual 

heat generated in the core is removed through SG1, see figure 18. 

Heat Transfer

0.00E+00

5.00E+04

1.00E+05

1.50E+05

2.00E+05

2.50E+05

0 2000 4000 6000 8000 10000 12000 14000 16000 18000 20000 22000 24000 26000 28000 30000 32000

Time (s)

Heat (W)

SG1 SG2

Figure 18. Calculated heat exchange through the steam generators. 
5-12

In the primary circuit, when the RHR is lost there is a rise in the primary pressure until the plant 

reaches a stable situation around 5.00 105 Pa, as shown in figure 19. This situation is maintained 

until the injection from the LPIS at 22350 s. It is also shown that there is no difference between 

RELAP-5 simulation results and experimental data until LPIS injection. The pressure is higher when 

the injection are produced due to the mass distribution in the primary circuit calculated by RELAP-5, 

which predicts more water inside the pressurizer.

Figure 19. Primary pressure evolution

After the LPIS injection, safety injections through the different accumulators start, in the same 

sequence as in the previous case, what raise the pressure in the primary circuit. In order to produce 

the second injection from the LPIS, the pressure of the system must be lowered. The secondary 

side depressurization begins at 31022 s, see figure 17. The primary depressurization starts at 

34750 s, see figure 19, and the injection from the LPIS is produced at 36600 s. Finally, at 38090 s 

an injection from the HPIS is produced.

At 38810 s., once all the injections have ceased it is possible to restart the RHR. Figure 20 shows 

the vessel level during the transient evolution. In this figure it is observed that the experimental level 

reaches more than 10 m but in the calculations performed by RELAP-5 the level in the primary 

circuit never reaches this value. This can be explained by the different mass distribution observed 

between the experimental data and the RELAP-5 calculations. Thus, the calculation predicts more 

water entering into the pressurizer that the experimental data as can be observed in figure 21, 

where the water level inside the pressurizer is represented. Moreover, the amount of water injected 

through the accumulators is lower in the simulation than in the experimental data, due to the 

pressure in the primary circuit calculated by RELAP-5 is higher than in the experiment as can be 

observed in figure 19.
5-13

Figure 20 Primary level 

Figure 21. Pressurizer level

In order to analyze the boron dilution in the primary circuit, the boron concentration on the loop seal 

one has been tracked. Figure 22 shows the evolution of the boron concentration at the loop seal of 

loop 1. In this figure it can be observed that RELAP5 calculations predict a decrease until the time 

the injections start. Since this moment there is an increase in the boron concentration in this part of 

the installation with the maximum reached when the RHR is recovered. The sharp decrease 
5-14

observed in the experimental data, after 15200 s is not reproduced by the calculations, which 

shows instead a smooth decrease.

Figure 22. Boron concentration at the loop seal 1

5.3. Comparison of F2.1 Run 1 and Run 2 plant evolution simulation.

One of the objectives of this work is focused on analysing the influence of the plant initial conditions 

on the plant behaviour after the RHR is lost. In this case, the two runs simulated differ in the amount 

of water in the reactor coolant circuit and in the temperature of the pressurizer. The effect of the 

variation in the initial plant situation can only be analyzed during the first part of the transient, until 

the plant reaches a stable situation, as after this time, different actions are taken in each case to 

mitigate the transient consequences. 

In this section, the evolution along the first part of the transient of some important variables, which 

show the difference in the plant behaviour, are presented. The figures show the evolutions of the 

calculations performed using RELAP5 for F2.1 Run1 and F2.1 Run2 respectively. 

Thus, Figure 23 shows the evolution of the primary pressure for F2.1 Run1 and F2.1 Run 2. As it 

can be observed in this figure for both cases the plant reaches a stable condition, but the value of 

the primary pressure at which the plant reaches a stable condition is different depending on the 

initial conditions considered. Thus, with the primary full of water up to 3⁄4 of loop and hot pressurizer 

the stabilization pressure is higher than supposing a lower level of water in the primary circuit with a 

cold pressurizer. 
5-15

Figure 23. F2.1Run1 and Run2 primary pressure calculation.

Regarding the mass distribution inside the primary circuit, in both cases RELAP5 predicts a 

different behaviour than in their experiments as explained in the previous sections. In fact, the same 

behaviour is observed in both Runs, for which RELAP5 calculations predict a higher level of water 

in this part of the system, see figures 12 and 21. Figure 24, shows the water level inside the 

pressurizer that reaches 10m for Run1 and Run2, which corresponds to the PZR full of water. The 

only difference observed in the evolution of the water level inside the pressurizer is that for 

F2.1Run2, primary level up to the RCL edge and cold pressurizer, the water starts entering the 

pressurizer earlier than in F2.1Run1.
5-16

Figure 24. F2.1Run1 and Run2 pressurizer level calculation

The evolution of the water level in the pressurizer agrees with the behaviour of the water inside the 

reactor vessel in each case. Thus, figure 25 shows the evolution of the level of water inside the 

vessel for F2.1Run1 and F2.1Run2 calculations. In this figure it can be seen a slight advancement 

in the decrease of the level inside the vessel and the final level reached is lower than in the 

calculations of F2.1Run1. So, the water coming out of the vessel comes into the pressurizer earlier 

in Run2 than in Run1.

Figure 25. F2.1Run1 and Run2 vessel level calculation

Finally, figure 26 presents the results obtained using RELAP-5 for the boron concentration evolution 
5-17

in the loop seal 1. For this variable, the results obtained in the calculations performed are quite 

similar in both situations. As shown in figure 26, the boron concentration is maintained for both 

cases until 10000 s., and after this time the boron concentration presents a smooth decrease in 

both cases. 

Figure 26. F2.1Run1 and Run2 boron concentration calculation
5-18
6-1

6. RUN STATISTICS

The calculations of F2.1 Run 1 and 2 have been performed using a server of Cluster IBM 1350 

with a biprocessor Intel Xeon with the following characteristics:

• x335 2.40GHz/100MHz/512KB L2, 512MB Memory, 331W, HS Open bay.

• x335 Processor 2.4GHz/512KB Upgrade. 

• 1GB PC2100 CL2.5 ECC DDR SDRAM RDIMM.

• 18.2GB 10K-RPM ULTRA 160 SCSI Hot-Swap SL HDD 

• Remote Supervisor Adaptor

In tables 4 and 5 there are exposed the relevant parameters of the run statistics of the simulation of 

experiment F2.1 RUN1 and RUN2, respectively.

Table 4: Run Statistics F2.1 RUN 1

RT CPU TS CPU/RT C DT GT

Steady-state 2800 10186.00 0.005 3.6379 600 101682 0.1670

Pressure control 9700 135953.61 0.01 14.0158 600 1412054 0.1605

Primary stabilization 18500 279384.75 0.005 15.1021 600 3234595 0.1440

First injection 27120 418987.17 0.005 15.4494 600 4954747 0.1409

End of transient 40940 647007.42 0.005 15.8038 600 7835986 0.1376

Table 5: Run Statistics F2.1 RUN 2

RT CPU TS CPU/RT C DT GT

Steady-state 2800 8490.7998 0.005 3.0324 600 100543 0.1407

Pressure control 9400 117548.69 0.01 12.505 600 1317185 0.1487

Primary stabilization 14000 175739.52 0.01 12.553 600 1946681 0.1505

First injection 22350 356435.66 0.005 15.948 600 3657256 0.1624

End of transient 41650 621093.51 0.005 14.912 600 6667797 0.1552

RT: Transient time (s)

CPU: Execution time (s)

TS: Maximum time step (s)

C: Total number of volumes

DT: Total number of time steps

GT: GT = (CPU*103

/(C*DT))
6-2
7-1

7. CONCLUSIONS

This work is focused on the simulation of a RHR failure with the plant in shutdown conditions and 

the primary circuit closed. In this conditions two experiments have been performed to asses the 

effect of the amount of water inside the primary circuit and the pressurizer initial conditions on the 

plant behaviour.

From the results presented in the previous sections it can be observed that in any case the plant 

reaches a stable situation when the residual heat sink is evacuated through the steam generators. 

The new situation reached maintains the plant integrity and the injections produced at the end of 

the transient assure the plant safety during the entire transient.

The results obtained in F2.1Run1 simulation agree with the experimental data although a significant 

difference in the mass distribution in the primary circuit is found, as can be observed in the 

pressurizer, U-tubes and vessel levels. Regarding the boron dilution in the loop seal, the simulation 

predicts a delay in the evolution of the boron concentration in this part of the facility. A similar result 

is obtained from the F2.1Run2 simulation when comparing the levels inside the primary, what 

implies that again in this calculation the mass distribution inside the primary circuit is different from 

the experimental data. And, also in F2.1Run2 the boron concentration in the loop seal differs from 

the experimental data as the simulation is delayed and there is a sharp deboration in the 

experimental data that is not predicted by the calculation. 

The comparison of the result obtained with different plant initial conditions, shows that the primary 

level and the temperature conditions in the pressurizer affect the plant behaviour. In any case a 

stable plant situation is reached, and it is always maintained under safe conditions, but there is a 

difference in the stabilization situation depending on the initial situation supposed. However, the 

influence of each initial condition separately on the plant behaviour cannot be analysed as the 

experimental data is performed varying both parameters simultaneously. 
7-2
8-1

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