NUREG/IA-0423 

International 

 Agreement Report

Analysis with TRACE Code of PKL-III Test F 1.2 

 

 

 

Prepared by:

J.L. Munoz-Cobo, S. Chiva, A. Escrivá

 

 Instituto de Ingenierίa Energética

 Univesitat Politécnica de Valéncia

Camino de Vera s/n. 46022 Valéncia

 

 

A. Calvo, NRC Project Manager

 

 Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: January 2013

Date Published: February 2013

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-0423 

International 

 Agreement Report

Analysis with TRACE Code of PKL-III Test F 1.2 

 

 

 

Prepared by:

J.L. Munoz-Cobo, S. Chiva, A. Escrivá

 

 Instituto de Ingenierίa Energética

 Univesitat Politécnica de Valéncia

Camino de Vera s/n. 46022 Valéncia

 

 

A. Calvo, NRC Project Manager

 

 Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: January 2013

Date Published: February 2013

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

The goal of this report is to explain the main results obtained in the simulations performed with 

the consolidated code TRACE for the OCDE TEST PKL-III F 1.2. The transient was produced by 

a set of successive extractions of refrigerant performed through the bottom of the vessel of the 

PKL-III facility in Germany, after each extraction phase no extraction was performed during a 

certain time, this no-extraction time was long enough to reach a quasi-stationary state so the 

new natural circulation conditions could be experimentally observed. Therefore, the goal of this 

test is to see how the inventory of refrigerant in the primary circuit affects to the natural 

circulation conditions in the primary circuit and when the natural circulation stops. The 

experiment had two different parts during, the first one were performed a set of extraction of 

refrigerant followed by no extraction intervals, until a minimum inventory was reached, during this 

set of extractions and the quasi-stationary periods that followed to each extraction period, the 

power was maintained constant. The second step of the experiment consisted in a set of 

injections symmetrical to the extractions, which refilled again the primary system although the

total number of refilling steps was smaller than the number of the extractions and, therefore, the 

initial conditions were not achieved. The total duration of the experiment was 90000 s. After each 

extraction step the relief valves of the secondary system were manipulated in order to maintain

constant the pressure in the primary system, this was done manually by the operator. In the code 

simulation, we simulate with TRACE the first 40000 s of the experiment until the natural 

circulation was completely stopped and the system enters in reflux condenser conditions.

The TRACE code was able to predict correctly the natural circulation mass flow rate along time. 

We note that the natural circulation mass flow rate at the beginning of the transient is 1.21 kg/s 

that is correctly predicted by TRACE, then after several successive extractions steps this mass 

flow rate increases reaching a maximum value of 3 kg/s. This value is also correctly predicted by 

the TRACE code. When the inventory is too low the natural circulation start to diminish 

progressively with the successive extractions until it stop.

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, TRAC-B, and currently TRACE. These complex tools, suitable state-ofthe-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

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/FAP

, within different periods of the associated national programs (e.g., CAMPEspaña). 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.

2

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

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. 

 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

in this international support of facilities and in the establishment of a large network of international

collaborations. In 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

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

Page

FOREWORD ............................................................................................................................... v

FIGURES ...................................................................................................................................viii

TABLES......................................................................................................................................viii

EXECUTIVE SUMMARY ............................................................................................................... ix

ACKNOWLEDGMENTS .............................................................................................................. xi

ABBREVIATIONS ..........................................................................................................................xiii

1. INTRODUCTION ................................................................................................................... 1-1

2. INITIAL AND BOUNDARY CONDITIONS ............................................................................ 2-1

2.1 Description of the PKL facility .................................................................................... 2-1

2.2 Fundamentals and goals of the injection experiments with borated water..............2-2

2.2.1 General information on the heat transfer and circulation regimens during 

a SBLOCA.................................................................................................2-2

2.2.2 Specific goals of the F1.2 experiment. ......................................................2-3

2.3 Initial and boundary conditions for the experiment PKL III F 1.2, comparative analysis 

with TRACE model initial conditions........................................................................2-4

2.3.1 Initial conditions.........................................................................................2-4

2.3.2 Sequence of events and procedures during the PKL III F 1.2 experiment 2-8

3. POST TEST RESULTS WITH THE TRACE CODE OF THE EXPERIMENT PKL III F 1.2 .3-1

3.1 Conditioning phase with the TRACE code ..............................................................3-1

3.2 Results with the TRACE code for the experiment F 1.2..........................................3-1

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

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

FIGURES

2.1 Primary system of the PKL-III facility.....................................................................................2-1

Page

2.2 Configuration of the safety system of PKL facility.................................................................2-3

2.3 TRACE model developed by the Polytechnic University of Catalunya and that was used 

as starting point to model the PKL-III facility.........................................................................2-6

2.4 Coolant distribution at the beginning of the F 1.2 experiment. The blue colour shows the 

liquid distribution while the red colour shows the steam distribution. ...................................2-7

3.1 Pressure evolution at the core exit. The experimental results are displayed in blue, and 

the TRACE results in red.......................................................................................................3-2

3.2 Evolution with time of the steam temperature in the hot leg of the second loop. The blue 

line denotes the experimental data while the red line denotes the TRACE simulation........3-3

3.3 Evolution with time of the liquid temperature in the hot leg of loop 2. The blue line 

displays the experimental results. .........................................................................................3-3

3.4 Pressure versus time at the exit of the secondary of the steam generator in the four loop. 3-4

3.5 Mass flow rate in kg/s versus time in the recirculation pump seal (cell 1, component 249) 

of the first loop. ......................................................................................................................3-5

3.6 Mass flow rate in kg/s versus time in the recirculation pump seal (cell 1, component 449) 

of the third loop. .....................................................................................................................3-5

3.7 Boron concentration in mass fraction versus time in the pump seal of the loop number 3..3-6

3.8 Void fraction versus time in cell 11 (Upper level) of the steam generator 1.........................3-7

3.9 Void fraction versus time in cell 5 (intermediate level) of steam generator 1.......................3-7

3.10 Void fraction versus time in the lower part of the SG primary in loop 1................................3-8

3.11 Void fraction versus time in the hot leg of loop 1 at component 210....................................3-8

3.12 Void fraction versus time in the cold leg of loop 1 at component 247. .................................3-9

3.13 Void fraction versus time in the RCP seal, in the cold leg of loop 1. ....................................3-9

3.14 Collapsed level of water in the steam generator of loop 1, calculated by the control block 

1526. ....................................................................................................................................3-10

TABLES

2.1 Primary initial conditions at t=0 for the experiment PKL III F 1.2. The primary system is 

full of sub-cooled water flowing in single phase natural circulation regime..........................2-4

Page

2.2 Initial conditions in the secondary of PKL III facility at the beginning of test F 1.2...............2-4

2.3 Comparison of the initial conditions of the PKL III facility for test F 1.2, and the data 

computed with the TRACE model at the beginning of the transient.....................................2-7

2.4 Sequence of events in the F 1.2 experiment ........................................................................2-9
ix

EXECUTIVE SUMMARY

This report presents the main results obtained with the NRC consolidated code TRACE for the 

simulation of the OCDE PKL-III test F 1.2 with a model of the PKL facility.

The transient was produced by a set of successive extractions of refrigerant performed through 

the bottom of the vessel of the PKL-III facility in Germany, after each extraction phase no 

extraction was performed during a certain time, this no-extraction time was long enough to reach a 

quasi-stationary state so the new natural circulation conditions could be experimentally observed. 

Therefore, the goal of this test is to see how the inventory of refrigerant in the primary circuit 

affects to the natural circulation conditions in the primary circuit and when the natural circulation 

stops. The experiment had two different parts, during the first one were performed a set of 

extraction of refrigerant followed by no extraction intervals, until a minimum inventory was 

reached, during this set of extractions and the quasi-stationary periods, that followed to each 

extraction period, the power was maintained constant. The second step of the experiment 

consisted in a set of injections symmetrical to the extractions, which refilled again the primary 

system, although the total number of refilling steps was smaller than the number of the extractions 

and, therefore, the initial conditions were not achieved. The total duration of the experiment was 

90000 s. After each extraction step the relief valves of the secondary system were manipulated in 

order to maintain constant the pressure in the primary system, this was done manually by the 

operator. In the code simulation, we simulate with TRACE the first 40000 s of the experiment until 

the natural circulation was completely stopped and the system enters in reflux condenser 

conditions.

The TRACE code was able to predict correctly the natural circulation mass flow rate along time. 

We note that the natural circulation mass flow rate at the beginning of the transient is 1.21 kg/s 

that is correctly predicted by TRACE. Then, after several successive extractions steps, this mass 

flow rate increases reaching a maximum value of 3 kg/s. this value is also correctly predicted by 

the TRACE code. When the inventory is too low, the natural circulation start to diminish 

progressively with the successive extractions until it stop.

xi

ACKNOWLEGMENTS

The thermal-hydraulic and nuclear engineering group of the UPV is indebted to the management 

board of the PKL-III project and to the people of AREVA NP. The CSN financed this work, and 

help to develop this report. Also we are indebted to the UPC and in particular to Francesc 

Raventos and Carmen Pretel that provides a model of the PKL-III facility for the TRACE code.

xiii

ABBREVIATIONS

CSN Spanish Nuclear Regulatory Commission

CRGT Control Rod Guide Tubes

ECCS Emergency Core Cooling System

CVCS Chemical and Volumetric Control System

HPIS High Pressure Injection System

LOCA Loss of coolant accident

LPIS Low Pressure Injection System

OCDE Organization for the Cooperation and Economic Development

PKL Pimärkreislauf facility in Erlangen

PRZ Pressurizer

PTS Pressurized Thermal Shock

PWR Pressurized Power Reactor

RCP Recirculation Pump

RHRS Residual Heat Removal System

SBLOCA Small Break Loss of coolant accident

SG Steam Generator

SRV Safety and Relief Valve

UNESA Spanish Electricity Producers Association

UPC Universidad Politécnica de Catalunya

TRACE TRAC/RELAP Advanced Computational Engine

1-1

1. INTRODUCTION

The present work is a contribution to the OCDE international collaborative research project PKLIII. The Spanish Nuclear Regulatory Commission coordinated the participation in this project, with 

the contribution of the Spanish Electricity Producers Association (UNESA). A consortium formed 

by the CSN, several Spanish Technical Universities and UNESA developed the Spanish 

participation in the project that was coordinated by the CSN and a steering committee.

The analysis of the experiment PKL-III F 1.2 with the TRACE code was assigned to the “thermalhydraulics and nuclear engineering group” of the Polytechnic University of Valencia. This test was 

performed in the PKL (Primärkreislauf) facility located in the research centre of AREVA 

FRAMATOME in Erlangen (Germany) which is operated by FRAMATOME ANP [1, 2].

The Test PKL III F 1.2 consisted in two stages, that we will call extraction and injection stages,

respectively. The first one formed by a set of successive phases that consisted in a set of 

extractions of refrigerant fluid, performed through the lower part of the vessel, followed by a period 

without extraction after each individual extraction. The period without refrigerant draining after 

each extraction phase was long enough to attain a pseudo-stationary state. During this set of 

steps, the power was maintained constant through the entire transient and, at the same time, the 

operator of the facility changed manually the area of the secondary relief valve in order to maintain 

constant the pressure in the primary system along the transient.

The total duration of the two stages was 90000 s. In this report, we will analyse the first 40000 s of 

the transient that, correspond to the extraction stage. We will compare the experimental data with 

TRACE results for a set of characteristics thermal-hydraulic variables.

Because no experimental data of the area of the relief valve or stem position through the transient

were recorded, the variation of this area was computed by trial and error, i.e. at each extraction 

step we selected the new area value that was able to maintain the pressure constant in the 

primary system during each quasi-stationary period. Because after each extraction step the fluid 

inventory in the system was smaller, then to maintain the pressure constant into the system we 

need to remove less heat from the primary. This was accomplished by raising the temperature in 

the secondary, to do this the facility operator adjust the valve stem position in order to increase the 

secondary pressure and in this way to reduce the primary secondary heat transfer. 

One subject of special interest in this test was to study the influence of the primary refrigerant 

inventory on the natural circulation mass flow rate, when the refrigerant inventory is too small the 

system enters in reflux condenser conditions and the natural circulation stops.

The goal of test series 1 was to obtain the multidimensional temperature distribution in the cold 

legs and the vessel downcomer during the ECCS injection for verification of computer codes and 

models. In test 1.2 the break was located in the bottom of the hot leg B, in such a way that the 

break does not disturb the stratification in the cold leg. 

The main goal of this work is to compare the experimental results with the predictions of the 

TRACE code and to analyze the set of phenomena that take place in the primary and secondary 

loops.

The work has been organized as follows; section 2 is devoted to study the facility model and the 

boundary conditions. In section 3 we study the transient sequence and we compare the TRACE 

results for the main physical magnitudes with the experimental data. Finally, section 4 contains the 

main conclusions of this study.

2-1

2. INITIAL AND BOUNDARY CONDITIONS

2.1 Description of the PKL facility

The experiment PKL III F 1.2 was performed in the PKL facility that reproduces, as displayed in 

figure 2.1, a four loops 1300 MWe PWR scaled with a factor of 1:145 in volume, and power. The 

PKL facility is scaled 1:1 in height. The facility contains all the components of the primary system 

denoted as (1) pressure vessel, (2) downcomer, (3) steam generators, (4) primary pumps, (5) 

pressurizer. As displayed in figure 2.1 the facility has four primary loops.

Figure 2.1 Primary system of the PKL-III facility.

The special characteristics of this facility are:

I) Four loops symmetrically arranged around the pressure vessel.

II) Identical lengths of the pipes connecting the same components in the four loops, and 

identical friction losses

III) The connexions of the hot and cold legs to the reactor vessel are located at the same 

height.

IV) The upper part of the reactor vessel downcomer modelled as a ring.

V) The modelling of a power plant with three loops is performed isolating one of the four 

loops.
2-2

The mass inventory in the primary circuit of the PKL-III facility at 100% of its capacity is 2500 kg,

this amount of water is enough to fill the pressurizer up to a height of 1.3 meters with an average 

density in the primary circuit of 890 kg/m3

. The heat losses of the primary circuit at 250 oC of 

temperature are 86 kW. 

The actual design of the PKL III facility allows the simulation of accidental scenarios under 

symmetrical and non-symmetrical conditions, and the research on individual effects of multiple 

failures. The laboratory performed several modifications in the original PKL facility to adapt it to a 

Westinghouse PWR design.

The PKL facility is equipped, in the primary and secondary circuits, with all the important 

operational and safety systems of the original plant design displayed at figure 2.2. These safety 

systems are:

In the primary circuit:

- Four independent safety injection systems at high and low pressure, which are connected 

to the hot and cold legs:

- Removal system of residual heat 

- 8 Accumulators

- Pressure control system (pressurizer)

- Volume control system

Secondary circuit:

- Main steam lines with all the characteristics of the original system (the condenser and the 

turbine are not simulated)

- Feed water system

- Emergency feed water system

- Depressurization system

2.2 Fundamentals and goals of the injection experiments with borated water

2.2.1 General information on the heat transfer and circulation regimens during a SBLOCA

During a SBLOCA transient the mass inventory in the primary system depends on the break size 

and the number of systems available for emergency cooling. Therefore, we will have different heat 

transfer mechanisms along time in the core, and in the steam generators, depending on the mass 

inventory and the enthalpy of the cooling water. If the primary system is practically full with subcooled cooling water, then the heat transfer regime to the secondary will be single-phase natural 

circulation. Then, it stars the production of steam in the reactor core channels when we reach 

saturation conditions in the primary. After some time, a two-phase natural circulation regime is 

established. If the primary system continues loosing cooling water through the break, then after a 

certain time, it is produced the phase separation in the upper part of the inverted U tubes ( ∩ ) . 

This fact causes the stop of the natural circulation and the system enters into the reflux condenser 

mode. This means that the steam produced in the core flows toward the steam generators, where 

it condenses. Then, the condensate flows toward the pressure vessel though the hot or cold legs,

depending on the position where the steam has condensed inside the tubes. During this 

operational mode, the removal of the core decay heat to the secondary is assured unless while the 

level of the two-phase mixture covers the upper reactor core level. We must also consider the fact 

that the condensate from the steam generators is free of dissolved boron, and bags of low-borated 

water can accumulate in the seal pump of the cold leg.
2-3

If the pressure of the primary system continues decreasing with time, then because the amount of 

coolant discharged through the break is smaller than the amount of coolant injected through the 

emergency cooling systems (ECCS), then it is possible to re-establish again natural circulation 

conditions. While it continues the refilling of the primary system and before the re-establishment of 

the natural circulation, the condensate, with de-borated water contained in the pump seal, will be 

transported first in the opposite direction of the ordinary flow, i.e. toward the steam generator exit. 

Nevertheless, when the natural circulation is re-established, these slugs of condensate are 

transported in the ordinary direction of the flow, i.e. toward the reactor pressure vessel. The 

amount of low-borated water that enters into the reactor core will depend on the amount of 

condensate that is formed and stored during the reflux condenser phase and how this low borated 

slugs mixes with the borated water in the downcomer and the lower plenum. 

Figure 2.2 Configuration of the safety system of PKL facility.

2.2.2 Specific goals of the F1.2 experiment.

The experiments of the PKL III series have shown that one of the conditions to have an

accumulation of non borated water in the pump seal is to have a small inventory of coolant in the 

primary system. In other words, it has been observed a continuous accumulation of non borated 

water in the pump seals only when the water level at the exit side of the steam generator drops 

below the tube sheet. For higher levels of water inside the steam generator tubes, it was not 

possible the accumulation of condensed non-borated water in the pump seal. In this last situation,

it exits a discontinuous transport of borated water from the inlet to the outlet of some of the U 

tubes of the steam generator, and obviously, this has an influence on the boron concentration in 

the pump seal. The water inventory is, in this last case, too big to have reflux condenser conditions

in all the tubes. This effect has been observed not only during the phase of inventory reduction in 

the primary system but also during the beginning of the refilling phase, when most of the U tubes 

of the steam generator were empty.

The consequence of these observed phenomena is that the size of any condensate slug is limited 

by the volumes of the pump seal and the pipe connecting the exit of the SG with the pump seal. 
2-4

Therefore, we must exclude any additional accumulation of non borated condensate inside the SG 

tubes. The condensing process and the transport of coolant inside the SG tubes dominate the 

phenomena explained in this section, and also we must consider the heterogeneous behaviour of 

the different tubes of the SG. This means that not all the SG tubes behaves in the same manner,

i.e. in some of the tubes we can have reflux condenser conditions while in others we can have 

transport of two-phase mixture. 

One of the main goals of the experiment F 1.2 is to improve our understanding of the physical 

processes that we have mentioned above, and to analyze the parameters that have a relevant 

influence on the system behaviour. The F 1.2 experiment have been planned in order to perform a 

parametric study about the importance of the water inventory inside the steam generators on the 

accumulation of non borated water in the pump seal. This research was conducted as a validation 

tool for thermal-hydraulic codes.

Therefore, the main goals are to know the maximum size of the non borated slugs that are formed 

in the pump seals and the minimum boron concentration inside these slugs when the water levels 

are maintained above the sheet of the SG tubes.

2.3 Initial and boundary conditions for the experiment PKL III F 1.2, comparative analysis

with TRACE model initial conditions

2.3.1 Initial conditions 

The experiment PKL III F 1.2 did not have a conditioning phase as in earlier experiments. For the 

test phase, the experiment starts from the initial conditions displayed at tables 2.1 and 2.2. The 

primary circuit was filled with sub-cooled coolant at the beginning of the test stage, which flows

through the loops and core in single-phase natural circulation regime. The boron concentration 

was homogeneous in all the primary circuit with a value of 2000 ppm. The pressure in the primary 

system was 12.0 bar, the fluid temperature at the core exit was 184 oC and the primary mass 

inventory was 2500 kg. The secondary pressure in the SG at the beginning of the test stage was 

5.9 bar with a collapsed level of water in the SG of 12 m.

Table 2.1 Primary initial conditions at t=0 for the experiment PKL III F 1.2. The primary 

system is full of sub-cooled water flowing in single phase natural circulation 

regime

Coolant inventory 2500 kg

Boron concentration 2000 ppm in all the primary loops

Heating power of the rods 600 kW (1.8%)

Pressure 12.0 bares

Fluid temperature at the core exit 184 oC

Subcooling temperature at the core exit 4 K

Coolant temperature in the pressurizer 189 oC

Water level in the pressurizer 1.3 m

Flow conditions Single phase natural circulation in the four loops

Table 2.2 Initial conditions in the secondary of PKL III facility at the beginning of test F 1.2 

Steam pressure in the 4 steam generators 5.9 bar

Steam temperature in the four steam generators 158 oC

Water collapsed level in the four steam generators 12 m

Feed water temperature 115-120 oC
2-5

The TRACE input-deck model for the PKL facility was provided to the authors of this study by the 

Polytechnic University of Catalunya (UPC), the nodalization of this model is displayed in figure 2.3.

To obtain the initial conditions for the experiment PKL III F 1.2, we performed a conditioning phase 

with the TRACE code. Also we performed a set of modifications in the input-deck model in order to 

comply with the boundary conditions of the F 1.2 experiment. The initial conditions of the model 

provided by UPC were a pressure in the primary system of 41.5 bars, with single phase natural 

circulation and a liquid temperature of 500 K. The secondary pressure in the UPC model was 28.5 

bars, and in this way the secondary acts as a heat sink. The power in the primary was set to 600 

kW and the mass inventory in the primary was 2500 kg. To achieve the initial conditions required 

by the experiment F 1.2, we performed a depressurization in the secondary at a constant rate until 

we achieve the desired initial conditions of the experiment, this phase lasts 9600 s. At the end of 

this conditioning phase, we have practically the desired initial conditions as displayed in table 2.3.

The figure 2.4 displays the initial conditions at the beginning of the experiment, and the degree of 

sub-cooling in the different parts of the primary circuit.

We observe that at the beginning of the experiment the pressurizer is practically empty of coolant 

fluid. The degree of sub-cooling at the beginning of the transient in the cold legs, in the 

downcomer and in the lower plenum is bigger than 20 K. The steam generators are full of coolant 

water up to a height of 12 meters that cover completely the U tubes.
2-6

Figure 2.3 TRACE model developed by the Polytechnic University of Catalunya and that 

was used as starting point to model the PKL-III facility.
2-7

Table 2.3 Comparison of the initial conditions of the PKL III facility for test F 1.2, and the 

data computed with the TRACE model at the beginning of the transient

Physical Magnitudes Initial Conditions for Test

PKL III F 1.2 Initial conditions in TRACE

Coolant inventory 2540 kg 2500 kg

Power of the heating rods 600 kW 599.1 kW

Pressure in the core 12 bar 12.16 bar

Fluid temperature at the core 

exit 184oC ( 457.15 K ) 182oC (455 K)Component 120 cell 7

Natural circulation mass flow 

rate 1.2 kg/s per loop 1.16 kg/s per loop

Secondary pressure 6 bar

6.2 bar

Component 610-cell 9

Secondary-SG.

Secondary Temperature 158oC ( 431.15 K )

157oC (430.15 K)

Component 610-cell 9

Secondary-SG

Pressure at the exit of the 

steam generator in the main 

steam line

5.9 bar 5.83 bar

Component 612-cell 2

Figure 2.4 Coolant distribution at the beginning of the F 1.2 experiment. The blue colour 

shows the liquid distribution while the red colour shows the steam distribution.
2-8

2.3.2 Sequence of events and procedures during the PKL III F 1.2 experiment

The experiment F 1.2 started with single-phase natural circulation conditions in the four loops. 

During the transient, the secondary pressure was controlled by opening and closing manually the 

relief valves, with the goal to maintain the pressure in the primary circuit at the desired level. At the 

1300 s of the beginning of the transient the main steam valve was partially closed, which produced 

an increase of the pressure in the secondary. This increase in the secondary pressure reduced the 

primary to secondary heat transference and the primary pressure started to increase. At time 3800 

s from the beginning of the transient we attained saturation conditions at the core exit and we 

started to accumulate steam in the vessel head. After the first accumulation of steam in the upper 

head of the vessel, the operators closed at time 4650 s the pressurizer isolation valve located in 

the purge line. Then a set of coolant draining stages began, followed by quasi-stationary states 

(stages without draining). Table 2.4 displays the sequence of events during the draining phase of 

this experiment.

Other boundary conditions considered during the F 1.2 test are:

- The LPIS system was not available

- The HPIS system was not available

- The accumulators were not available

- The RHRS system was not available

- The recirculation pumps (RCP) were stopped

- The butterfly valves of the RCP, were closed to simulate the hydraulic resistance of the 

RCP pumps.

- The system CVCS was modified in order to use it to increase the coolant inventory. 

Injecting cold water with a 2000 ppm concentration of boron water into the lower plenum.
2-9

Table 2.4 Sequence of events in the F 1.2 experiment

Time (seconds) Events Inventory of coolant in 

the primary system (kg)

0 Starting of the test 2500

1300 Partial closing of the MSV and the secondary 

pressure starts to increase 2500

3800 The saturation temperature at the core exit is 

attained 2495

4650 Isolation of the pressurizer 2415

5100 Steady state 2415

5850 Start of draining 2415

6100 End of draining 2365

6890 Start of draining 2365

7350 End of draining 2240

9660 Start of draining 2240

10150 End of draining 2110

13150 Start of draining 2110

13850 End of draining 1990

19500 Start of training 1990

20300 End of draining 1860

22300 Start of draining 1860

23150 End of draining 1730

25300 Start of draining 1730

26150 End of draining 1605

28360 Start of draining 1605

29200 End of draining 1470

32620 Start of draining 1470

33450 End of draining 1310

37250 Start of draining 1310

38100 End of draining 1120

40410 Start of draining 1120

41200 End of draining 935

44150 End of the data file of the day 8, and beginning of 

the data file of day 9 935

45690 Start of draining 935

46850 End of draining 810

49880 Start of draining 810

51100 End of draining 685

53050 Start of the coolant injection test 685
2-10
3-1

3. POST TEST RESULTS WITH THE TRACE CODE OF THE EXPERIMENT PKL III F 

1.2

3.1 Conditioning phase with the TRACE code

As we mentioned earlier in this experiment there is not conditioning phase, table 2.3 displays the 

results obtained with the TRACE code when we simulate the beginning of the experimental phase. 

In our case, as we mentioned earlier, we departure from a PKL model for TRACE provided by 

UPC, with initial and boundary conditions different from the conditions of the F 1.2 experiment. To 

attain the experimental conditions of the F 1.2 experiment, we modified these boundary conditions 

according to the test data provided by the experimentalists and then we performed a 

preconditioning phase that consisted in a cooling by reducing the secondary pressure at a 

constant rate. At the end of this preconditioning phase we attained the conditions displayed in 

table 2.3, that are very close to the initial conditions for the experiment F 1.2. It is remarkable the 

good agreement between TRACE results and the experimental data for the natural circulation 

mass flow rate at the beginning of the experimental phase. In this case the prediction of TRACE

code was 1.16 kg/(s loop) while the experimental data was 1.2 kg/(s loop).

The small difference in the power displayed at table 2.3 was produced by the fact that in the UPC 

model there was a small source of 1 kW of heat in the vessel head. This negligible source heated 

the steam accumulated in the vessel head much above the saturation temperature and gives 

problems during the test execution. To solve this problem we reduce this heat source to a value of 

0.01 kW. 

3.2 Results with the TRACE code for the experiment F 1.2

We have performed the analysis with the TRACE code for the phase of successive coolant 

extractions through the lower part of the vessel until the natural circulation in the primary circuit is 

completely stopped performed during the experiment F 1.2. We have simulated a total time of 

transient of 37400 s, plus the 9600 s of the pre-conditioning phase. When the natural circulation 

stops, the system enters in reflux condenser conditions and the TRACE code uses a very small 

integration time step of 10-4 s, and because it is necessary to perform a manual adjustment of the 

relief valve stem position (valve area) in order to maintain the primary pressure at 12 bar, then the 

time necessary to perform this adjustment with this small time step becomes very long. For this 

reason we decided to simulate only the first phase of the transient until the natural circulation in 

the primary was stopped completely.

We will give the total time that include the 9600 s of preconditioning phase in TRACE code, and 

between brackets the experimental time. The transient starts with natural circulation conditions 

with a mass flow rate of 1.2 kg/(s loop). Then a set of extractions that lasts from 750 to 850 s are 

performed through the lower part of the reactor vessel. Following each extraction, we do not purge 

the vessel for a variable period that lasts from 2500 to 4500 s, during this time the system attains

quasi-stationary conditions. The coolant draining performed during the first five extractions 

promotes an increase in the natural circulation mass flow rate of the primary, that rises from 1.2 to 

3 kg/s per loop, this change is very well predicted by the TRACE code, although the rise starts 

earlier in time. This increment in the mass flow rate per loop is produced because we pass from

single-phase natural circulation regime to two-phase natural circulation due to the reduction of the

coolant inventory in the primary, and to the transference of heat from the rods to the coolant in the 

reactor core. This causes an increase of the natural circulation driving force produced by the 

difference of densities between the cold and hot legs. If the coolant extraction through the core 

lower part continues, then after the sixth extraction, at t=32600 s (23000 s), starts a progressive 

reduction of the mass flow rate circulating through the primary. This decrease in the circulating 

mass flow rate is produced by the fact that the successive inventory reductions increase the void 
3-2

fraction not only in the hot leg but also in the cold leg. Therefore, the difference of density between 

the hot and cold legs decreases and this causes a reduction of the natural circulation driving force, 

until a point where the natural circulation is completely stopped. This happens at time 47000 s 

(37400 s), from that moment the time step in the TRACE code becomes very small, around 10-4 s.

The first magnitude that will be analyzed will be the pressure evolution with time. We observe at 

figure 3.1 that the primary pressure is kept more or less constant during this phase of the 

experiment, except when we perform an extraction. During each extraction we observe that the 

pressure diminishes to recover after a certain time. The pressure reduction is produced by the 

reduction of coolant inventory, and the recovery that follows is due to the transference of heat to 

less mass inventory (higher specific enthalpy), and the reduction of heat transference from the 

primary to the secondary through the SG due to the increase of the pressure in the secondary, 

which is produced by manually partial closing the relief valves.

Figure 3.1 Pressure evolution at the core exit. The experimental results are displayed in 

blue, and the TRACE results in red.

When figure 3.1 is observed, we notice that approximately at time t=19260 s (9660 s), the 

pressure diminish faster than in previous extractions, after that the pressure recovers again. This 

fact is produced by a coolant draining that begins at that time 19260 s and ends at 19750 s (10150 

s). The recovering of the primary pressure is due to the partial closing of the relief valve in the

secondary in order to increase the pressure in the primary. The main difficulty in simulating this 

transient is to know how much is necessary to close the relief valve to raise the pressure in the 

primary to the previous level of 12 bar. When the mass inventory in the primary diminishes, these 

adjustments in the relief valve area are more difficult to be performed because the pressure reacts 

with bigger changes to smaller changes in the relief valve flow area.

The evolution of the steam and liquid temperatures in the hot leg along the transient are displayed 

at figures 3.2 and 3.3, we notice that these variables are well predicted by the TRACE code and 

that its evolution with time is similar to the pressure evolution.

The evolution of the secondary pressure is displayed in figure 3.4. We observe that the secondary 

pressure increases continuously from 5.9 bar to about 10 bar in about 15000 s, then the 

secondary pressure remains constant during the rest of the transient. As we explained earlier, the 

continuous closing of the relief valves performed during the draining steps produces this rise in the 
3-3

secondary pressure. As we can see in figure 3.4, the secondary pressure is approximately 

predicted by the TRACE code.

Figure 3.2 Evolution with time of the steam temperature in the hot leg of the second loop. 

The blue line denotes the experimental data while the red line denotes the 

TRACE simulation.

Figure 3.3 Evolution with time of the liquid temperature in the hot leg of loop 2. The blue

line displays the experimental results.
3-4

Figure 3.4 Pressure versus time at the exit of the secondary of the steam generator in the 

four loop.

Next, we are going to comment the evolution along the transient of the natural circulation mass 

flow rate in loops 1 and 3. In figure 3.5 we display the mass flow rate evolution in loop 1, and in 

figure 3.6 we display the mass flow rate evolution in loop 3. We observe that at the beginning of 

the transient the mass flow rate is 1.21 kg/(s loop) in both loops, that is very well predicted by the 

TRACE code, that gives practically the same value. From time 10000 s (400 s), to time 20000 s 

(10400 s) the experimental mass flow rate remains constant and at time 20000 s the mass flow 

rate rise suddenly from 1.2 to 2.3 kg/(s loop). The TRACE code gives a good prediction of the 

evolution of the mass flow rate during this time interval, the only difference is that at around 16000 

s the mass flow rate rise from 1.26 to approximately 2.0 kg/(s loop), and then at t=20000 s rise 

again to 2.6 kg/(s loop). Therefore, the TRACE code predicts very well the increase in the natural 

circulation mass flow rate due to the increase of the void fraction in the core and in the hot leg. 

However, the mass flow rate starts to increase earlier in the TRACE code than in the experiment. 

Then, at time 23400 s (13850 s) the experimental mass flow rate rises again due to the fifth 

extraction and attains its maximum value of 3.1 kg/(s loop). This value is very well predicted by the 

TRACE code that gives the same value, i.e. 3.1 kg/(s loop), however the TRACE code displays 

bigger oscillations than the ones observed experimentally.

Then, at time 30000 s (20400 s), due to new extractions starts to diminish the natural circulation 

mass flow rate due to the appearance of some void fraction at the exit side of the U tubes of the 

steam generator. The progressive extractions of coolant increase more the void fraction in the cold 

leg. And at time 50000 s (40400 s), the void fraction predicted by the TRACE code in the pipe 

connecting the steam generator exit with the pump seal is 0.9, this means that the void fraction is

very high also in the cold leg, and the driving force for the natural circulation becomes very low 

and the natural circulation flow stops. The TRACE code predicts that at time 50000 s there is not 

natural circulation flow in loop 3, this result match the experimental one as displayed in figure 3.6.
3-5

Figure 3.5 Mass flow rate in kg/s versus time in the recirculation pump seal (cell 1, 

component 249) of the first loop.

Figure 3.6 Mass flow rate in kg/s versus time in the recirculation pump seal (cell 1, 

component 449) of the third loop.

Another parameter to analyze is the boron concentration at the pump seal and at the exit of the 

steam generator. In this case, it is observed in figure 3.7 that the boron concentration remains 

constant for the first 43000 s (32400 s) and then diminishes for 300 s, to recover again to its initial 

value of 2000 ppm, this small change in the boron concentration is reproduced by the TRACE 

code. At the end of the natural circulation period, the steam generator enters in reflux condenser 

conditions and starts the accumulation of condensate in the pump seal.
3-6

Figure 3.7 Boron concentration in mass fraction versus time in the pump seal of the loop 

number 3.

Next, we are going to perform a study of the void fraction evolution in the different components of 

the PKL III facility during the phase of successive extractions performed in the Test F 1.2. In the 

steam generator the void fraction start to rise in the upper part of the U tubes at time 18000 s 

(8400 s), but the first noticeable increase in the void fraction is produced at time 23000 s (13400

s), when the void fraction attains a value of 0.55, as it is displayed in figure 3.8. This increase in 

the void fraction is also produced at lower levels, but the void fraction attains logically smaller 

values as displayed in figure 3.9, where the void fraction at time 23000 s (13400 s) rise to 0.35. 

While, at the entrance of the U tubes (cell1) the void fraction is still lower at the same time and 

equal to 0.3, as displayed in figure 3.10. Therefore, a big part of the void fraction at the different 

levels of the SG is due to the steam produced in the core, only in the upper part of the U tubes the 

higher value of the void fraction is produced by a smaller collapsed water level produced by a 

smaller water inventory. The next important increase in the void fraction in the SG upper level is 

produced at time 33000 s (23400 s), when ends the sixth extraction, in this case it is produced a 

bigger increment in the void fraction that rises above 0.9. This new increase in the void fraction is 

produced by the fact that the collapsed liquid level has decreased as result of the successive 

extractions, what is noticed mainly in the upper part of the steam generator. However, in the hot 

leg the void fraction increases progressively, with the successive extractions as displayed in figure 

3.11 and at time 33000 s (23400 s) is equal to 0.5 that agrees with the value at the entrance of the 

steam generator.
3-7

Figure 3.8 Void fraction versus time in cell 11 (Upper level) of the steam generator 1.

Figure 3.9 Void fraction versus time in cell 5 (intermediate level) of the steam generator 1.
3-8

Figure 3.10 Void fraction versus time in the lower part of the SG primary in loop 1.

Figure 3.11 Void fraction versus time in the hot leg of loop 1 at component 210.

Concerning the time evolution of the void fraction in the cold leg, displayed in figure 3.12 for loop 

1, we notice that in spite of the successive extractions performed trough the vessel lower part, the 

steam generator acts as a heat sink and it condenses the steam, and as a consequence the void 

fraction in the cold leg is negligible until time 40000 s (30400 s), only when the mass inventory is 

very small start to increase the void fraction in the cold leg. At the end of the natural circulation the 

void fraction in the cold leg rises to 0.9.
3-9

Figure 3.12 Void fraction versus time in the cold leg of loop 1 at component 247.

We also notice that when is finishing the natural circulation stage and we are entering into the 

reflux condenser mode, the condensate start to accumulate in the pump seal at the cold leg, and 

the void fraction start to diminish in the RCP seal (Component 249) as it is displayed in figure 3.13. 

The void fraction in the RCP seal becomes very small at time 47000 s as displayed in figure 3.13. 

Figure 3.13 Void fraction versus time in the RCP seal, in the cold leg of loop 1.

Concerning the evolution with time of the level of condensed water in the steam generator 1, we 

notice in figure 3.14, that this level diminish continuously with the successive extractions. The 

collapsed level evolution in the steam generator is well predicted by the TRACE code, however 

the code calculations display bigger oscillations than the experimental results. The water level in 

the steam generator drops from 10.35 m at the beginning of the transient to 1 m at time 47000 s 
3-10

(37400 s), while the water level predicted by the code at this time is slightly higher, as it is 

displayed in figure 3.14.

Figure 3.14 Collapsed level of water in the steam generator of loop 1, calculated by the 

control block 1526.
4-1

4. CONCLUSIONS

In this work we have performed a simulation with the TRACE code of the coolant extraction phase

during the experiment PKL III F 1.2, until the natural circulation is completely stopped and the 

system pass from natural circulation regime to reflux condenser conditions. In general we have 

noticed a good prediction with the TRACE code of the pressure and natural circulation mass flow 

rate evolutions along this phase of test that last for 37000 s. The experiment consisted in a set of 

successive coolant extractions performed through the lower part of the reactor vessel, while the 

primary pressure was maintained constant at 12 bar by adjusting manually the stem position of the 

relief valve in the secondary. Previous to the test simulation we performed with TRACE a 

preconditioning phase to attain the initial conditions of the experiment. In table 2.3 we display the 

initial conditions for TRACE and the initial conditions for the experiment. The initial conditions for 

both cases in the main thermal-hydraulic variables are very close.

The main conclusions obtained during the analysis of this experiment performed with the TRACE 

code are:

1) We observe a progressive increase in the natural circulation mass flow rate during the set 

of extractions that finish at time 23250 s (13850 s). During this interval the natural

circulation mass flow rate evolves from 1.2 kg/(s loop), at the beginning of the test, to a 

maximum of 3.1 kg/(s loop) at time 23000 s (14400 s), this maximum value is well 

predicted by the TRACE code. The increase of the mass flow rate produced by the 

successive extractions is due to an increase in the natural circulation driving force,

produced by an increase in the void fraction in the hot leg and in the core, while the void 

fraction remains practically zero in the cold legs and in the cold side of the U tubes. This 

situation is maintained for a long time 8000 s until time 30000 s (20400 s). At this time start 

to appear small amounts of steam in the cold legs and in the cold side of the U tubes that 

reduce the natural circulation driving force.

2) At time 47000 s (37000 s) the mass flow rate is very small, this is produced by the 

reduction of the mass inventory in the primary circuit, that produces an increases of the 

void fraction that reach a value close to one in the upper part of the U tubes, that stops the 

natural circulation through the steam generator. In this situation only flows steam that 

condenses and we enter into the reflux condenser mode. This reduction in the mass flow 

rate is well predicted by the TRACE code as displayed in figures 3.5 and 3.6.

3) It is noticed a progressive pressure increase in the secondary, from 6 bar at the beginning 

of the transient to 10 bar, as displayed in figure 3.4. The reason of this is that when 

performing the successive extractions in the primary circuit, the operator must maintain 

constant the primary pressure at a value of 12 bars. To achieve this goal the operator close 

the steam relief valve in the secondary in order to rise the temperature in the secondary. 

This fact diminishes the heat transference from the primary to the secondary and as a 

consequence increases the pressure in the primary system. The TRACE simulation 

reproduces well this increase in the secondary pressure. Although the main problem is that 

is not available the flow area of the relief valves versus time, this was obtained by trial and 

error. 

4) The reduction of the natural circulation mass flow rate is observed from time 34000 s 

(24400 s), and progresses with time. The causative mechanism of this reduction is the 

appearance of voids in the cold legs as it is displayed in figures 3.12 and 3.13. This 

increase in the void fraction provokes that the natural circulation driving force diminishes. 

The TRACE code reproduces pretty well this reduction in the natural circulation mass flow 

rate, and through the analysis of the void fraction in the hot and cold legs it is possible to 

understand the origin of the reduction in the mass flow rate.
4-2

5) The collapsed level of water in the steam generator is well predicted by the TRACE code, 

as displayed in figure 3.14. The level decreases with time with periods were the level is 

practically constant, which are produced by the quasi-stationary periods between 

successive coolant extractions. During the time interval from 39000 s (29400 s) to 43000 s 

(33400 s) the steam generator water collapsed level displays big oscillations. Two draining 

of coolant are performed during this time interval. These draining performed when the 

mass inventory is small could induce these bigger oscillations of the water level in the 

steam generator.

6) It is observed in figure 3.13 that when the natural circulation diminishes and we enter in 

reflux condenser conditions the void fraction in the pump seal decreases to practically 0. at 

time 50000 s (40400 s). This is produced because the primary system has entered into 

reflux condenser condition, the natural circulation has stopped, and the condensed water 

from the steam generator started to accumulate in the RCP seal. 

In general we can say that TRACE predictions for this case are very good and the natural 

circulation mass flow rates are predicted with a great precision in comparison with previous codes
5-1

5. REFERENCES

1. Kiger, K.T. and F. Gavelli. “Boron mixing in complex geometries: flow structure details”. 

Nuclear Engineering and Design. Vol. 208, No. 1: pp. 67-85. August 2001.

2. Umminger, K., W. Kastner, J. Liebert and T. Mull “Thermal hydraulics of PWRS with respect 

to boron dilution phenomena. Experimental results from the test facilities PKL and UPTF”. 

Nuclear Engineering and Design. Vol. 204: pp. 191-203. February 2001

3. Marzo M. “Ex vessel transport and mixing of a deborated slug in a PWR primary geometry”. 

Nuclear Engineering and Design Vol. 210: pp. 169-175. December 2001

4. Wegner “PKL III RELAP5/MOD3 input model” Report Framatome ANP 

NGES1/2002/en/0059. 2002.

5. Chiva, S., J.L. Muñoz-Cobo and S. Méndez, “Análisis Post-Test del experimento de dilución 

de boro PKL E-2.1 con el código RELAP V.2 e interpretación de los resultados”. 30 Reunión 

Anual de la Sociedad Nuclear Española, 29 de septiembre – 1 octubre del 2004.




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NUREG/IA-0423 Analysis with TRACE Code of PKL-III Test F 1.2