NUREG/IA-0243 

 International 

Agreement Report

 

Development of a Vandellos II 

NPP Model using the TRACE 

Code: Application to an Actual 

Transient of Main Coolant Pumps 

Trip and Start-up 

Prepared by:

O. Lozano, C-P. Chiang, C. Llopis, 

L. Batet, F. Reventόs

Department of Physics and Nuclear Engineering

Technical University of Catalonia

ETSEIB, Av. Diagonal 647, Pav. C

08028 Barcelona

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
AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

NRC’s Public Electronic Reading Room at 

http://www.nrc.gov/reading-rm.html. Publicly released 

records include, to name a few, NUREG-series 

publications; Federal Register notices; applicant, 

licensee, and vendor documents and correspondence; 

NRC correspondence and internal memoranda; 

bulletins and information notices; inspection and 

investigative reports; licensee event reports; and 

Commission papers and their attachments.

NRC publications in the NUREG series, NRC 

regulations, and Title 10, Energy, in the Code of 

Federal Regulations may also be purchased from one 

of these two sources.

1. The Superintendent of Documents

 U.S. Government Printing Office

 Mail Stop SSOP

 Washington, DC 20402-0001

 Internet: bookstore.gpo.gov

 Telephone: 202-512-1800

 Fax: 202-512-2250

2. The National Technical Information Service

 Springfield, VA 22161-0002

 www.ntis.gov 

 1-800-553-6847 or, locally, 703-605-6000

A single copy of each NRC draft report for comment is 

available free, to the extent of supply, upon written 

request as follows:

Address: Office of the Chief Information Officer, 

Reproduction and Distribution

 Services Section

 U.S. Nuclear Regulatory Commission

 Washington, DC 20555-0001

E-mail: DISTRIBUTION@nrc.gov 

Facsimile: 301-415-2289 

Some publications in the NUREG series that are 

posted at NRC’s Web site address

http://www.nrc.gov/reading-rm/doc-collections/nuregs 

are updated periodically and may differ from the last 

printed version. Although references to material found 

on a Web site bear the date the material was accessed, 

the material available on the date cited may 

subsequently be removed from the site.

Non-NRC Reference Material

Documents available from public and special technical 

libraries include all open literature items, such as 

books, journal articles, and transactions, Federal 

Register notices, Federal and State legislation, and 

congressional reports. Such documents as theses, 

dissertations, foreign reports and translations, and 

non-NRC conference proceedings may be purchased 

from their sponsoring organization.

Copies of industry codes and standards used in a 

substantive manner in the NRC regulatory process are 

maintained at—

The NRC Technical Library 

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852-2738

These standards are available in the library for 

reference use by the public. Codes and standards are 

usually copyrighted and may be purchased from the 

originating organization or, if they are American 

National Standards, from—

American National Standards Institute

11 West 42nd Street

New York, NY 10036-8002

www.ansi.org 

212-642-4900

Legally binding regulatory requirements are stated only 

in laws; NRC regulations; licenses, including technical 

specifications; or orders, not in NUREG-series 

publications. The views expressed in contractorprepared publications in this series are not necessarily 

those of the NRC.

The NUREG series comprises (1) technical and 

administrative reports and books prepared by the staff 

(NUREG-XXXX) or agency contractors 

(NUREG/CR-XXXX), (2) proceedings of conferences 

(NUREG/CP-XXXX), (3) reports resulting from 

international agreements (NUREG/IA-XXXX), 

(4) brochures (NUREG/BR-XXXX), and 

(5) compilations of legal decisions and orders of the 

Commission and Atomic and Safety Licensing Boards 

and of Directors’ decisions under Section 2.206 of 

NRC’s regulations (NUREG-0750).

DISCLAIMER: This report was prepared as an account of work sponsored by an agency of the U.S. Government. 

Neither the U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or implied, 

or assumes any legal liability or responsibility for any third party’s use, or the results of such use, of any information, 

apparatus, product, or process disclosed in this publication, or represents that its use by such third party would not 

infringe privately owned rights.
iii

ABSTRACT 

The Thermal-Hydraulics Studies Group (GET) of the Technical University of Catalonia (UPC) 

has developed a model of Vandellòs II NPP (CNV II) for TRACE using as a previous RELAP5 

model of the plant. The model simulates the components of the primary and secondary circuits, 

along with the main heat structures (nuclear fuel, SG tubes and pressurizer heaters). Passive 

heat structures and neutron kinetics are not simulated in the model. The model includes some 

of the plant control systems. 

The main predicted steady state nominal parameters successfully match plant and design 

values. A real loss of off-site power with the subsequent Reactor Coolant Pumps start-up 

sequence (occurred in August 24th of 1993) has been simulated. The asymmetric behaviour of 

the flows in the vessel has been useful in qualifying the 3D capabilities of the TRACE VESSEL 

component. 

Mass flows of primary loops are among the most important parameters in the analyzed 

transient. The simulation results are very similar to the data recorded at the plant. The similarity 

of calculated and measured values is a guarantee of the validity of the model. 
iv
v

FOREWORD 

This report represents one of the assessment/application calculations submitted in fulfillment of 

the bilateral agreement for cooperation in thermal hydraulic activities between the Consejo de 

Seguridad Nuclear (CSN) and the US Nuclear Regulatory Commission (USNRC) in the form of 

Spanish contribution to the Code Assessment and Management Program (CAMP) of the 

USNRC, whose main purpose is the validation of TRACE code.

The CSN and UNESA (the association of the Spanish utilities), together with some relevant 

universities, have set up a coordinated framework (CAMP-Spain), whose main objectives are 

the fulfillment of the formal CAMP requirements and the improvement of the quality of the 

technical support groups that provide services to the Spanish utilities, the CSN, the research 

centers and the engineering companies

This report is one of the Spanish utilities contributions to the above mentioned CAMP-Spain 

program and has been reviewed by the AP-28 Project Coordination Committee for the 

submission to the CSN. 

UNESA 

December 2009 
vi
vii

CONTENTS

Page

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

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

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

ABBREVIATIONS........................................................................................................................ xi

ACKNOWLEDGEMENTS...........................................................................................................xiii

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

2 PLANT DESCRIPTION......................................................................................................2-1

2.1 Reactor vessel .................................................................................................................2-2

2.2 Reactor core ....................................................................................................................2-3

2.3 Primary system ................................................................................................................2-3

2.4 Steam generators ............................................................................................................2-3

2.5 Secondary system ...........................................................................................................2-3

2.6 Control and protection systems........................................................................................2-6

3 TRANSIENT DESCRIPTION .............................................................................................3-1

4 MODEL DESCRIPTION.....................................................................................................4-1

4.1 Primary hydrodynamic volumes .......................................................................................4-1

4.2 Secondary hydrodynamic volumes ..................................................................................4-3

4.3 Heat structures and power components ..........................................................................4-5

4.4 Control systems ...............................................................................................................4-6

5 SIMULATION .....................................................................................................................5-1

5.1 Steady state calculation ...................................................................................................5-1

5.2 Transient features ............................................................................................................5-1

6 RESULTS ..........................................................................................................................6-1

7 CONCLUSIONS.................................................................................................................7-1

8 REFERENCES ..................................................................................................................8-1

Figures 

Page

Figure 1. Outline of the inner part of the reactor vessel.............................................................2-2

Figure 2. General view of a steam generator.............................................................................2-5

Figure 3. General layout of the primary circuit components in the model of CNV II ..................4-2

Figure 4. Sketch of the vessel model.........................................................................................4-2

Figure 5. Represents the hydrodynamic model of one SG. The three steam generators are 

identical..............................................................................................................................4-4
viii

Figure 6. Drawing of the hydrodynamic model of the steam generators..............................4-4 

Figure 7. Sketch of the simulation of secondary system lines in the CNV II model. ..................4-5

Figure 8. Diagram of the pressurizer pressure control...............................................................4-7

Figure 9. Auxiliary Feed Water mass flow rate ..........................................................................5-4

Figure 10. Nuclear power...........................................................................................................6-1

Figure 11. Primary (mass and volume) flow rates .....................................................................6-2

Figure 12. Primary mean temperature .......................................................................................6-3

Figure 13. Primary pressure ......................................................................................................6-5

Figure 14. Pressurizer level .......................................................................................................6-5

Figure 15. Secondary pressure (SG).........................................................................................6-6

Figure 16. Steam generators level.............................................................................................6-7

Figure 17. Azimuthal downcomer flows in axial level 5..............................................................6-8

Figure 18. Azimuthal downcomer flows in axial level 7..............................................................6-8

Figure 19. Azimuthal downcomer flows in axial level 9..............................................................6-9

Figure 20. Sketch of the nomenclature followed in the VESSEL component (arrows represent 

positive directions)..................................................................................................6-9

Tables 

Page

Table I. Main features of Vandellòs II NPP................................................................................2-1

Table II. Principal events recorded during the transient of 24/08/93 in CNV II ..........................3-2

Table III. Comparison of key steady state parameters ..............................................................5-2
ix

EXECUTIVE SUMMARY

Vandellòs II is a three-looped PWR of Westinghouse design located in the northeast coast of 

Spain. It started its commercial operation in March 1988. Its nominal power is currently of 1,087 

MWe (2,940.6 MWt). 

The Thermal-Hydraulics Studies Group (GET) of the Technical University of Catalonia (UPC) 

has developed a model of Vandellòs II NPP (CNV II) for TRACE using as a previous RELAP5 

model of the plant. The TRACE model does not include all the scope of the RELAP5 model and 

is still pending of the usual and complete qualification process that the RELAP5 model has 

undertaken. 

TRACE code version used is 5.0RC3. A direct translation to TRACE has been attempted 

(excluding the reactor vessel and control blocks) using the plug-in ―R5 to TRACE‖ of the SNAP 

program. It has been found that the resulting translation had some errors that have been 

corrected by ―hand work‖. 

The model simulates the components of the primary and secondary circuits, along with the main 

heat structures (nuclear fuel, SG tubes and pressurizer heaters). Passive heat structures are 

not simulated in the model, but probably they will be incorporated in a near future. Neutron 

kinetics has not been modelled at the moment. The model includes some of the plant control 

systems (Pressurizer level and pressure, steam generator level and main feed water flow, 

turbine ...) 

The first calculation performed with the model has been that of the nominal steady state. The 

main predicted parameters successfully match plant and design values. 

New TRACE model for CNV II has an advantage compared to the previous RELAP5/Mod3.2 as 

it simulates the reactor pressure vessel using a three dimensional component (the VESSEL 

component of TRACE). 

A real loss of off-site power (occurred in August 24th of 1993) has been simulated using the 

TRACE model. The event produced the Reactor Coolant Pump (RCP) trip and, after this, the 

reactor and turbine trips. The subsequent RCP start-up sequence produced an asymmetric 

behaviour of the flows in the vessel which has been useful in qualifying the 3D capabilities of the 

TRACE VESSEL component. 

Transient data (with a 2 seconds record frequency) are available from a total of 95 variables. 

The TRACE calculation of the transient has been done imposing, on the one hand, the thermal 

power in the core and, on the other hand, the Auxiliary Feed Water mass flow rate (different in 

each of the steam generators), in order to reproduce the plant conditions. 

Mass flows of primary loops after the shutdown and subsequent start-up of the pumps are 

among the most important parameters in the analyzed transient. The simulation results are very 

similar to the data recorded at the plant. 

The signs and the magnitudes obtained for azimuthal mass flows in the downcomer of the 

vessel show significant consistency. 
x

The results that best fit the data recorded at the plant are SG pressures, SG levels and primary 

average temperature. This shows a certain quality of the developed model. 

Regarding average temperature, TRACE model values are higher than those of the plant, which 

is explained by the fact that the average temperature control system is, by now, not included in 

the model. 

The similarity of calculated and measured values is a guarantee of the validity of the model. 
xi

ABBREVIATIONS 

AFW Auxiliary Feed Water 

ANAV Asociación Nuclear Ascó-Vandellòs 

CNV II Vandellos II NPP 

GET Thermal-Hydraulic Studies Group (UPC) 

MFW Main Feed Water 

NPP Nuclear Power Plant 

PWR Pressurized Water Reactor 

RCP Reactor Coolant Pumps 

RELAP Reactor Excursion and Leak Analysis Program 

SG Steam Generator 

UPC Universitat Politècnica de Catalunya (Technical University of Catalonia)

UNESA Asociación Española de la Industria Eléctrica (Association of the Spanish 

Utilities)
xii
xiii

ACKNOWLEDGEMENTS 

Authors are grateful to Association of the Spanish Utilities (UNESA) for funding the work 

presented in this report and to Asociación Nuclear Ascó-Vandellós (ANAV) for their continuous 

help and cooperation. Specially, authors recognize the valuable advices from Mr. Ismael Sol 

(ANAV).
1-1

1 INTRODUCTION

The Thermal-Hydraulics Studies Group (GET) of the Technical University of Catalonia (UPC) 

collaborates since 1991 with the operation responsible of Catalan Nuclear Power Plants in 

different tasks related to operation support 78. Among them, UPC’s GET takes care of 

maintaining the thermal-hydraulic models of Ascó (two units) and Vandellòs II NPP that have 

been prepared with the aim of simulating a wide variety of transients using RELAP5/Mod3.2. 

GET has also participated, in the past, in several activities related to the development of the socalled Consolidated Thermal-Hydraulic Code that culminated in the current status of the TRACE 

code 10. 

Recently, due to an agreement with the ―Asociación de la Industria Eléctrica Española 

(UNESA)‖, GET has developed a model of Vandellòs II NPP (CNV II) using TRACE. The model 

is considered preliminary because it does not include all the scope of the RELAP5 model and 

also because it is still pending of the usual and complete qualification process 9. 

The TRACE version used is 5.0RC3. In preparing the model, use has been made of SNAP 

program in its version 0.26.6. The ―plug-in‖ R5 to TRACE, version 0.1.3., and the AptPlot, 

version 6.1.1., for post-process are other used pieces of software. 

The first step has been to predict a nominal steady state using the preliminary model of CNV II. 

The main predicted parameters successfully match plant and design values. 

TRACE model of CNV II has an advantage compared to that of RELAP5/Mod3.2, due to the fact 

that TRACE code allows simulating the vessel using a three dimensional component. 

A loss of off-site power (occurred in August 24th of 1993) is the transient simulated. The event 

produced the Reactor Coolant Pump (RCP) trip and, after this, the reactor and turbine trips. The 

system went to hot zero power and, afterwards, the RCP start-up sequence took place. The 

asymmetric nature of the transient helps qualifying 3D capabilities of the TRACE code. 
2-1

2 PLANT DESCRIPTION

Vandellòs II is a three looped PWR NPP of Westinghouse design owned by Endesa (72%) and 

Iberdrola (28%), and operated by ANAV 1. It is located near Tarragona, in the northeast of 

Spain, and uses the Mediterranean Sea water as heat sink. It started its commercial operation 

in March 1988. Its nominal power is currently of 1,087 MWe (2,940.6 MWt). 

 

The reactor vessel is Westinghouse designed. The plant has three Westinghouse (model F) 

steam generators of U-tubes kind without pre-heaters. Feed water enters directly the upper part 

of the downcomer through J-shaped nozzles. 

The main features of the plant are shown in Table I: 

Table I. Main features of Vandellòs II NPP 

Reactor thermal power (MWt) 2940.6

Electrical power (MWe) 1087

Fuel UO2

Number of fuel bundles 157

Number of cooling loops 3

Reactor operating pressure 

(MPa) 15,4

Mean coolant temperature (K):

 Hot zero power

 Full power

564,8

582,3

Steam generator (SG) Westinghouse 

tipo F

Number of tubes in one SG 5626

Total tubes length in one SG 

(m) 98759

Tubes inner diameter (m) 0.0156

Tubes wall material INCONEL

Coolant Recirculation Pumps Westinghouse D 

100

Volume of the primary (m3) 106.19

Volume of pressurizer (PZR) 

(m3)

39.65

PZR heaters power (kW) 1400

Some of these general features depend on the plant configuration. The table refers to the 

current one. The analyzed transient occurred before different technological improvements took 

place in the plant; related changes in the model will be presented in the appropriate section. 

The most relevant features of main equipment of the Nuclear Steam Supply System (NSSS) are 

presented below 12. 
2-2

2.1 Reactor vessel

The reactor vessel of Vandellòs II NPP is that of a three looped PWR. It is a cylindrical vessel 

with a hemispherical lower plenum and removable bolted cover which ensures water tightness. 

It contains the core, the structure supporting it, the control rod banks, the thermal shielding and 

other components directly related to the core. 

The three inlet and the three outlet nozzles are located between the main flange and the upper 

part of the core. The vessel is designed and manufactured following the specific requirements of 

ASME code for nuclear equipment. 

The vessel is made of carbon steel. The inner surfaces in contact with the cooling water are 

coated with a layer of austenitic stainless steel in order to minimize corrosion. 

FigureFigure 1 shows an outline of the inner part of the reactor vessel.

Figure 1. Outline of the inner part of the reactor vessel. 
2-3

2.2 Reactor core

The reactor core consists of 157 fuel assemblies of different enrichment and burnup. Each fuel 

assembly has 264 active fuel rods, with the corresponding Zircaloy-4 cladding, which contain 

the UO2 pellets. 

2.3 Primary system

Primary pipes (hot, intermediate and cold legs) are made of steel. Their diameter is 74 cm. Hot 

legs length is 7.3 m. 

Each loop of Vandellòs II NPP Reactor Coolant System has a one stage centrifugal vertical 

pump, with a system controlling seal leakage. The main pumps supply the water flow needed to 

cool the reactor. 

The pressurizer is the component which controls the pressure in the Reactor Coolant System 

during normal operation and helps smoothing pressure changes during transients. Vandellòs II 

NPP pressurizer houses 78 vertical type heaters, with a total power of 1400 kW. 

2.4 Steam generators

Steam Generators (SG) are the components which perform the heat transfer from primary to 

secondary circuits and produce the steam needed in the turbine. Each of the three SG 11 has 

5,626 tubes made of INCONEL. The tubes are inverted U type, with an inner diameter of 1.56 

cm and a total length of 98,759 m. 

Figure 2 shows the outline of the SG. 

The lowest part of SG primary side is hemispherical and split in two by a plate. The reactor 

coolant comes in through one of the mentioned two halves, then through the Inconel tubes and 

finally comes out through the other half. The heat from the reactor coolant is transferred through 

the wall of the tubes. 

The secondary side is a shell that acts as a pressure barrier around the tube bundle. Separators 

and driers are located in the upper part. The water supply comes through the feed water nozzle 

and the feed ring designed to prevent water hammer.

The heat transferred by the reactor cooling water through U tubes raises the temperature of the 

secondary side to saturation, producing steam with high moisture content. After this, the steam 

goes through turbo separators and dryers that remove liquid water from vapour flow to the 

turbine. The water separated by the separators mixes with the main feed water and comes 

through a downcomer to the inlet of the tube bundle.

2.5 Secondary system

The main feed water (MFW) is driven by two turbo pumps. Pressure and flow of MFW system is 

controlled through the speed of turbo pumps and the opening of MFW valves. 
2-4

The Auxiliary Feed Water (AFW) system is driven by one turbo pump and two motor pumps. 

The flow is manually controlled through AFW valves. 

The current turbine of Vandellòs II NPP consists of one high pressure and three low pressure 

sections mechanically coupled by a shaft to the electrical generator.

Moisture Separators Reheaters are located between the high and low pressure turbine 

components. The steam from the header enters the high pressure turbine component through 4 

inlet valves, which control the steam flow. In series, the stop valves are designed to isolate and 

trip the turbine. 

The role of the steam dump system is to discharge directly to the condenser the power balance 

between nuclear and turbine. Its normal operation is in temperature mode (its performance is 

determined by changes in the reactor average temperature). It can also work in pressure mode. 
2-5

Figure 2. General view of a steam generator. 
2-6

2.6 Control and protection systems 

The plant has, among others, the following control and protection systems 4, 5, 13: 

 Reactor control system (by average temperature and power mismatch) 

 Primary system pressure control 

 Pressurizer pressure control system 

 Steam dump control system 

 Feed water control system 

 Reactor protection system 

 Pressurizer relief and safety valves system 

Steam generators relief and safety valves system
3-1

3 TRANSIENT DESCRIPTION 

The analyzed transient occurred on August 24, 1993 in Vandellòs II NPP, when a momentary 

loss of offsite power caused the shutdown of the main reactor coolant pumps (RCP) and 

subsequently the trip of the reactor and the turbine. The system went to hot stop and the 

operators proceed to the sequential RCP’s start-up.

Data from the transient are available. The available records cover the period of time between 

16:15:00 on the 24th until 17:30:00 on the same day, which means a total of 4500 seconds, of 

which 618 seconds belong to the steady-state period (before the loss of supply) and the rest to 

the transient.

The data come from the data acquisition system of CNV II 3. Measurements are available from 

a total of 95 variables. The frequency of recording information available is 2 seconds and so a 

total of 2251 values for each studied variable are considered. The records of the data 

acquisition system do not include the start-up of the last coolant pump, in particular that of the 

first loop, which was started after 17:30:00. 

The transient began at 16:25:18. Henceforth, and to facilitate the understanding of the timing, 

the origin of time (t = 0s) is fixed at 16:15:00, the time at which the recorded data are available. 

During the time between t = 0s and t = 618s, the plant was operating at full power (2775 MW on 

the date of the transient) in steady-state condition and at nominal levels. 

At time 618 s a sudden and temporary loss of offsite power, caused a rapid loss of tension in 

the main and auxiliary transformers that connect the plant with the outside electrical grid. This 

loss of electricity caused the shutdown of the turbine, the RCPs and the reactor. 

Also, the pressurizer heaters and the steam-dump system became unavailable. 

Following the shutdown of the coolant pumps, the water flow in the three primary loops 

decreased and natural circulation was established. 

After the turbine trip, the reference temperature becomes that of zero power and the average 

temperature tends to that value. This decrease in coolant temperature leads to an increase in 

density and, therefore, to a depressurization of the primary system. 

The turbine and reactor trip caused the isolation of the main feed water and the start of the two 

motor pumps and the turbo pump of the AFW system which, from now on, provides water to 

steam generators. Modulation of AFW is done manually. 

At time 1522 seconds, RCP-C (the coolant pump of the third loop) is manually started, reaching 

a flow rate in the 3rd loop around 107%. The start of this pump causes the appearance of 

reverse flow in the other two loops. 

At time 2320 seconds, operators manually isolate the main steam. Secondary pressure and 

temperature are recovered. 
3-2

The start-up of the second coolant pump (RCP-B) is carried out manually at 2584 seconds. This 

brings water flows in loops 2 and 3 to 106% of the nominal value and increases the reversed 

flow in the first loop. 

From then until the end of the available data (4500 s), the system evolves towards its nominal 

values, recovering the SG levels. 

Table II shows the events of the transient: 

Table II. Principal events recorded during the transient of 24/08/93 in CNV II 

Time (s) Event

0 Start of data recording (16:15:00)

618 Momentary loss of offsite power (16:25:18)

620 RCPs trip

620 Turbine and reactor trip

635 Manual Control of the Auxiliary Feed Water system

1522 Start-up of RCP C (third loop)

2320 Main Steam isolation

2584 Start-up of RCP B (second loop)

4500 End of data (17:30:00)
4-1

4 MODEL DESCRIPTION 

The following are the main features of the model of Vandellòs II NPP using the code TRACE. 

This model has been developed from the existing for RELAP5, which has proven its 

performance. Both models share technical information and the corresponding descriptive 

document 2. 

A direct translation to TRACE has been tried using the plug-in ―R5 to TRACE‖ of the SNAP 

program. This has been done excluding the reactor vessel, as it is three-dimensional in TRACE 

and one-dimensional in RELAP5. Control blocks have also been excluded because the 

translation does not perform correctly for this particular. 

This way, a TRACE model was obtained for the primary and the secondary systems. This 

model, a priori, only needed to be checked and corrected for the elevations of the different 

components to be correct. It was found that the resulting translation had many additional errors. 

The errors (and the solution applied) are the following: 

 Some junctions appeared twice after translation; so, some of them had to be deleted.

 The friction coefficient obtained by translation was zero on all sides of the nodalization 

volumes of each component. The present value of these coefficients has been 

introduced based on the available information of the RELAP5 model. 

The translation does not cover the difference between RELAP5 and TRACE in terms of 

homologous curves of pump components. Overlapping or null values were found after 

translation, as well as inconsistencies in the figures. Such curves have been 

implemented from scratch according to the values given in the descriptive document 2. 

 The correspondence between the types of valves in both codes is not direct. For this 

reason some of them have been changed. 

To describe in more detail the model, the following parts will be considered: 

 Primary hydrodynamic volumes 

 Secondary hydrodynamic volumes 

 Heat structures and power components 

 Control systems 

4.1 Primary hydrodynamic volumes

The model of the primary circuit for TRACE includes the following main parts: 

 Reactor vessel 

 Primary loops, including SG U-tubes

 Reactor Coolant Pumps 

 Pressurizer 

Figure 3 shows an outline of the general layout of the components of the primary circuit. 

The main features of the model for the primary circuit of CNV II are presented below. 
4-2

To simulate the reactor pressure vessel the TRACE VESSEL component has been used. This 

component allows the simulation of a 3-dimensional region. For the model of CNV II, 12 axial 

levels, 2 radial and 3 azimuthal nodes have been considered (Figure 4). 

Carga / 

CVCS descarga

Lazo 2

Lazo 1

presionador

C.N. VANDELLÒS II

turbina

descarga al 

condensador

consumos propios

rociado

GV 1

A.A.

GV 2

A.A.

GV 3

A.A.

220

320

420

450 460

465 470 475

350 360

365 370 375

250 260

265

252

270 275

230 240 330 340 430 440

510

520

645 745 845

410310210

219

319

419

NODALIZACION GENERAL

2

555

366 367

466 467

Figure 3. General layout of the primary circuit components in the model of CNV II 

Figure 4. Sketch of the vessel model 
4-3

The downcomer is located in the outer ring from axial level number 2 to level 9. The core takes 

axial levels from 3 to 8 of the inner part of the vessel.

The primary circuit of CNV II includes three loops. Each loop consists of the following parts and 

components: 

1. Vessel outlet 

2. Hot leg 

3. SG inlet 

4. SG tubes 

5. SG outlet 

6. Intermediate leg 

7. RCP

8. Cold leg 

9. Vessel inlet 

RCPs are modelled using the PUMP component of TRACE. The values of the RCP technical 

features come from design information (nominal speed, friction torque, torque, inertia ...). The 

homologous curves correspond also to the design. 

The pressurizer is connected to hot leg 3 by the TEE 419, as can be seen in Figure 3. The 

specific component PRESSURIZER available in the code has not been used. The two 

components used (510 and 520) are PIPEs. 

The pressurizer spray system has been simulated in a simplified way using components FILL of 

TRACE. 

4.2 Secondary hydrodynamic volumes

The model of the secondary circuit for TRACE includes the following main parts: 

SG

 MFW and AFW systems 

 Steam lines and relief valves 

 Turbine 

 Steam-Dump 

 Steam consumptions 

The secondary of the steam generators includes the following parts and components: 

1) FW feed ring

2) downcomer 

3) boiler

4) turbo separators 

5) driers and steam dome 
4-4

GENERADORES DE VAPOR

601

610

de rama caliente 230 240 a rama fría

salida vapor

250

600

231

620

630

640

645

220

650

622

678 675

677

676

1345

1223

AAP

AAA

650 750 850

677 777 877

676 776 876

1345 2345 3345

678 778 878

675 775 875

1223 2223 3223

645 745 845

640 740 840

630 730 830

622 722 822

620 720 820

610 710 810

601 701 801

600 700 800

240 340 440

231 331 431

230 330 430

GV1 GV2 GV3

primario secundario AA

Figure 5. Represents the hydrodynamic model of one SG. The three steam generators are 

identical. 

The mass of vapour and liquid water inside of SG, as well as the recirculation rate have been 

checked and they agree with design information. 

MFW and AFW are simulated as boundary conditions.

GENERADORES DE VAPOR

601

610

de rama caliente 230 240 a rama fría

salida vapor

250

600

231

620

630

640

645

220

650

622

678 675

677

676

1345

1223

AAP

AAA

650 750 850

677 777 877

676 776 876

1345 2345 3345

678 778 878

675 775 875

1223 2223 3223

645 745 845

640 740 840

630 730 830

622 722 822

620 720 820

610 710 810

601 701 801

600 700 800

240 340 440

231 331 431

230 330 430

GV1 GV2 GV3

primario secundario AA

Figure 6. Drawing of the hydrodynamic model of the steam generators 

The model includes the steam lines, the steam isolation valves, the steam header, the turbine 

valves (for control and for stop), the steam-dump (simplified form) and steam consumptions. 
4-5

Figure 7 shows a nodalization diagram of the TRACE model. 

GENERADOR

VAPOR 1

SECUNDARIO: LINEAS, VALVULAS ALIVIO, TURBINA, STEAM-DUMP Y CONSUMOS PROPIOS

650

653

654

900

1

907

turbina

steam dump

consumos propios

910

912

917

939

940

668

GENERADOR

VAPOR 2

750

753

754

755

767

768

GENERADOR

VAPOR 3

850

853

854

868

652 752 852

865

866

765

766

665

666

Figure 7. Sketch of the simulation of secondary system lines in the CNV II model.

4.3 Heat structures and power components

Next, the main features of the different heat structures and power components used in the 

TRACE model are described. The main heat structures modelled are: 

 Heat transfer from primary to secondary sides through SG tubes 

 Heat transfer from nuclear fuel to the primary coolant 

 Pressurizer heaters 

 

Other components such as high and low pressure MFW heaters are not simulated. 

The heat structures that could represent the heat losses from the different components (vessel, 

steam generators, piping, etc.) to the environment are not simulated. The passive internal heat 

structures are not part of the model. The former have a limited impact on the transient analysis. 

The latter are awaiting the results of other studies, carried out in parallel, with RELAP5 and will 

probably be implemented in the near future.
4-6

The U-tubes are simulated by 3 heat structures, one for each loop and steam generator. Each 

one consists of 14 nodes: nodes 1 to 6 correspond to the ascending part of the tubes; nodes 7 

and 8 correspond to the upper curve of the tube; and nodes 9 to 14 to the descending part.

The material of the pipes is INCONEL 600. The table of properties of this material has been 

obtained from the RELAP5 model.

The heat structures and power components modelling heat transfer between the fuel and 

primary circuit are justified below. Taking into account that the 3D vessel has 3 azimuthal 

sectors, 3 cylindrical heat structures have been modelled. Each heat structure corresponds to a 

power component. These heat structures simulate fuel rods. Each structure is divided in 6 axial 

nodes. 

Finally, there is a power component to provide the heating of the 3 heat structures that simulate 

the fuel rods. Power has been introduced as a boundary condition. The reactor kinetics has not 

been modelled in this preliminary model of CNV II because in the transient discussed below 

kinetics plays a minor role. For this particular analysis, the residual power has been introduced 

as a boundary condition using a table.

Backup and control heaters are simulated as a single heat structure, associated with a power 

component. The pressure control system determines the power of this component. 

4.4 Control systems

The model includes the following control systems 6: 

 Pressurizer level 

 Pressurizer pressure 

 SG level and MFW flow 

 Turbine control 

 Secondary relief 

 Steam-dump 

As an example the diagram of the pressurizer pressure control system is shown in Figure 8. 
4-7

Figure 8. Diagram of the pressurizer pressure control 

PI 

Ppres

(30)

157.2kg/cm2

(-23)

k=0.01

tau=10

cte=6

1.0197e-5

(-36)

-35 -25

3.25

(-33)

-26 -34 -37

·0.5

1.17

1.73

-32

-39

(-1.05;1.0)

(1.0597;0.0)

-38

3.3937

(-44)

-22

-45

-46
5-1

5 SIMULATION 

5.1 Steady state calculation

The model described above has been used for calculating a steady-state. 

Table III shows the comparison of the values of key parameters in the model with those of the 

plant (collected by instrumentation or established in the design).

5.2 Transient features

In Table II the chronology of events corresponding to the analyzed transient has been shown. 

Main boundary conditions have to do with the following areas: 

 

 Reactor power 

 Trip and start-up of RCP 

 Turbine trip 

 Steam isolation 

 MFW and AFW 

 Steam-dump 
5-2

Table III. Comparison of key steady state parameters 

Plant TRACE

Absolute 

Error

Relative Error 

(%)

Nuclear power (%) 100.56 99.43 -1.13 -1.12

Temperature hot leg 1 

(K) 598.55 603.04 4.50

not applicable 

(na)

Temperature hot leg 2 

(K) 598.12 602.33 4.21 na

Temperature hot leg 3 

(K) 599.29 602.39 3.10 na

Mean temperature loop 

1 (K) 581.92 585.22 3.31 na

Mean temperature loop 

2 (K) 581.29 584.75 3.46 na

Mean temperature loop 

3 (K) 581.85 584.86 3.01 na

Temperature cold leg 1 

(K) 565.29 567.41 2.12 na

Temperature cold leg 2 

(K) 564.46 567.16 2.71 na

Temperature cold leg 3 

(K) 564.40 567.32 2.92 na

Pressurizer pressure 

(kPa) 15518.26 15513.50 -4.75 -0.03

Pressurizer level (%) 59.10 59.97 0.87 1.47

Pressure SG 1 (kPa) 6747.78 6770.45 22.66 0.34

Pressure SG 2 (kPa) 6726.34 6742.43 16.09 0.24

Pressure SG 3 (kPa) 6949.33 6745.42 -203.91 -2.93

Level SG 1 (%) 51.10 50.00 -1.10 -2.14

Level SG 2 (%) 49.79 50.00 0.21 0.41

Level SG 3 (%) 49.37 50.00 0.63 1.27

Steam flow 1 (kg/s) 521.21 514.13 -7.08 -1.36

Steam flow 2 (kg/s) 526.54 512.49 -14.05 -2.67

Steam flow 3 (kg/s) 535.04 505.54 -29.50 -5.51

Recirculation ratio* 3.27 3.13 -0.14 -4.23

 * Design 

Reactor power

In the steady-state, the reactor is at full power (100%). The value of power used is 

2798 MWth. This value was used in a preliminary RELAP5 calculation and is maintained for 

possible comparisons RELAP5-TRACE, despite of being slightly higher than the other two 

values referenced below: 

 2775 MWth (nominal value for the configuration of the plant when the transient 

took place) 

2791.2 MWth (value obtained performing an energy balance calculation) 

 
5-3

Relative errors, respectively 0.8% and 0.2%, are considered low enough. 

Power values after the reactor trip are obtained from a RELAP5 calculation. 

Trip and start-up of RCP 

In order to simulate the trip and start-up of RCPs, the following logical variables haven been 

defined:

 variable 1: it stops RCP 260 (loop 1) at time t=620 s.

 variable 2: it stops RCP 360 (loop 2) at time t=620 s and starts it at time 

t=2584 s.

 variable 3: it stops RCP 460 (loop 3) at time t=620 s and starts it at time 

t=1522 s.

Turbine trip

The turbine trips at time 620s.

Pressurizer heaters control

Pressurizer heaters are disconnected at time 620 s and they connected again at time 700 

s. 

Steam isolation

Steam isolation valves remain open until time t=2320 s and close after this time. 

MFW

MFW stops at time t=620 s. 

Agua de Alimentación Auxiliar

0

5

10

15

20

25

30

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) Caudal (kg/s)

AAA2 planta

AAA2 trace

Figure 9. Auxiliary Feed Water mass flow rate 
5-4

Agua de Alimentación Auxiliar

0

5

10

15

20

25

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) Caudal (kg/s)

AAA1 planta AAA1 trace

Figure 9. Auxiliary Feed Water mass flow rate (Continued)

AFW

In order to control AFW flow, plant data have been introduced using TRACE FILL 

components which simulate the AFW inlet. Figure 10 shows the corresponding curves. 

Steam dump

The steam dump is not available due to the loss of offsite power. 

Agua de Alimentación Auxiliar

0

5

10

15

20

25

30

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) Caudal (kg/s)

AAA3 planta AAA3 trace
6-1

6 RESULTS 

In this last section, the main results of the simulation are presented and discussed. 

As stated above, nuclear power has been introduced according to the results obtained with 

RELAP5 and it substantially conforms to that of the plant. 

Figure 10 shows the plant nuclear power compared with that introduced in the TRACE model. It 

has been assumed that the calculated steady-state power fits the average power of the reactor. 

In the long term, a difference exists between registered data and TRACE values, due to the fact 

that the former are estimated from neutron detectors in the power range and are not accurate 

enough. 

Potencia nuclear

0

20

40

60

80

100

120

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) potencia (%)

planta

trace

Figure 10. Nuclear power
6-2

Caudales del primario

-40

-20

0

20

40

60

80

100

120

140

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) caudal (%)

caudal1 planta

caudal2 planta

caudal3 planta

caudal1 trace

caudal2 trace

caudal3 trace

lazo 3 lazos 2 y 3

lazo 1 lazos 1 y 2

Caudales del primario

-40

-20

0

20

40

60

80

100

120

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) caudal (%)

caudal1 planta

caudal2 planta

caudal3 planta

caudal1 trace

caudal2 trace

caudal3 trace

lazo 3 lazos 2 y 3

lazo 1 lazos 1 y 2

Figure 11. Primary (mass and volume) flow rates

Figure 11 compares the evolution of the predicted flow in the three loops with data collected in 

the plant.
6-3

It should be noted that the plant flow measurement is based on Δp at the elbows and the results 

are given in percentages. Just as with nuclear power, it has been assumed that the average 

flow (in kg/s) of the first 618 seconds is the average flow in the plant for each loop. 

It should be also noted that the measurements at the plant for low flows are unreliable and that 

the flow meters always give positive values. With these considerations, in view of Figure 11, it 

can be concluded that the simulation results are significantly similar to the measured values. 

Regarding average temperature, TRACE model values are higher than those of the plant. This 

deviation will not be treated as part of the scope of this study since the average temperature 

control system is, by now, not included in the model. 

As shown in Figure 12, at the beginning of the transient, a sharp drop in average temperature 

takes place as a result of tripping the reactor. It is noted that the magnitude of the fall obtained 

by TRACE is similar to that recorded at the plant. 

Temperatura media primario

530

540

550

560

570

580

590

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) temperatura (K)

planta

trace

Figure 12. Primary mean temperature

At the time the transient took place, temperature was measured in the plant using 

thermocouples located in small pipes derived from the main loops. If forced circulation was lost, 

actual temperature was not reflected in the thermocouples. This explains the discrepancy 

between the results and the plant data. When forced circulation is recovered temperature values 

become similar. 
6-4

Finally, it should be noted that Figure 12 shows the only values of average temperature that 

were available. Those values were collected by an instrument that was not able to record 

temperatures below 275 o C. 

Figure 13 shows the sharp decrease of primary pressure at the beginning of the transient. The 

simulation reproduces it, but slightly more pronounced. 

In addition to the first pressure drop (1), Figure 13 shows the effect of the restart of RCP-C. The 

event (2) is the isolation of steam which, in TRACE calculation, causes a pressure increase 

more significant than in the plant. Finally, event (3) is the restart of RCP-B.

Pressurizer level follows pressure and temperature time trend. 

The pressurizer level initially shows a sharp decrease and then evolves as a function of average 

temperature as well as of the flow of the chemical and volume control system, as shown in 

Figure 14. 

The simulation results are similar to those recorded in the plant. The model shows a larger 

decrease in the level, as it also happens with the primary pressure. 

Pressurizer level follows pressure and temperature time trend. 

The pressurizer level initially shows a sharp decrease and then evolves as a function of average 

temperature as well as of the flow of the chemical and volume control system, as shown in 

Figure 14. 

The simulation results are similar to those recorded in the plant. The model shows a larger 

decrease in the level, as it also happens with the primary pressure. 

Figure 15 shows that, at the beginning of the transient, after the loss of offsite power, a sudden 

increase of secondary pressure occurs, which is due to the closure of turbine inlet valves. Later 

on, this pressure slowly decreases due to the AFW contribution and to the extraction of steam 

consumptions. 

When RCP C restarts there is a small pressure increase with the improvement of heat transfer 

from primary to secondary. 

When the steam is manually isolated, the pressure is recovered as there is no extraction of 

steam. 

The simulation offers for SG pressure time trend is very similar to that recorded at the plant. 
6-5

Presión del primario

1,00E+07

1,10E+07

1,20E+07

1,30E+07

1,40E+07

1,50E+07

1,60E+07

1,70E+07

1,80E+07

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) presión (Pa)

planta

trace

1 2

3

Figure 13. Primary pressure

nivel presionador

0

10

20

30

40

50

60

70

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) nivel (%)

planta

trace

 

Figure 14. Pressurizer level
6-6

Presión GV3

0,00E+00

1,00E+06

2,00E+06

3,00E+06

4,00E+06

5,00E+06

6,00E+06

7,00E+06

8,00E+06

9,00E+06

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) presión (Pa)

p GV3 planta

p GV3 trace

Presión GV1

0,00E+00

1,00E+06

2,00E+06

3,00E+06

4,00E+06

5,00E+06

6,00E+06

7,00E+06

8,00E+06

9,00E+06

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) presión (Pa)

p GV1 planta

p GV1 trace

Figure 15. Secondary pressure (SG)

As can be seen in 

Figure , the SG level (measured in narrow range) shows a sharp decrease at the beginning of 

the transient due to the reactor trip, the increase of secondary pressure, and the consequent 

void collapse. 

In the long term the AFW systems allows the SG levels recovery.

The evolution of the SG level obtained with the TRACE model is essentially the same as that 

recorded at the plant. This similarity is especially remarkable when one considers that the 

measurement range of this level is relatively small. 

Presión GV2

0,00E+00

1,00E+06

2,00E+06

3,00E+06

4,00E+06

5,00E+06

6,00E+06

7,00E+06

8,00E+06

9,00E+06

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) presión (Pa)

p GV2 planta

p GV2 trace
6-7

Nivel de del GV1

-10

0

10

20

30

40

50

60

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) nivel (%)

nivel GV1 planta

nivel GV1 trace

Nivel del GV2

-10

0

10

20

30

40

50

60

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) nivel (%)

nivel GV2 planta

nivel GV2 trace

Figure 16. Steam generators level

Figures 16, 17 and 18 show the azimuthal mass flows at different downcomer axial levels. In 

these figures, ―sector n‖ mass flow is the mass flow from ―sector n‖ to ―sector n+1‖, with the 

direction ―n‖ to ―n+1‖ positively defined (see Figure 19). 

When the situation is symmetrical, either at full power (period 0-600 seconds) or when the three 

RCPs are stopped (period 600-1500 seconds), the azimuthal flows are virtually null. 

Between 1500 and 2600 seconds, RCP-3 is running while the other two are stopped. There is 

inverted flow in loops 1 and 2 and therefore the nozzle 3 drives cold water through the junction 

from 3 to 1 (named ―sector 3‖ in the figures) with a positive flow and through the junction from 3 

to 2 (named ―sector 2‖) with a negative flow. The flow at the junction from 1 to 2 (named ―sector 

1‖) is less visible and near zero. 

Nivel del GV3

-10

0

10

20

30

40

50

60

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) nivel (%)

nivel GV3 planta

nivel GV3 trace
6-8

caudales azimutales del downcomer (nivel axial 5)

-100

-80

-60

-40

-20

0

20

40

60

80

100

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) caudal (kg/s)

sector1

sector2

sector3

Figure 17. Azimuthal downcomer flows in axial level 5 

caudales azimutales del downcomer (nivel axial 7)

-200

-150

-100

-50

0

50

100

150

200

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) caudal (kg/s)

sector1

sector2

sector3

Figure 18. Azimuthal downcomer flows in axial level 7
6-9

caudales azimutales del downcomer (nivel axial 9)

-1500

-1000

-500

0

500

1000

1500

0 500 1000 1500 2000 2500 3000 3500 4000 4500 5000

tiempo (s) caudal (kg/s)

sector1

sector2

sector3

Figure 19. Azimuthal downcomer flows in axial level 9

 

Figure 20. Sketch of the nomenclature followed in the VESSEL component (arrows 

represent positive directions)

sector 1

3

2

junction 1

junction 3
6-10

Comparing the different axial levels, it can be verified that, in all cases, the azimuthal flows are 

higher when they are closer to nozzles. 

This follow-up to the hydrodynamic behaviour of the downcomer in symmetry and asymmetry 

condition may appear irrelevant but it connects with the interest of accident scenarios in which 

the effect of core bypass should be properly assessed.
7-1

7 CONCLUSIONS

This report reflects how the TRACE model of CNV II has been developed, how the steady state 

has been checked and how the transient has been calculated and its results compared with 

data recorded in the plant. 

The model includes the main systems of Vandellòs II NPP. The TRACE version used is 5.0RC3. 

Most of the information of the plant comes from an existing RELAP5 model. The only brand new 

area of development is the 3D vessel. A direct translation to TRACE has been tried using the 

plug-in ―R5 to TRACE‖ of the SNAP program but some errors have been detected that have 

been fixed: 

 Some junctions appeared twice after translation 

 The friction coefficient obtained by translation was zero on all sides of the nodalization 

volumes of each component. 

The translation of homologous curves of pump components was not done correctly.

 The correspondence between the types of valves in both codes is not direct. 

It should be noted that among the most important parameters in the analyzed transient there are 

the mass flows of primary loops, after the shutdown and subsequent start-up of the pumps. The 

simulation results are very similar to the data recorded at the plant. 

The signs and the magnitudes obtained for azimuthal mass flows in the downcomer of the 

vessel show significant consistency. 

The results that best fit the data recorded at the plant are SG pressures, SG levels and primary 

average temperature. This shows a certain quality of the developed model. It should be noted 

that SG level obtained is very similar to that of the plant. Since the narrow range for the level 

measurement is small, the results are relevant. 

The similarity of results obtained basically for average temperature and primary and secondary 

pressure guarantee the validity of the model. 
8-1

8 REFERENCES 

1. ANAV, web site of Asociación Nuclear Ascó – Vandellòs (ANAV), www.anav.es (2008) 

2. C.N. Vandellòs II, (in Spanish) Nota de cálculo del modelo de C.N. Vandellòs II con 

RELAP5/Mod3.2

3. C.N. Vandellòs II, data recorded in the plant by the Data Acquisition System 

4. C.N. Vandellòs II, Logics Diagrams 

 

5. C.N. Vandellòs II, Loop Diagrams 

6. C. Llopis, (in Spanish) Modelos avanzados de sistemas de control y protección de una 

central nuclear de agua a presión: contribución a la seguridad y a la disponibilidad, PhD 

thesis UPC, 2006.

7. C. Llopis, F. Reventós, L. Batet, C. Pretel, I. Sol; Analysis of low load transients for the 

Vandellòs-II NPP. Application to operation and control support; Nuclear Engineering and 

Design 237 (2007) 2014–2023; 2007

8. F. Reventós, L. Batet, C. Llopis, C. Pretel and I. Sol, Thermal-Hydraulic Analysis Tasks for 

ANAV NPPs in Support of Plant Operation and Control; Science and Technology of 

Nuclear Installations; Article ID 153858, 13 pages, doi:10.1155/2008/153858; 2008.

9. F. Reventós, L. Batet, C. Pretel, M. Salvat, and I. Sol, Advanced qualification process for 

ANAV integral plant models, Nuclear Engineering and Design, vol. 237, no. 1, pp. 54–63, 

2007.

10. US NRC. TRACE V5.0 THEORY MANUAL. Volume 1: Field Equations, Solution Methods, 

and modelling Techniques.

11. Westinghouse; Model F steam generator termal-hydraulic data report (WTP-ENG-TN-81-

001) 

12. Westinghouse Nuclear Española, (in Spanish) Descripción del SNGV de Westinghouse,

1983. 

13. Westinghouse Energy Systems International; Precautions, limitations and setpoints. C.N. 

Vandellòs II