NUREG/IA-0245 

 

 International 

 Agreement Report

 

Assessment of TRACE 5.0 against 

ROSA Test 6-1, Vessel Upper Head 

SBLOCA 

Prepared by: 

S. Gallardo, V. Abella, G. Verdú 

Universidad Politécnica de Valencia 

ETSII 

Camί de Vera s/n 

46021 Valencia, SPAIN 

 

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

April 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.
NUREG/IA-0245

Assessment of TRACE 5.0 against ROSA Test 6-1, 

Vessel Upper Head SBLOCA 

Manuscript Completed: June 2010 

Date Published: April 2011 

Prepared by: 

S. Gallardo, V. Abella, G. Verdú 

Universidad Politécnica de Valencia 

ETSII 

Camí de Vera s/n 

46021 Valencia, SPAIN 

Antony Calvo, NRC Project Manager 

Prepared for:

Division of Systems Analysis 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

iii

ABSTRACT

The purpose of this work is to provide an overview of the results obtained in the simulation of a 

pressure vessel upper head Small Break Loss-Of-Coolant Accident (SBLOCA) under the 

assumption of total failure of High Pressure Injection System (HPIS) in the Large Scale Test Facility 

(LSTF) via the thermal-hydraulic code TRACE5. 

The work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA). Test 

6-1 simulated a PWR pressure vessel upper-head SBLOCA with a break size equivalent to 1.9% of 

the cold leg break. The break size and core uncover caused the primary depressurization. When 

the accident management (AM) action was initiated by fully opening the steam generator relief 

valves (detection of high core exit temperature of 623 K) primary pressure was lower than the 

steam generator secondary-side pressure. A detailed model has been developed following these 

assumptions. 

Results of the simulation with TRACE5 are compared with the experimental in several graphs, 

observing an acceptable general behaviour in the entire transient. In conclusion, this work 

represents a good contribution for assessment of the predictability of computer codes such as 

TRACE5.
iv
v

FOREWORD

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

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

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

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

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

safety acceptance criteria for the licensing analysis are verified. 

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

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

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

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

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

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

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

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

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

maintaining a single, internationally recognized code version. 

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

 Share a well documented code assessment data base. 

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

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

management and operator procedures-related studies). 

 Maintain and improve user expertise and guidelines for code applications. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

community during the coming decade. 

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

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

 

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

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

Safety of Nuclear Installations (CSNI). 
vi

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

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

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

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

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

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

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

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

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

system. 

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

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

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

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

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

accidents such as: 

Temperature stratification and coolant mixing during ECCS coolant injection

 Water hammer-like phenomena 

 ATWS 

 Natural circulation with super-heated steam

 Primary cooling through SG depressurization 

 Pressure vessel upper-head and bottom break LOCA

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

period of CAMP-España activities focused on: 

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

experiments. 

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

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

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

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

studies. 

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

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

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

hydraulics analysis of accidents of the Spanish nuclear power plants. 

________________________________

Francisco Fernández Moreno, Commissioner 

Consejo de Seguridad Nuclear (CSN) 
vii

CONTENTS

Page

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

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

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

Acknowledgments ....................................................................................................................xiii

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

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

2. ROSA facility description....................................................................................................2-1

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

4. Applied method: TRACE5 model of ROSA facility............................................................4-1

5. Results and discussion.......................................................................................................5-1

5.1. System pressures ......................................................................................................... 5-3

5.2. Break ............................................................................................................................. 5-5

5.3. Primary loops mass flow rates ...................................................................................... 5-7

5.4. Vessel collapsed liquid levels ....................................................................................... 5-9

5.5. Maximum fuel rod surface temperature ...................................................................... 5-12

5.6. Maximum core exit temperature .................................................................................. 5-14

5.7. Hot and cold legs liquid levels ..................................................................................... 5-14

5.8. Steam generator relief valve flow rate ......................................................................... 5-17

5.9. U-tubes collapsed liquid level ..................................................................................... 5-18

5.10. Steam generators secondary-side liquid level ............................................................ 5-20

5.11. Pressurizer liquid level ................................................................................................5-22

5.12. Accumulator liquid level ..............................................................................................5-22

6. Conclusions .........................................................................................................................6-1

7. References............................................................................................................................ 7-1
viii

Figures

Page

Figure 1 Model nodalization used for simulation ....................................................................... 4-1

Figure 2 3D Vessel nodalization and connections visualized with SNAP ................................. 4-2

Figure 3 Steam generator nodalization ..................................................................................... 4-4

Figure 4 Primary and secondary pressures (0 to 1 NT) ............................................................ 5-4

Figure 5 Primary and secondary pressures (0 to 0.35 NT) ....................................................... 5-4

Figure 6 Break mass flow rate .................................................................................................. 5-6

Figure 7 Discharged inventory through the break ..................................................................... 5-6

Figure 8 Primary loop A mass flow (0 to 1 NT) ......................................................................... 5-7

Figure 9 Primary loop A mass flow (0 to 0.35 NT) .................................................................... 5-8

Figure 10 Primary loop B mass flow (0 to 1 NT) ....................................................................... 5-8

Figure 11 Primary loop B mass flow (0 to 0.35 NT) .................................................................. 5-9

Figure 12 Upper plenum collapsed liquid level ....................................................................... 5-10

Figure 13 Core collapsed liquid level ...................................................................................... 5-10

Figure 14 Downcomer collapsed liquid level ........................................................................... 5-11

Figure 15 Upper head collapsed liquid level ........................................................................... 5-11

Figure 16 Relation between the Upper head collapsed liquid level and break mass flow rate. 

Experimental and TRACE5 results .......................................................................................... 5-12

Figure 17 Maximum fuel rod surface temperature .................................................................. 5-13

Figure 18 Core power ............................................................................................................. 5-13

Figure 19 Maximum exit core fluid temperature ...................................................................... 5-14

Figure 20 Collapsed liquid level in the hot leg A..................................................................... 5-15

Figure 21 Collapsed liquid level in the hot leg B ..................................................................... 5-15

Figure 22 Collapsed liquid level in the cold leg A ................................................................... 5-16

Figure 23 Collapsed liquid level in the cold leg B ................................................................... 5-16

Figure 24 SG A relief valve mass flow .................................................................................... 5-17

Figure 25 SG B relief valve mass flow .................................................................................... 5-17

Figure 26 SG U-tube up-flow side collapsed liquid levels in loop with PZR ............................ 5-18

Figure 27 SG U-tube up-flow side collapsed liquid levels in loop with PZR ............................ 5-19

Figure 28 SG U-tube down-flow side collapsed liquid levels in loop with PZR ....................... 5-19

Figure 29 SG U-tube down-flow side collapsed liquid levels in loop with PZR ....................... 5-20

Figure 30 Steam generator A. Secondary-side collapsed liquid level ..................................... 5-21
ix

Figure 31 Steam generator B. Secondary-side collapsed liquid level ..................................... 5-21

Figure 32 Pressurizer liquid level ............................................................................................ 5-22

Figure 33 Accumulator tank liquid level .................................................................................. 5-23
x

Tables

Page

Table 1. Control logic and sequence of major events in the experiment ................................... 3-1

Table 2. Predetermined core power decay curve ...................................................................... 3-2

Table 3. Pumps relative rotational speed ................................................................................... 3-2

Table 4. Core protection system logic ........................................................................................ 3-3

Table 5. Number of heaters per heat structure .......................................................................... 4-3

Table 6. Steady state conditions. Comparison between experiment and TRACE5 ................... 5-1

Table 7. Chronological sequence of events. Comparison between experiment and TRACE5 .. 5-2
xi

EXECUTIVE SUMMARY

The purpose of this work is to provide an overview of the results obtained in the simulation of a 

pressure vessel upper head Small Break Loss-Of-Coolant Accident (SBLOCA) under the 

assumption of total failure of High Pressure Injection System (HPIS) in the Large Scale Test Facility 

(LSTF) via the thermal-hydraulic code TRACE5. 

The work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA). Test 

6-1 simulated a PWR pressure vessel upper-head SBLOCA with a break size equivalent to 1.9% of 

the cold leg break. The break size and core uncover caused the primary depressurization. When 

the accident management (AM) action was initiated by fully opening the steam generator relief 

valves (detection of high core exit temperature of 623 K) primary pressure was lower than the 

steam generator secondary-side pressure. A detailed model has been developed following these 

assumptions. 

Results of the simulation with TRACE5 are compared with the experimental in several graphs, 

observing an acceptable general behaviour in the entire transient. In conclusion, this work 

represents a good contribution for assessment of the predictability of computer codes such as 

TRACE5.
xii
xiii

ACKNOWLEDGEMENTS 

This paper contains findings that were produced within the OECD-NEA ROSA Project. The authors 

are grateful to the Management Board of the ROSA Project for their consent to this publication, and 

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

agreement STN/1388/05/748. 
xiv
xv

ABBREVIATIONS

AFW Auxiliary Feed Water 

AM Accident Management 

CAMP Code Assessment and Management Program 

CCFL Counter-current Flow Limiting 

CPU Central Processing Unit 

CSN Consejo de Seguridad Nuclear (Spanish nuclear regulatory commission) 

ECCS Emergency Core Cooling System 

FW Feedwater 

HPI High Pressure Injection 

JAEA Japan Atomic Energy Agency 

kg kilogram(s) 

LSTF Large Scale Test Facility 

l/s liter per second 

m meter(s) 

mm millimeter(s) 

MPa megapascal 

kg/cm2 kilogram per square centimeter 

oC degrees Celsius 

oK degrees Kelvin 

MSIV main steam isolation valve 

MW megawatt(s) 

MWe megawatt(s) electric 

MWt megawatt(s) thermal 

NPP nuclear power plant

NRC U.S. Nuclear Regulatory Commission 

PCT Peak Cladding Temperature 

PV Pressure Vessel 

PWR Pressurized Water Reactor 

PZR Pressurizer 

RV Relief Valve 

s second(s) 

SBLOCA Small Break Loss-Of-Coolant Accident 

SG Steam Generator 

SNAP Symbolic Nuclear Analysis Package

TRACE TRAC/RELAP Advanced Computational Engine 
xvi
1-1 

1 INTRODUCTION 

The Loss-Of-Coolant Accident (LOCA) is one of the most important design basis accidents. Many 

works with the goal of improving safety methodologies, protocols and safety-related operating limits 

for a plant, can be found in literature [1-8], together with works focused on testing thermalhydraulic 

codes [9-13]. In the last years, there has been a significant interest in the development of codes 

and methodologies for ―best-estimate‖ analysis of LOCAs In this frame, improvement of operating 

efficiencies in Small Break Loss-Of-Coolant Accidents (SBLOCA) is nowadays a concern. 

The aim of the present work is to describe the main results achieved by the authors using the 

thermal-hydraulic code TRACE5 [9, 10], in the frame of OECD/NEA ROSA Project Test 6-1 (SBPV-9 in JAEA) [14] with the purpose of testing the behavior of the code for this transient. The 

experiment 6-1 of the OECD/NEA ROSA (SB-PV-9 in JAEA) project was managed during 17th of 

December 2005 in the Large Scale Test Facility (LSTF) [15] of the Japanese Atomic Energy Agency 

(JAEA). The LSTF simulates a PWR reactor, Westinghouse type, of four loops and 3423 MW of 

thermal power, scaled to 1/48 in volume and two loops. 

The experiment simulates an SBLOCA in the upper head of the vessel of a Pressurize Water 

Reactor (PWR) of four loops. The size of the break is equivalent to 1.9% of the diameter of the cold 

leg. 
1-2 
2-1 

2 ROSA FACILITY DESCRIPTION 

In this section, a brief description of the LSTF facility (in the Tokai Research Establishment of the 

JAERI) is presented. The primary coolant system consists of the pressure vessel (PV), the primary 

loop A with the pressurizer (PZR) and the symmetrical primary loop B. Each loop contains a primary 

coolant pump (PC) and a steam generator (SG). The secondary-coolant system consists of the jet 

condenser (JC), the feedwater pump (PF), the auxiliary feedwater pumps (PA) and related piping 

system in addition to two SG secondary systems. The ECCSs consist of the high pressure charging 

pump (PJ), the high pressure injection pump (PL), the residual heat removal (RHR) system and the 

primary gravity injection tank (PGIT). The nitrogen gas is supplied to the accumulator tanks and 

some gas injection locations. The coolant discharged from the primary system is stored in the break 

flow storage tank (ST). The pressure vessel (PV) is composed of an upper head above the upper 

core support plate, the upper plenum between the upper core support plate and the upper core 

plate, the core, the lower plenum and the downcomer annulus region surrounding the core and 

upper plenum. LSTF vessel has 8 upper head spray nozzles (of 3.4 mm inner-diameter). 8 control 

rod guide tubes (CRGTs) form the flow path between the upper head and upper plenum. The 

maximum core power of the LSTF is limited to 10 MW which corresponds to 14% of the 

volumetrically scaled PWR core power and is sufficiently capable to simulate PWR decay heat 

power after the reactor scram. 

Regarding to the steam generators, each of them contains 141 U-tubes which can be classified in 

different groups depending on their length (an average length of 19.7 m can be considered, with a 

maximum height of 10.62 m and a minimum height of 9.156 m). U-tubes have an inner diameter of 

19.6 mm and an outer diameter of 25.4 mm (with 2.9 mm wall thickness). As a consequence, the 

total inner and outer surface areas are 171 and 222 m2, respectively. On the other hand, vessel, 

plenum and riser of steam generators have an inner height of 19.840, 1.183 and 17.827 m, 

respectively. The downcomer is 14.101 m in height. 
2-2 
3-1

3 TRANSIENT DESCRIPTION 

The control logic of the transient is listed in Table 1. The experiment was initiated by quickly 

opening the break valve (inner diameter of 13.8 mm), at time zero. Simultaneously, rotational speed 

of primary coolant pumps was increased up to 1500 rpm. A scram signal was generated when the 

pressurizer pressure dropped to a determined value. This signal produces the initiation of the core 

power decay curve, calculated by considering the stored heat in the fuel rods and the delayed 

neutron fission power, as it can be seen in Table 2. The initial core power corresponds to 14% of 

the nominal power of a PWR volumetrically scaled (1/48).

Table 1 Control logic and sequence of major events in the experiment 

Break Time zero

Generation of scram signal Primary pressure drops to a 

determined value

Pressurizer (PZR) heater off Generation of scram signal or 

PZR liquid level below a 

determined value

Initiation of core power decay curve simulation Generation of scram signal

Initiation of primary coolant pump coastdown Generation of scram signal

Turbine trip (closure of stop valve) Generation of scram signal

Closure of main steam isolation valve Generation of scram signal

Termination of main feedwater Generation of scram signal

Generation of safety injection (SI) signal Determined value of primary 

pressure

Initiation of auxiliary feedwater Generation of SI signal

Initiation of steam generator (SG) secondary-side depressurization 

as accident management (AM) action by fully opening relief valves

Core exit temperature 

reaches determined 

maximum

Initiation of accumulator system Determined value of primary 

pressure

Initiation of low pressure injection system Determined value of PV lower 

plenum pressure 
3-2

Table 2 Predetermined core power decay curve 

Normalized 

Time

Normalized

Power

Normalized 

Time

Normalized

Power

Normalized 

Time

Normalized

Power

0 1 0.0267 0.3042 0.2 0.1832

0.006 1 0.0333 0.2763 0.266 0.1577

0.0067 0.8150 0.05 0.2423 0.333 0.1487

0.01 0.5366 0.0667 0.2263 0.5 0.1342

0.0133 0.4504 0.1 0.2079 0.666 0.1238

0.0167 0.3906 0.133 0.2000 1 0.1096

0.02 0.3538 0.166 0.1913

At the same time, the primary coolant pump coastdown is initiated, also using a pre-determined 

rotational speed curve (Table 3).

Table 3 Pumps relative rotational speed 

Normalized 

Time

Relative

rotational 

speed

Normalized 

Time

Relative

rotational 

speed

Normalized 

Time

Relative

rotational 

speed

0 1.000 0.01 0.280 0.026 0.125

0.00066 0.850 0.013 0.220 0.03 0.110

0.0017 0.730 0.016 0.185 0.033 0.100

0.0033 0.540 0.02 0.160 0.083 0.000

0.0067 0.370 0.023 0.140 -- ---

A turbine trip is activated by closing the steam generators main steam isolation valves (MSIVs). The 

closure of MSIVs produces an increasing of SG secondary-side pressure and a temporary rise in 

the primary pressure, followed by a new decrease due to the core power decay effect. 

Simultaneously, main feedwater flow of both SGs is stopped. The safety injection signal (SI) is 

generated when the primary pressure decreases to a determined value. From this moment on, the 

relief valves (RV) in both steam generators, begin opening and closing in order to maintain the 

pressure between two fixed values. When the core exit temperature reaches a determined value, 

the accident management (AM) action is initiated by fully opening relief valves of both steam 
3-3

generators. The core power is automatically decreased by the core protection system when the 

maximum fuel rod surface temperature exceeds a certain maximum, as it can be seen in Table 4. 

Table 4 Core protection system logic 

Control of

core power to

Maximum fuel rod

surface temperature (K)

75% 958

50% 968

25% 969

10% 970

0% 973

When primary pressure drops to a determined value, the accumulation system starts to inject water 

in cold legs. In the last part of the experiment, nitrogen gas is injected in both cold legs, throughout 

the accumulators. Finally, when primary pressure drops to a determined value, the Low Pressure 

Injection System (LPIS) is initiated. 
3-4
4-1 

4 APPLIED METHOD: TRACE5 MODEL OF ROSA FACILITY 

TRACE5 code is designed to perform best-estimate analyses of loss-of-coolant accidents (LOCAs) 

or operational transients. However, it has some limitations in use, for example, model cannot be 

applied directly to those transients in which one expects to observe thermal stratification of the 

liquid phase in the 1D components. Furthermore, it is not appropriate for modeling situations in 

which transfer of momentum plays an important role at a localized level. In this work, the LSTF has 

been modeled with 88 hydraulic components (7 BREAKs, 13 FILLs, 29 PIPEs, 2 PUMPs, 1 

PRIZER, 21 TEEs, 14 VALVEs and 1 VESSEL). In order to characterize the heat transfer 

processes, 48 Heat Structure components (Steam Generator U-tubes, core power, pressurizer 

heaters and heat losses) have been considered. Figure 1 shows the nodalization of the model using 

SNAP (Symbolic Nuclear Analysis Package software) [16].

Figure 1 Model nodalization used for simulation 
4-2 

In order to model the pressure vessel, a 3D–VESSEL component has been considered (Figure 2). 

A nodalization consisting of 19 axial levels, 4 radial rings and 10 azimuthal sectors has been 

selected. This nodalization characterizes with an acceptable detail the actual features of the LSTF 

vessel. Increasing the number of axial levels, azimuthal sectors or radial rings, does not improve 

significantly the agreement with experimental results, but increases CPU time. For each axial level, 

volume and effective flow area fractions have been set according to technical specifications 

provided by the organization [14]. Active core is located between levels 3 and 11. Level 12 

simulates the upper core plate. Levels 13 to 15 characterize the vessel upper plenum. In level 16, 

the upper core support plate is located. Finally, upper head is defined between levels 17 to 19. 3DVESSEL is connected to different 1D components: 8 Control Rod Guide Tubes (CRGT), hot leg A 

and B (level 15), cold leg A and B (level 15) and a bypass channel (level 14). Control rod guide 

tubes have been simulated by PIPEs components, connecting levels 13 and 19 and allowing the 

flow between upper head and upper plenum. 

Figure 2 3D Vessel nodalization and connections visualized with SNAP 

30 HTSTRs simulate the fuel assemblies in the active core. A POWER component manages the 

power supplied by each HTSTR to the 3D-VESSEL. Fuel elements (1008 in total) were distributed 

into the 3 rings: 154 elements in ring 1, 356 in ring 2 and 498 in ring 3 and also characterized by 

HTSTR components. In both axial and radial direction, peaking factors were considered. The power 

ratio in the axial direction presents a peaking factor of 1.495. On the other hand, depending on the 

radial ring, different peaking factors were considered (0.66 in ring 1, 1.51 in ring 2 and 1.0 in ring 3). 

The number of fuel rod components associated with each heat structure has been determined from 

the technical documentation given, taking into account the distribution of fuel rod elements in the 

vessel, as it can be seen in Table 5. 

A detailed model of SG (geometry and thermal features) has been developed, due to the fact that 

TRACE5 does not include any pre-determined steam generator component. A representation of the 
4-3 

SG nodalization can be seen in Figure 3. Both boiler and downcomer components of secondaryside, have been modelled by TEEs components. U-tubes have been classified into three groups 

according to each average length and heat transfer features. Steam-separator model can be 

invoked in TRACE5 setting a friction coefficient (FRIC) greater than 1022 at a determined cell edge, 

allowing only to flow through the cell interface gas phase. Heat transfer between primary and 

secondary sides has been performed by using HTSTR components. Cylindrical-shape geometry 

has been used to best fit heat transmission. Critical heat flux flag has been set in order to use an 

AECL-IPPE Table, calculating critical quality from Biasi correlation [10]. Inner and outer surface 

boundary conditions for each axial level, has been set to couple HTSTR component to hydro 

components (primary and secondary fluids). Different models varying the number of U-tube groups 

were tested (1, 3 and 6 groups). It was found that results do not apparently change, using these 

models. However, in order to best fit the collapsed liquid level in U-tubes without drastically 

increasing CPU time, a 3-group configuration was finally chosen. Heat losses to environment have 

been added to secondary-side walls. 

Table 5 Number of heaters per heat structure 

HTSTR Number of 

heaters HTSTR Number of heaters HTSTR Number of heaters

310 17 320 44 330 60

311 17 321 40 331 54

312 10 322 23 332 32

313 12 323 32 333 45

314 20 324 40 334 56

315 17 325 42 335 61

316 16 326 38 336 57

317 12 327 26 337 31

318 14 328 30 338 45

319 17 329 39 339 57
4-4 

Figure 3 Steam generator nodalization 

Regarding to the break simulation, it is important to take into account the necessity of activating the 

Choke flow model in the break when critical flow conditions are expected to appear. Choke model 

predicts for a given cell the conditions for which choked flow is expected to occur, providing three 

different models in one: subcooled-liquid, two-phase and single-phase vapor model. The break has 

been simulated by means of a VALVE component connected to a BREAK component in order to 

establish the boundary conditions. This BREAK has been modelled following the recommendations 

of the TRACE5 user’s manual [9]. In this case, since the break is simulated to discharge in a big 

volume space (the storage tank), a dxin=1.0*10-6 (length) and a volin=1.0*106 (volume) has been 

selected with the purpose of providing a large area. 
5-1 

5 RESULTS AND DISCUSSION 

Steady-state conditions achieved in the simulation were in reasonable agreement with the 

experimental values, as it can be seen in Table 6. 

Table 6 Steady state conditions. Comparison between experiment and TRACE5 

Item Relative Error (%)

(loop without PZR)

Pressure vessel

Core power (MW) 0.00

Primary loop

Hot leg fluid temperature (K) 0.20

Cold leg fluid temperature (K) 0.07

Mass flow rate (kg/s / loop) 2.80

―downcomer‖-to-hot leg bypass 

(kg/s) 2.00

Pressurizer

Pressure (MPa) 0.50

Liquid level (m) 4.10

Accumulator system

Pressure (MPa) 0.20

Temperature (K) 0.50

Steam generators

Secondary-side pressure (MPa) 0.70

Secondary-side liquid level (m) 5.90

Steam flow rate (kg/s) 4.10

Main feedwater flow rate (kg/s) 0.70

Main feedwater temperature (K) 0.10

Auxiliary feedwater temp (K) 0.06
5-2 

Table 7 lists the chronology sequence of events during the transient and the comparison in 

Normalized Time between the experiment and TRACE5 results. 

Table 7 Chronological sequence of events. Comparison between experiment and 

TRACE5 

Event

Experiment 

Normalized 

Time

TRACE5 

Normalized 

Time

Valve open 0.0 0.0

Scram signal (determined value of primary pressure) 0.007 0.006

S.I. signal (determined value of primary pressure) 0.008 0.01

Phase change (one phase liquid to two-phase) in the 

break 0.01 0.02

Pumps stop 0.09 0.09

Phase change (two-phase to one phase vapor) in the 

break 0.24 0.23

Primary pressure lower than in the secondary 0.27 0.25

Initiation of secondary depressurization (determined 

exit temperature of the ―upper plenum‖) 0.37 0.37

Initiation of core protection system (determined vessel 

temperature) 0.41 0.43

Accumulators start (determined value of primary 

pressure) 0.44 0.44

Initiation of LPIs (determined value of lower plenum 

pressure) 1 0.93

Variables presented in this section follow the requirements for an exhaustive analysis of the 

transient. The most important parameters that will be studied in this paper are the following: 

Pressures at both primary and secondary circuits, mass flow rate and inventory at the break, 

primary mass flow, vessel collapsed-liquid levels, maximum fuel rod surface temperature, core exit 

temperature, collapsed-liquid levels in hot and cold legs, mass flow in SG relief valves, liquid level 

in SG secondary-side and liquid level in the accumulators. 
5-3 

5.1 System pressures

Figures 4 and 5 compare the primary and secondary pressures. The primary pressure begins to 

decrease at time zero (when the break is produced). In the experiment, the scram signal is 

generated at 0.007 NT after the break, when the primary pressure decreases to a determined value. 

The generation of the scram signal causes the main steam valve (MSIV) of steam generators to 

close and the beginning of the primary coolant pumps coastdown. The SI signal is generated at 

0.008 NT when the primary pressure decreases to the determined value. The secondary pressure 

rapidly increases after the closure of the MSIVs. From this moment on, the secondary pressure 

starts to oscillate by means of opening and closing the relief valves (RV) of steam generators until 

0.37 NT, when the depressurization of steam generators secondary-side begins. 

The relatively big size of the break results in a rapid depressurization of the primary side, especially 

until 0.24 NT when the break flow changes to single-phase steam. The pressure in the primary 

decreases below the pressure of the secondary in the steam generator at 0.27 NT, almost 

simultaneously to the core uncover. Anyway, according to the experiment report [14], the 

depressurization of the secondary is not effective until 0.27 NT, when the pressure of the primary 

becomes lower than the pressure of the secondary. 

The first part of the transient (between 0 and 0.24 NT) shows a perfect agreement between 

experimental and simulated values. Discrepancies appear in the second part of transient, when 

two-phase break flow turns to single-phase steam flow (Figures 4 and 5). Deviation between both 

primary pressures might be due to an improper treatment of the steam critical flow at the break by 

TRACE5, as it will be explained in the following section. The accumulator system is initiated at 

about 0.44 NT when the primary pressure decreases to a determined value, producing an 

increasing in the primary pressure of experimental values. The entrance of water coming from the 

accumulators produces a peak of pressure, also coincident with the refill of the pressure vessel. 

This effect is not registered in the simulation. The coolant injection from the accumulator system is 

finished when the primary pressure decreases to the determined value, followed by the discharge of 

nitrogen gas from the accumulator tanks. 

TRACE5 adequately reproduces the liquid level decrease of accumulators (as it will be shown in 

other section) and the entrance of nitrogen gas. Once accumulators are empty of water, both 

primary pressures of experimental and TRACE5 become almost the same. When vessel lower 

plenum pressure reaches the determined value, the low pressure injection (LPI) system actuates at 

about 1 NT in the experiment, in agreement with TRACE5, which registers this event slightly before. 
5-4 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Values

Primary and Secondary Pressures

Measured Primary Pressure

Measured Secondary Pressure SGA

Measured Secondary Pressure SGB

TRACE5 Primary Pressure

TRACE5 Secondary Pressure SGA

TRACE5 Secondary Pressure SGB

Figure 4 Primary and secondary pressures (0 to 1 NT) 

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Values

Primary and Secondary Pressures

Measured Primary Pressure

Measured Secondary Pressure SGA

Measured Secondary Pressure SGB

TRACE5 Primary Pressure

TRACE5 Secondary Pressure SGA

TRACE5 Secondary Pressure SGB

Figure 5 Primary and secondary pressures (0 to 0.35 NT) 
5-5 

5.2 Break

Break flow rate is shown in Figure 6. Figure 7 shows the discharged inventory. The primary 

pressure and the inventory decrease sharply until the saturation pressure of the coolant is reached. 

Then, the pressure slowly decreases as the break flow changes to saturated two-phase flow. 

TRACE5 adequately reproduces the flow through the break, and so the discharged inventory in 

both cases (single and two phases). From 0.25 NT on, the upper plenum begins to empty and so 

the two-phase flow through the guide tubes connecting the upper plenum and the upper head ends. 

Thereafter, the code overpredictes the break mass flow. This fact can be seen in Figure 7 for time 

up to 0.24 NT. TRACE5 allows to apply Choke flow [9] conditions and use discharge coefficients 

(one liquid single-phase and two-phase coefficients). 

Between 0.25 and 0.4 NT TRACE predicts single-phase vapour, but during this period the total 

mass flow rate calculated through the break is slightly higher than the experimental measurement. 

In other time intervals (for example between 0.65 and 0.75 NT) a two-phase liquid-vapor mass flow 

is leaving through the break, but before this time period, the total discharged inventory was higher 

than the experimental measurement. This permits to conclude that a single vapour phase discharge 

coefficient is needed in the simulation. Other codes, such as RELAP, include the possibility to have 

into account this discharge coefficient for one-phase vapour. 

On the other hand, to best simulate the break mass flow rate it is necessary to adequately model all 

the flowpaths reaching the break. In the simulation three flowpaths have been studied: control rod 

guide tubes connecting the upper head and the upper plenum, the spray nozzles between the 

upper plenum and the downcomer and the bypass between the upper plenum and the downcomer 

through the hot leg. The bypass between the upper plenum and the downcomer has been modeled 

by means of a TEE component. Friction and flow are of bypass has been adjusted during the 

steady state calculations in order to achieve a mass flow rate through the bypass similar to the 

obtained in the experiment. NFF factors have been set to -1 in the secondary side of this TEE 

component. Regarding to the control rod guide tubes, results that better fits experimental 

measurements are obtained adjusting the flow are of the control rod guide tubes to the actual flow 

area and considering NFF=-1. Finally, spray nozzles between upper plenum and upper head have 

been modeled taking into account the technical specifications provided by JAEA. 
5-6 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

-0.2

0

0.2

0.4

0.6

0.8

1

Normalized Time

Normalized Values

Discharged primary coolant mass

Measured discharged primary coolant mass

TRACE5

Figure 6 Break mass flow rate 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.2

0.4

0.6

0.8

1

1.2

Normalized Time

Normalized Values

Break Flow Rate

Measured break flow rate

TRACE flow rate

Figure 7 Discharged inventory through the break 
5-7 

5.3 Primary loops mass flow rates

In general, mass flow through both loops is adequately reproduced, as it can be seen in the 

following Figures. Figures 8 and 9 show the mass flow of loop A (measured in TRACE5 at pump 

position). In addition, Figures 10 and 11 show the loop B mass flow. Almost symmetric flow rates 

were observed among both loops during the entire transient. According to these results, TRACE5 

successfully reproduces the natural flow circulation in primary loops under conditions of upper 

plenum SBLOCA. 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

-0.2

0

0.2

0.4

0.6

0.8

Normalized Time

Normalized Values

Primary loop A mass flow rates

Primary Loop A masflow

TRACE5 A

Figure 8 Primary loop A mass flow (0 to 1 NT) 
5-8 

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

-0.2

0

0.2

0.4

0.6

0.8

Normalized Time

Normalized Values

Primary loop A mass flow rates

Primary Loop A masflow

TRACE5 A

Figure 9 Primary loop A mass flow (0 to 0.35 NT) 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

-0.2

0

0.2

0.4

0.6

0.8

1

1.2

Normalized Time

Normalized Values

Primary loop B mass flow rates

Primary LoopB massflow

TRACE5 B

Figure 10 Primary loop B mass flow (0 to 1 NT) 
5-9 

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

-0.2

0

0.2

0.4

0.6

0.8

Normalized Time

Normalized Values

Primary loop B mass flow rates

Primary LoopB massflow

TRACE5 B

Figure 11 Primary loop B mass flow (0 to 0.35 NT) 

5.4 Vessel collapsed liquid levels

The following Figures (12, 13 and 14) show a comparison between the collapsed liquid levels in the 

upper plenum, core and downcomer, respectively for both experimental and TRACE5 results. In the 

experiment, the collapsed liquid level is computed from differences in pressure between the upper 

and lower parts of each region, and the coolant densities. The collapsed water level decreased due 

to a loss of the RCS coolant inventory through the break. Coolant in the upper plenum entered the 

upper-head through control rod guide tubes (CRGTs) until the entrance holes at the CRGTs bottom 

are exposed to steam. In Figure 12 it can be seen that the experimental upper plenum liquid level 

remains almost constant until 0.24 NT. TRACE5 reproduces this phenomenon delaying the 

entrance of steam approximately 0.03 NT (to 0.27 NT). Immediately after, core liquid level starts to 

decrease, producing the core uncover until 0.44 NT (Figure 13). Simultaneously with the core 

empty, the downcomer liquid level decreases, until the entrance of water coming from accumulators 

(0.48 NT, approximately) begins (Figure 14). In the experiment, the core and downcomer liquid 

levels indicate temporal manometric fluctuations during the coolant injection from the accumulator 

since the local pressure decreases in the cold legs. The liquid level decreases in the core and 

increases in the downcomer. This phenomenon is not reproduced with TRACE5. 
5-10

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.05

0.1

0.15

0.2

0.25

Normalized time

Normalized values

PV collapsed liquid levels

Measured Upper Plenum Collapsed level

TRACE5

Figure 12 Upper plenum collapsed liquid level

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

Normalized time

Normalized values

PV collapsed liquid levels

Measured Core Collapsed level

TRACE5

Figure 13 Core collapsed liquid level 

The most important discrepancy is observed in the last stretch of the transient (between 0.68 and 1 

NT) in the downcomer liquid level. In the model, a decrease in the liquid level occurs, which is not 

registered in the experiment. This phenomenon has a straight relationship with the mass 

discharged inventory. 
5-11

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized time

Normalized values

PV collapsed liquid levels

Measured Downcomer Collapsed level

TRACE5

Figure 14 Downcomer collapsed liquid level 

Upper head liquid level is shown in Figure 15. A good agreement can be seen between 

experimental and TRACE5 results. It is important to remark that mass flow rate is conditioned by 

the upper head liquid level as it can be observed in Figure 16. 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.2

0.4

0.6

0.8

1

1.2

Normalized Time

Normalized Values

Liquid level

Upper Head collapse liquid level

TRACE

Figure 15 Upper head collapsed liquid level 
5-12

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.5

1

Time

Experimental break flow and UH collapsed level 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.5

1

experimental break flow rate

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.5

1

Time (s)

TRACE break flow and UH collapsed level

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

1

TRACE break flow rate

Figure 16 Relation between the Upper head collapsed liquid level and break mass 

flow rate. Experimental and TRACE5 results 

5.5 Maximum fuel rod surface temperature

TRACE5 reproduces precisely the evolution of the maximum fuel rod temperature in the core. The 

maximum temperature is reached at 0.41 NT. At this moment, the core protection system activation 

is produced, reducing the core power according to a programmed decay power curve (Figure 18 

and Table 4). After the activation of this system and the entrance of water from accumulators, 

TRACE5 fuel rod surface temperature is in good agreement with experimental values (Figure 17). 
5-13

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Temperature (K)

Fuel rod surface T

Fuel rod surf T

Fuel rod surf T

Fuel rod surf T

Fuel rod surf T

Fuel rod surf T

TRACE5

Figure 17 Maximum fuel rod surface temperature 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Values

Core power

total core power

TRACE5 A

Figure 18 Core power 
5-14

5.6 Maximum core exit temperature

Figure 19 shows a comparison between TRACE5 and the experimental core-exit fluid temperature. 

Experimentally, core exit temperatures are observed to keep the saturation temperature until 0.31, 

even after the beginning of the core uncover, partly due to the condensation produced in the hot 

leg. It is also shown that core exit temperatures depend on the radial position. In the experiment, 

the temperature peak occurs in the centre, while the exit temperatures of the middle ring are almost 

the same to those of the outer ring. The accident management action begins at 0.37 NT 

(experimentally and with TRACE5) immediately after the core exit temperature reaches a maximum. 

In TRACE5, thermocouples have been located in the level 13 of the vessel (in the upper plenum 

region). 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0.55

0.6

0.65

0.7

0.75

0.8

0.85

0.9

0.95

1

Normalized Time 

Normalized Value

Core Exit Temperature

Measured Core Exit T

TRACE5 Core Exit T

Figure 19 Maximum exit core fluid temperature 

5.7 Hot and cold legs liquid levels

The following Figures (20, 21, 22 and 23) show the liquid level in hot and cold legs, respectively. 

Experimentally, liquid level was obtained with a three gamma ray beam densitometer. The hot leg 

liquid level is kept constant in a value until 0.24 NT, suddenly decreasing from this time on. Hot legs 

remain empty until the entrance of the liquid from the accumulators. Experimentally, this 

phenomenon takes place at 0.55 NT. Regarding the collapsed liquid level in cold legs, it is 

important to remark the good agreement between TRACE5 and the experiment, until the entrance 

of accumulators (0.44 NT, approximately). 
5-15

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Values

Hot leg A liquid level

Hot leg A liquid level

TRACE5 A

Figure 20 Collapsed liquid level in the hot leg A 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Values

Hot leg Bliquid level

Hot leg B liquid level

TRACE5 B

Figure 21 Collapsed liquid level in the hot leg B 
5-16

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Values

Cold leg A liquid level

Cold leg A liquid level

TRACE5 A

Figure 22 Collapsed liquid level in the cold leg A

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Values

Cold leg B liquid level

Cold leg B liquid level

TRACE5 B

Figure 23 Collapsed liquid level in the cold leg B 
5-17

5.8 Steam Generator relief valve flow rate

A good agreement has been achieved between TRACE5 and the experiment, as it can be observed 

in Figures 24 and 25. These Figures perfectly show the periods corresponding to the relief valves 

actuation and the continuous depressurization of secondary-sides. 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.2

0.4

0.6

0.8

1

Normalized Time

Normalized Values

mass flow-reliefvalveA

TRACE

Figure 24 SG A relief valve mass flow 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.2

0.4

0.6

0.8

1

Normalized Time

Normalized Values

mass flow-reliefvalveB

TRACE

Figure 25 SG B relief valve mass flow 
5-18

5.9 U-tubes collapsed liquid level

U-tubes have been grouped according to similar lengths. Experimentally six different types were 

considered. In the TRACE5 simulation and due to the calculation time cost together with the 

obtained results, a 3-group classification has been adopted. Collapsed liquid levels obtained with 

TRACE5 are satisfactory, properly reproducing the clearance of the tubes. In both experimental and 

TRACE5 results, collapsed liquid levels in up-flow and down-flow sides are balanced. The natural 

two-phase circulation ends about 0.17 NT before the U-tubes lose the collapsed liquid level in the 

up-flow side. Results are shown in Figures 26, 27, 28 and 29. 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

-0.1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

Normalized Time

Normalized Values

SG U-tube upflow-side collapsed liquid levels in loop with PZR

SGA Tube1 Inlet

SGA Tube2 Inlet

SGA Tube3 Inlet

SGA Tube4 Inlet

SGA Tube5 Inlet

SGA Tube6 Inlet

TRACE51

TRACE52

TRACE53

Figure 26 SG U-tube up-flow side collapsed liquid levels in loop with PZR 
5-19

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

-0.1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

Normalized Time

Normalized Values

SG U-tube upflow-side collapsed liquid levels in loop with PZR

SGA Tube1 Inlet

SGA Tube2 Inlet

SGA Tube3 Inlet

SGA Tube4 Inlet

SGA Tube5 Inlet

SGA Tube6 Inlet

TRACE51

TRACE52

TRACE53

Figure 27 SG U-tube up-flow side collapsed liquid levels in loop with PZR

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

-0.1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

Normalized Time

Normalized Values

SG U-tube downflow-side collapsed liquid levels in loop with PZR

SGA Tube1 Outlet

SGA Tube2 Outlet

SGA Tube3 Outlet

SGA Tube4 Outlet

SGA Tube5 Outlet

SGA Tube6 Outlet

TRACE51

TRACE52

TRACE53

Figure 28 SG U-tube down-flow side collapsed liquid levels in loop with PZR 
5-20

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35

-0.1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

Normalized Time

Normalized Values

SG U-tube downflow-side collapsed liquid levels in loop with PZR

SGA Tube1 Outlet

SGA Tube2 Outlet

SGA Tube3 Outlet

SGA Tube4 Outlet

SGA Tube5 Outlet

SGA Tube6 Outlet

TRACE51

TRACE52

TRACE53

Figure 29 SG U-tube down-flow side collapsed liquid levels in loop with PZR 

5.10 Steam generators secondary-side liquid level

The following Figures (30 and 31) show the collapsed liquid level of the secondary side of the 

steam generator. Experimentally, the liquid level is kept above a determined value so that the Utubes are covered. The liquid level starts to decrease immediately after the initiation of the AM 

action. The most relevant discrepancy between the experimental values and those obtained with 

the model, is that with TRACE5 the sharp drop of level registered in the experiment in each relief 

valve actuation has not been reproduced. 
5-21

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0.65

0.7

0.75

0.8

0.85

0.9

0.95

1

Normalized Time

Normalized Values

SGA secondary-side liquid levels

SGA wide range

TRACE5 A

Figure 30 Steam generator A. Secondary-side collapsed liquid level

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0.65

0.7

0.75

0.8

0.85

0.9

0.95

1

Normalized Time

Normalized Values

SGB secondary-side liquid levels

SGB wide range

TRACE5 B

Figure 31 Steam generator B. Secondary-side collapsed liquid level
5-22

5.11 Pressurizer liquid level

Figure 32 shows the pressurizer water level. Liquid level starts to decrease immediately after the 

break. Experimentally, it becomes completely empty at 0.01 NT. TRACE5 reproduces the 

experiment adequately, but during the period between 0.01 and 0.03 NT, it predicts a liquid level 

slightly higher than the experimental. 

0 0.05 0.1 0.15

-0.1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

Normalized Time

Normalized Liquid level

PZR collapsed liquid level

Measured PZR liquid level

TRACE5 PZR liquid level

Figure 32 Pressurizer liquid level 

5.12 Accumulator liquid level

Figure 33 shows the accumulator tank of loop A collapsed liquid level. In this case, there is no 

experimental data available. Accumulators start to empty at 0.43 NT, a little bit after the maximum 

fuel surface temperature is reached. 
5-23

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

Normalized Time

Normalized Liquid level

ACC collapsed liquid level

TRACE5 ACC liquid level

Figure 33 Accumulator tank liquid level 
5-24
6-1 

6 CONCLUSIONS 

This paper contains results obtained in the simulation of the OECD/NEA ROSA Project Test 6-1 

with the code TRACE5. Experimental results show that liquid level in the upper head controls the 

break flow rate. Phase change from two-phases to one single-phase steam is produced when 

coolant flow through control rod guide tubes is over. TRACE5 corroborates this fact. 

Depressurization produced by the break causes the primary pressure to become lower than SG 

secondary-side pressure at 0.27 NT. At this moment steam generators heat extraction starts to be 

ineffective. Approximately at 0.37 NT the accident management (AM) action is initiated, consisting 

in the steam generators secondary-side depressurization (activated when core fluid temperature 

exceeds a maximum temperature). This action appears to be insufficient to avoid the fuel surface 

rod temperature excursion (at 0.41 NT). Core protection system decreases the core power 

automatically at 0.41 NT. TRACE5 successfully reproduces all these events but it is necessary to 

take into account a limitation that needs to be improved in the future: the lack of a single-phase 

steam discharge coefficient. Comparison between experimental and TRACE5 results shows that 

discharged mass inventory through the break predicted by TRACE5 is well reproduced when fluid in 

upper head is one single liquid phase or two-phase in critical flow conditions. However, when fluid 

turns into steam single-phase, discharge mass flow is approximately 10% higher than the 

experiment. This overestimation of TRACE5 causes a lower primary pressure than the 

experimental. At 0.44 NT, water coming from accumulator tanks is injected in the cold leg of both 

loops. Upper plenum, core and downcomer refill is well reproduced by TRACE5. The main 

discrepancy after that is the liquid level decrease observed in the downcomer of the pressure 

vessel. 

Finally, TRACE5 produces the low injection system activation at 0.93 NT, approximately, when 

primary pressure drops to a determined value, in good agreement with the experiment. 

Variable Meaning Code File

n_var

Total stored data: total number of 

the control variables that will move 

the control rod banks

PARCS v2.7 pdmr_varM.f

var_leidas(1000)

It stores the read data, that is, the 

ID’s fo the RELAP5 control 

variables which will move the 

control rod banks in PARCS v2.7.

PARCS v2.7 pdmr_varM.f
6-2 
6-3 
7-1 

7 REFERENCES 

1. Stephen M. Bajorek, Nikolay Petkov, Katsuhiro Ohkawa, Arthur P. Ginsberg. ―Realistic 

Small And Intermediate-Break Loss of Coolant Accident Analysis Using WCOBRA/TRAC‖. 

Nuclear Technology Vol. 136. Oct. 2001. 

2. Steven T. Polkinghorne, Cliff B. Davis, Richard T. McCracken. ―Analysis of an Advanced 

Test Reactor Small Break Loss of Coolant Accident With An Engineered Safety Feature To 

Automatically Trip The Primary Coolant Pumps‖. Nuclear Technology Vol. 132. Oct. 2000. 

3. Kyoo Hwan Bae, Guy Hyung Lee, Hee Cheol Kim, Quun S. Zee. ―SBLOCA Long Term 

Cooling Procedure for the Integral Type PWR‖. Annals of Nuclear Energy 34 (2007) 333–

338. March 2007. 

4. Afshin Heyadat, Hadi Davilu, Jalil Jafari. ―Loss of Coolant Accident Analyses on Tehran 

Research Reactor by RELAP5/MOD3.2 code‖. Progress in Nuclear Energy 49 (2007) 

511e528. 2007.

5. Y. Koizumi, H. Asaka, H. Kumamaru, M. Osakabe, K. Tasaka, Y. Mimura, ―Investigation of 

Break Orientation Effect During Cold Leg Small Break LOCA at ROSA-IV LSTF,‖ J. Nucl. 

Sci. Technol. 25 1988. 

6. Mitsuhiro Suzuki, Takesi Takeda, Hideaki Asaka, Hideo Nakamura. ―Effects of Secondary 

Depressurization on Core Cooling in PWR Vessel Bottom Small Break LOCA Experiments 

with HPI Failure and Gas Inflow‖. Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 

43, No. 1, p. 55–64. 2006.

7. Hiroshige Kumamaru, Yutaka Kukita, Hideaki Asaka. ―RELAP5/MOD3 Code Analyses of 

LSTF Experiments on Intentional Primary-Side Depressurization Following SBLOCAS with 

Totally Failed HPI. NUCLEAR TECHNOLOGY VOL. 126. June 1999. 

8. Chien-Hsiung Lee, I-Ming Huang, Chin-Jang Chang. ―Using An IIST 1% Cold-Leg SBLOCA 

Experiment With Passive Safety Injection To Assess The RELAP5/MOD3.2 Code‖. Nuclear 

Technology Vol. 135. Aug. 2001. 

9. Nuclear Regulatory Commission, 2007a. TRACE5 V5.0. User’s manual. Volume 1: Input 

Specification. Division of Risk Assessment and Special Projects. Office of Nuclear 

Regulatory Research. U. S. 

10. Nuclear Regulatory Commission, 2007b. TRACE5 V5.0. Theory manual. Field Equations, 

Solution Methods and Physical Models. Division of Risk Assessment and Special Projects. 

Office of Nuclear Regulatory Research. U. S. 

11. RELAP5/MOD3.3 code manual. Volume II: User’s guide and input requirements. December 

2001. U.S. Nuclear Regulatory Commission. 
7-2 

12. Wolfert, K., Teschendorff, V., Lerchl, G., et. al., 1989. The thermal–Hydraulic Code ATHLET 

for Analysis of PWR and BWR Systems, NURETH-4. Karlsruhe 1989. Proc. vol. II, 1234–

1239, 1989

13. Micaëlli, J.C., Barré, F., Bestion, D., CATHARE Code Development and Assessment 

Methodologies. Trans. of the ANS, Winter Meeting San Francisco, October 29–November 2, 

vol. 73, 509–510, 1995.

14. Thermohydraulic Safety Research Group, Nuclear Safety Research Center, Japan Atomic 

Energy Agency, 2006. Final Data Report of ROSA/LSTF Test6-1 (1.9% Pressure Vessel 

Uppe-head Small Break LOCA Experiment SB-PV-09 in JAEA). 

15. The ROSA-V Group, 2003. JAERI-Tech. ROSA-V Large Scale Test Facility (LSTF) system 

description for the third and fourth simulated fuel. assemblies. 

16. Nuclear Regulatory Comission and Applied Programming Technology, 2007. Symbolic 

Nuclear Analysis Package (SNAP).