NUREG/IA-0244 

 International 

 Agreement Report

 

Assessment of TRACE 5.0 against 

ROSA Test 6-2, Vessel Lower Plenum 

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 

February 2011 

 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

Under the International Code Assessment and Maintenance Program (CAMP) 

 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0244

 

Assessment of TRACE 5.0 against ROSA Test 6-2,

Vessel Lower Plenum SBLOCA. 

 

 

 

Draft Report for Comment 

 

Manuscript Completed: June 2010

Date Published: February 2011

Prepared by:

S. Gallardo, V. Abella, G. Verdú

Universidad Politécnica de Valencia

ETSII

Camí de Vera s/n

46021 Valencia, SPAIN

A. 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 lower plenum 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 6-2 (SB-PV-10 in JAEA). An 

asymmetrical steam generator secondary-side depressurization is produced as an accident 

management action at the steam generator in the loop without PZR, after the generation of the 

safety injection signal in order to achieve a determined depressurization rate in the primary system. 

A detailed model has been developed with TRACE5 following these assumptions. 

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

acceptable general behavior in the entire transient. In conclusion, this work represents a small 

contribution for assessment of the predictability of thermal hydraulic 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

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

Page

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

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

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

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

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

2. ROSA LSTF description ......................................................................................................2-1

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

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

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

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

5.2. Transient .......................................................................................................................5-2

5.3. System pressures .........................................................................................................5-4

5.4. Break.............................................................................................................................5-6

5.5. Primary loops mass flow ...............................................................................................5-7

5.6. Vessel collapsed liquid levels .....................................................................................5-10

5.7. Peak cladding temperature .........................................................................................5-13

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

5.9. Steam generator main steam mass flow rate..............................................................5-18

5.10. Steam generator relief valve flow rate.........................................................................5-20

5.11. Steam generator u-tubes collapsed liquid level ..........................................................5-22

5.12. Steam generators secondary-side liquid level ............................................................5-26

5.13. Pressurizer liquid level ................................................................................................5-28

5.14. Accumulator liquid level ..............................................................................................5-29

5.15. Hot leg temperature ....................................................................................................5-30

5.16. Downcomer fluid temperature.....................................................................................5-31

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

8. 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.....................................................................................4-3

Figure 4 Primary and secondary pressures (0.0 to 1.0 NT)......................................................5-5

Figure 5 Primary and secondary pressures (0.0 to 0.25 NT)....................................................5-5

Figure 6 Break mass flow rate (0.0 to 1.0 NT)..........................................................................5-6

Figure 7 Discharged inventory through the break (0.0 to 1.0 NT).............................................5-7

Figure 8 Primary mass flow, loop A (0.0 to 1.0 NT)..................................................................5-8

Figure 9 Primary mass flow, loop A (0.0 to 0.2 NT)..................................................................5-8

Figure 10 Primary mass flow, loop B (0.0 to 1.0 NT)................................................................5-9

Figure 11 Primary mass flow, loop B (0.0 to 0.4 NT)................................................................5-9

Figure 12 Upper plenum collapsed liquid level (0.0 to 1.0 NT)...............................................5-10

Figure 13 Upper plenum collapsed liquid level (0.0 to 0.5 NT)...............................................5-11

Figure 14 Core collapsed liquid level (0.0 to 1.0 NT)..............................................................5-11

Figure 15 Downcomer collapsed liquid level (0.0 to 1.0 NT) ..................................................5-12

Figure 16 Upper head collapsed liquid level ...........................................................................5-12

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

Figure 18 Core power (0.0 to 1.0 NT).....................................................................................5-14

Figure 19 Collapsed liquid level in hot leg A (0.0 to 1.0 NT)...................................................5-15

Figure 20 Collapsed liquid level in hot leg A (0.0 to 0.4 NT)...................................................5-15

Figure 21 Collapsed liquid level in hot leg B (0.0 to 1.0 NT)...................................................5-16

Figure 22 Collapsed liquid level in hot leg B (0.0 to 0.4 NT)...................................................5-16

Figure 23 Collapsed liquid level in cold leg A (0.0 to 1.0 NT) .................................................5-17

Figure 24 Collapsed liquid level in the cold leg B (0.0 to 1.0 NT) ...........................................5-17

Figure 25 Steam mass flow through the main line (0.0 to 1.0 NT)..........................................5-18

Figure 26 Steam mass flow through the main line (0.0 to 0.1 NT)..........................................5-18

Figure 27 Steam mass flow through the main line (0.0 to 1.0 NT)..........................................5-19

Figure 28 Steam mass flow through the main line (0.0 to 0.1 NT)..........................................5-19

Figure 29 SG A relief valve mass flow rate (0.0 to 1.0 NT).....................................................5-20

Figure 30 SG A relief valve mass flow rate (0.0 to 0.2 NT).....................................................5-21

Figure 31 SG B relief valve mass flow rate (0.0 to 1.0 NT).....................................................5-21
ix

Figure 32 SG B relief valve mass flow rate (0.0 to 0.2 NT).....................................................5-22

Figure 33 Collapsed liquid level in the U-tubes of the SG A. Inlet (0.0 to 1.0 NT)..................5-23

Figure 34 Collapsed liquid level in the U-tubes of the SG A. Inlet (0.0 to 0.3 NT)..................5-23

Figure 35 Collapsed liquid level in the U-tubes of the SG A. Outlet (0.0 to 1.0 NT) ...............5-24

Figure 36 Collapsed liquid level in the U-tubes of the SG A. Outlet (0.0 to 0.3 NT) ...............5-24

Figure 37 Collapsed liquid level in the U-tubes of the SG B. Inlet (0.0 to 1.0 NT)..................5-25

Figure 38 Collapsed liquid level in the U-tubes of the SG B. Outlet (0.0 to 1.0 NT) ...............5-25

Figure 39 Steam generator A. Secondary-side collapsed liquid level (0.0 to 1.0 NT) ............5-26

Figure 40 Steam generator A. Secondary-side collapsed liquid level (0.0 to 0.2 NT) ............5-27

Figure 41 Steam generator B. Secondary-side collapsed liquid level (0.0 to 1.0 NT) ............5-27

Figure 42 Steam generator B. Secondary-side collapsed liquid level (0.0 to 0.4 NT) ............5-28

Figure 43 PZR liquid level.......................................................................................................5-29

Figure 44 Accumulator liquid level in loop A...........................................................................5-29

Figure 45 Accumulator liquid level in loop B...........................................................................5-30

Figure 46 Maximum hot leg fluid temperature ........................................................................5-30

Figure 47 Downcomer fluid temperature (level 5.5 m)............................................................5-31
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..................................................................................................... 3-3

Table 4. Curve of predetermined pressure drop after the beginning of the AM management action in 

the secondary-side of the steam generator in the loop without PZR................................................... 3-3

Table 5. Core protection system logic........................................................................................................... 3-4

Table 6. Number of heaters per heat structure .......................................................................................... 4-3

Table 7. Steady-state condition. Comparison between experimental and simulated values....... 5-1

Table 8. Chronological sequence of events. Comparison between experiment and TRACE ..... 5-3
xi

EXECUTIVE SUMMARY

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

pressure vessel lower plenum 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 6-2 (SB-PV-10 in JAEA). An 

asymmetrical steam generator secondary-side depressurization is produced as an accident 

management action at the steam generator in the loop without PZR, after the generation of the 

safety injection signal in order to achieve a determined depressurization rate in the primary system. 

A detailed model has been developed with TRACE5 following these assumptions. 

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

acceptable general behaviour in the entire transient. In conclusion, this work represents a small 

contribution for assessment of the predictability of thermal hydraulic 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

ACC Accumulator

AFW Auxiliary Feed Water

AIS Accumulator Injection System

AM Accident Management

CCFL Counter-current Flow Limiting

CP Coolant Pump

ECCS Emergency Core Cooling System

HPI High Pressure Injection

JAEA Japan Atomic Energy Agency

JC Jet Condenser

LSTF Large Scale Test Facility

LPI Low Pressure Injection System

MFW Main Feed Water

MSIV Main Steam Isolation Valve

PCT Peak Cladding Temperature

PGIT Primary Gravity Injection Tank

PV Pressure Vessel

PWR Pressurized Water Reactor

PZR Pressurizer

RHR Residual Heat Removal

RV Relief Valve

SBLOCA Small Break Loss-of-coolant Accident

SG Steam Generator

ST Storage Tank
xvi
1-1

1 INTRODUCTION

When a Loss-Of-Coolant Accident (LOCA) in a PWR reactor occurs, the Emergency Core Cooling 

System (ECCS) automatically actuates. ECCS, for a Westinghouse-type PWR, consists of three 

injection systems: High Pressure Injection (HPI) system, accumulator (ACC) injection system, and 

Low Pressure Injection (LPI) system. These systems are activated successively as the primary 

pressure decreases. However, the reactor coolant inventory can keep on decreasing if HPI system 

fails during a Small Break LOCA (SBLOCA) with a break discharge flow rate too small to 

depressurize the primary system to the ACC injection pressure. In this frame, an intentional 

asymmetrical depressurization of one Steam Generator (SG) secondary-side, by means of an 

Accident Management (AM) action is studied. The main goal of the experiment is to determine if this

depressurization is able to achieve a primary system cooling rate of -55 K/h and therefore, if this 

action is enough to ensure the core cooling in case of nitrogen gas inflow in the primary circuit. The 

advantage of this procedure over the primary-side steam relief is that it does not cause loss of the 

reactor coolant inventory.

In addition, in the last years, there has been a significant interest in the development of codes and 

methodologies for “best-estimate” analysis of LOCAs, as it can be found in the literature [7, 9-15]. 

Extensive research activities have been performed to develop thermalhydraulic codes, such as 

RELAP5 [6], ATHLET [7] and CATHARE [8], which enable a more realistic simulation of nuclear 

reactor systems. 

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

thermal-hydraulic code TRACE5 [4, 5], in the frame of OECD/NEA ROSA Project Test 6-2 (SB-PV10 in JAEA) [1], with the purpose of testing the behaviour of the code for this transient. A post-test 

analysis was performed with the main objective of assessing the code's capability in predicting the 

thermal-hydraulic responses to the simulated secondary-side depressurization procedure. The 

experiment 6-2 of the OECD/NEA ROSA project was conducted during 15th and 16th of December, 

2005 in the Large Scale Test Facility (LSTF) [3] of the Japanese Atomic Energy Agency (JAEA) [1, 

2]. 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 a SBLOCA in the lower 

plenum of the pressurized vessel.
1-2
2-1

2 ROSA LSTF DESCRIPTION

This section consists in a sketched description of the LSTF facility (in the Tokai Research 

Establishment of the JAEA) [3]. Two loops compose the primary coolant system: the primary loop A 

with the pressurizer (PZR) and the symmetrical primary loop B. Both include a primary coolant 

pump (PC) and a steam generator (SG). On the other hand, the secondary-coolant system includes 

a jet condenser (JC), a feedwater pump (PF), the auxiliary feedwater pumps (PA) and two SG 

secondary systems with a related piping system. 

The ECCSs consist of the following sub-systems: 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). A break flow storage tank (ST) stores the discharged coolant from the primary 

system. 

The pressure vessel (PV) is composed of the following elements: The upper head located above 

the upper core support plate; the upper plenum situated between the upper core support plate and 

the upper core plate; the core; the lower plenum and the downcomer annulus region which 

surrounds the core and upper plenum. LSTF vessel is structured with 8 spray nozzles (of 3.4 mm 

inner-diameter) at the upper head, and 8 control rod guide tubes (CRGTs) which lead the flow path 

between the upper head and the upper plenum. 

Each steam generator (SG) contains 141 U-tubes grouped 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). All the U-tubes are characterized with an inner diameter of 19.6 mm and an outer 

diameter of 25.4 mm (2.9 mm of wall thickness). The total inner and outer surface areas are 

therefore 171 and 222 m2, respectively. Regarding to the vessel, plenum and riser of steam 

generators, the inner heights are 19.840, 1.183 and 17.827 m, respectively. The downcomer is 

14.101 m.
2-2
3-1

3 TRANSIENT DESCRIPTION

The control logic of the transient is listed in Table 2. The break unit is connected to the PV lower 

plenum nozzle at 1.735 m below the core bottom elevation. After the break started, the primary 

coolant was discharged through the break and accumulated in the storage tank.

The experiment was initiated by quickly opening the break valve at time zero. The inner diameter of 

the break is 3.2 mm, satisfying the assumption of 0.1% of the size of the cold leg. Simultaneously, 

rotational speed of primary coolant pumps was increased up to 1500 rpm. A scram signal was 

generated when the PZR pressure dropped below a determined value. This signal produces the 

initiation of the core power decay curve, calculated by considering the stored heat in fuel rods and 

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). When liquid level PZR is lower 

than certain height, heaters (backup and proportional) are turned off. NT stands for Normalized 

Time and NV for Normalized Value.

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

Event Condition

Break. Time zero.

Generation of scram signal. Primary pressure drops to 0.81 NV.

PZR heaters off. Generation of scram signal or PZR 

liquid level below 0.306 NV.

Initiation of the core power decay curve. Generation of scram signal.

Initiation of primary pumps stopping curve. Generation of scram signal.

Turbine signal (turbine trip). Generation of scram signal.

Main steam isolation valve (MSIV) closure. Generation of scram signal.

Feed water termination. Generation of scram signal.

Generation of safety injection (SI) signal. Primary pressure drops to 0.76 NV.

Initiation of auxiliary feed water. Generation of safety injection SI 

signal.

Asymmetrical depressurization at the secondary of the 

steam generator in the loop without PZR.

Little after the generation of the SI 

signal.

Initiation of the accumulation system. Primary pressure drops to 0.28 NV.

Initiation of low pressure injection system. Pressure of the lower plenum of the 

vessel below 0.07 NV.
3-2

Table 2 Predetermined core power decay curve

Normalized Time Normalized

Power Normalized Time NormalizedPower

0 1 0.0033 0.3042

0.00075 1 0.0042 0.2763

0.00083 0.8150 0.0063 0.2423

0.0013 0.5366 0.0083 0.2263

0.0017 0.4504 0.0125 0.2079

0.0021 0.3906 0.0167 0.2000

0.0025 0.3538 0.0208 0.1913

Normalized Time Normalized

Power Normalized Time

Normalized

Power

0.025 0.1832 0.208 0.0936

0.033 0.1577 0.25 0.0886

0.0417 0.1487 0.33 0.0814

0.0625 0.1342 0.41 0.0763

0.0833 0.1238 0.83 0.0629

0.125 0.1096 - -

0.166 0.1003 - -

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

rotational speed curve (Table 3). Pumps are completely stopped 0.01NT after scram signal 

generation. 
3-3

Table 3 Pumps relative rotational speed

Normalized

Time

Relative

rotational 

d

Normalized

Time

Relative

rotational

d

Normalized

Time

Relative

rotational 

d 0 1.000 0.0013 0.280 0.0033 0.125

0.000083 0.850 0.0017 0.220 0.0037 0.110

0.000208 0.730 0.0021 0.185 0.0042 0.100

0.00041 0.540 0.0025 0.160 0.0104 0.000

0.00083 0.370 0.0029 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 primary pressure, followed by a new decrease due to core power decay effect. Simultaneously, 

main feedwater flow of both steam generators is stopped. The Safety Injection (SI) signal is 

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

Relief Valves (RV) in both SGs, begin opening and closing in order to maintain the pressure 

between two fixed values.

Around 0.0012 NT after the SI signal, the action of asymmetrical depressurization in the secondaryside of the SG B (in loop without PZR) begins in order to achieve a depressurization rate of -55 K/h 

in the primary system. This depressurization is produced by closing the RV and gradually opening 

the manually operated valve (MV), according to a pre-determined pressure drop curve (see Table 

4). On the contrary, the relief valve of Steam Generator A (loop with PZR) is closed after initiation of 

the accident management action. 

Table 4 Curve of predetermined pressure drop after the beginning of the AM 

management action in the secondary-side of the steam generator in the loop without PZR

Normalized 

time

Normalized 

Value 

(pressure)

Normalized 

time

Normalized 

Value 

(pressure)

Normalized 

time

Normalized 

Value 

(pressure)

0 0.504 0.0013 0.329 0.0031 0.161

0.000041 0.496 0.0019 0.261 0.0037 0.122

0.00062 0.409 0.0025 0.209 0.0044 0.094
3-4

When primary pressure drops to 0.28 NV, the accumulator 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 0.07 NV, the Low Pressure Injection System 

(LPIS) is initiated. 

In order to protect the facility, the core power is automatically decreased by the core protection 

system when the maximum fuel rod surface temperature excess 958 K, as it can be seen in Table 

5.

Table 5 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
4-1

4 APPLIED METHOD: TRACE5 MODEL OF ROSA LSTF

TRACE5 code is designed to perform best-estimate analyses of LOCAs or operational transients. 

However, it has some limitations in use; for example, those transients in which one expects to 

observe thermal stratification of the liquid phase in the 1D components cannot be directly modeled. 

Furthermore, it is not appropriate for modeling situations in which transfer of momentum plays an 

important role at a localized level [4]. 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, PZR heaters and heat losses) have been considered. 

Figure 1 shows the nodalization of the model using SNAP (Symbolic Nuclear Analysis Package 

software).

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 [3]. 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. The upper core support 

plate is located in level 16. Finally, upper head is defined between levels 17 to 19. 3D-VESSEL 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 controls 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 6.
4-3

Table 6 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

Figure 3 Steam generator nodalization
4-4

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 

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 gas phase to flow through the cell interface. 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 [4, 5]. 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.

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 [5]. 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

5.1 Steady-state

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

experimental values. In Table 7, the relative errors (%) between experimental and simulated results 

for different items are listed. It is important to remark that in any case, the maximum difference 

between experiment and simulation is 5%. In order to achieve the steady state conditions, the 

duration of simulation was stated to 5000 seconds. 

Table 7 Steady-state condition. Comparison between experimental and simulated values

Item Relative Error (%)

(Loop with PZR)

Core Power 0.0

Hot leg Fluid Temperature 0.1

Cold leg Fluid Temperature 0.1

Mass Flow Rate 5.0

PZR Pressure 0.2

PZR Liquid Level 4.3

Accumulator System Pressure 0.2

Accumulator System Temperature 0.5

SG Secondary-side Pressure 0.7

SG Secondary-side Liquid Level 3.8

Steam Flow Rate 0.4

Main feedwater Flow Rate 0.0

Main feedwater Temperature 0.3

Auxiliary Feedwater Temperature 0.3
5-2

5.2 Transient

In order to analyze results, the full transient has been divided into three phases, each of them 

characterized by a significant action or event. The first one, from 0 to 0.101 NT, starts at time 0 

when the break is produced and finishes when the Accident Management (AM) action is initiated. 

The following phase consists of a long depressurization process promoted by an asymmetrical

steam generator depressurization action (0.101 -0.401NT). In the third part of the transient, the 

depressurization process is degraded by the non-condensable gas inflow from the Accumulator

Injection System AIS tanks before the LPI actuation (0.401-0.961 NT). LPI produces a repeated 

core quench process. 

Table 8 lists the chronology sequence of events during the transient and the comparison in 

Normalized Time (1 NT = Total Transient Time) between the experiment and TRACE results. 

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 discharge coolant inventory 

through 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

Table 8 Chronological sequence of events. Comparison between experiment and TRACE

Event Experiment.

Normalized Time (s)

TRACE. Normalized 

Time (s)

Break. 0 0

Power increase of the proportional heaters of the 

PZR. 0.00008 0.00008

Power increase of the base heaters of the PZR. 0.0006 0.00064

PZR heaters turned off due to low water level. 0.0134 0.014

Scram signal generation (primary pressure 

drops to a determined value). 0.0227 0.0222

Pumps predetermined stopping curve initiation. 0.0228 0.023

Core power decay curve initiation. 0.0235 0.023

SI signal generation (primary pressure drops to 

a determined value). 0.0294 0.0292

Pumps stop. 0.0328 0.0328

Steam generator in loop without PZR

depressurization 0.101 0.101

Closure of Relief Valve at both loops with and 

without PZR. 0.107 0.106

End of Auxiliary feed water in SG in loop with 

PZR. 0.105 0.105

Change of phase in break mass flow (from liquid 

to two- phase) 0.178 0.196

Initiation of accumulator injection to cold leg in 

loop with PZR. 0.204 0.204

Initiation of accumulator injection to cold leg in 

loop without PZR. 0.205 0.204

Initiation of nitrogen gas injection to the cold leg 

of loop with PZR. 0.401 0.382

Initiation of nitrogen gas injection to the cold leg 

of loop without PZR. 0.442 0.383

Initiation of Low Pressure Injection. 0.900 Not reached
5-4

5.3 System pressures

Figure 4 compares 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 

0.0234 NT after the break, when the primary pressure decreases to 0.81 NV (Normalized Value). 

The generation of the scram signal causes the MSIV of steam generators to close and the 

beginning of the primary coolant pumps coastdown. The SI signal is generated at 0.0294 NT when 

the primary pressure decreases to a determined value. The secondary pressure increases rapidly 

during a short time after the closure of the MSIVs. From this moment on, the secondary pressure 

starts to oscillate between 0.501 NV and 0.488 NV by means of opening and closing the relief 

valves (RV) of steam generators, allowing a mass flow rate of 0.48 NV to flow through RVs.

In the second part of the transient, a controlled depressurization of the secondary side of SG B (in 

loop without PZR) is produced. At same time, relief valve of SG A is fully closed. Behavior of SG A 

and SG B are completely different, due to the accident management action itself. SG B secondary 

pressure began to decrease after the AM action started. On the other hand, the SG A secondary 

pressure temporarily increased after the closure of SG RV because of a continued primary-tosecondary heat transfer while the secondary pressure is lower than the primary pressure. After a 

delay, primary pressure begins following the pressure drop of the SG B secondary-side due to 

single-phase liquid natural circulation. Approximately, at 0.16 NT, starts a two-phase natural 

circulation flow, decreasing primary pressure. At this time, SG A secondary pressure gradually 

decreased and remained high the rest of the transient. The slight loss of pressure in SGA is due to 

thermal contraction of coolant and heat losses to environment. Furthermore, depressurization of SG 

B strongly depends on the predetermined pressure drop curve established by the accident 

management action. 

The good agreement between both experimental and simulated secondary pressures has been 

achieved by introducing the appropriate opening sequence of the accident management valve (AM 

valve).

Accumulator system is initiated at about 0.204 NT when the primary pressure decreased to 0.28 

NV. The coolant injection from the accumulator system was finished when the primary pressure 

decreased to 0.1 NV, followed by the discharge of nitrogen gas from the accumulator tanks. TRACE 

reproduces adequately the liquid level decrease of accumulators (as will be presented in other

section) and the entrance of nitrogen gas. Approximately, at 0.44 NT accumulators get completely 

empty and nitrogen-gas starts to get injected into both cold legs. In the experiment, an 

accumulation of non-condensable gas is produced in the top of the U-tubes, causing two 

consecutive temporal increasing in pressure. On the contrary, the SGB secondary pressure slightly 

decreased when the primary pressure increased. This effect has not been reproduced by TRACE5 

because nitrogen transport is characterized with the same equations than water vapor-phase. 

When vessel lower plenum pressure reached a determined value, the low pressure injection (LPI) 

system actuated at about 0.96 NT (in the experiment) but this low pressure value was not reached 

in the simulation, given as a result the no actuation of the LPI system. This lack of inventory in the 

system can be seen in Figures 14 and 15 (PV core and upper plenum collapsed liquid level, 

respectively).
5-5

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

1.4

Normalized time

Normalized values

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 (0.0 to 1.0 NT)

0 0.02 0.04 0.06 0.08 0.1 0.12 0.14 0.16 0.18 0.2 0

0.2

0.4

0.6

0.8

1

Normalized time

Normalized values

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 (0.0 to 0.25 NT)

In general, TRACE5 reproduces successfully the pressure evolution during the whole transient. The 

main disagreements are found during the nitrogen flowing in the primary circuit, especially in the 

circulation through the U-tubes.
5-6

5.4 Break

Figures 6 and 7 show the mass flow through the break and the coolant inventory, respectively. 

Experimentally, mass flow in the break is obtained from the liquid level in the break flow storage 

tank. In order to adjust the break mass flow and the discharged primary coolant inventory with 

TRACE, a sensitivity analysis varying the discharge coefficient was performed. A discharge 

coefficient of 0.78, using Choked flow model [4, 5], fits successfully both mass flow break and 

discharged coolant inventory. TRACE5 estimates a change of flow from liquid single-phase to twophase at around 0.16 NT, which completely agrees with the experimental measurements. A proper 

adjust of the discharged mass flow inventory gives a high reliability to the TRACE model. At this 

point is very important to remark the conditions at which the discharge is carried out with TRACE. 

The break has been simulated throughout a PIPE component, which connects with a VALVE 

component and finally a BREAK in order to establish the boundary conditions. The BREAK has 

been modelled following the recommendations of the user’s guide of TRACE. 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

-0.4

-0.2

0

0.2

0.4

0.6

0.8

1

Normalized Time NT

Normalized Value

 

Break mass flow rate

TRACE5

Figure 6 Break mass flow rate (0.0 to 1.0 NT)
5-7

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 NT

Normalized Value

 

Discharge Coolant Inventory

TRACE flow rate

Figure 7 Discharged inventory through the break (0.0 to 1.0 NT)

5.5 Primary loops mass flow

Primary mass flow in loop A and B are shown in figures 8, 9, 10 and 11. The initial steady state 

flows were rapidly increased by primary coolant pumps (PC) at the break time and were maintained 

at 1 NV in each loop until the start of pump coast-down. In the first part of the transient, mass flow 

starts to decrease due to the pump coast-down. After the pumps were stopped, only natural 

circulation flow in each loop contributed to cool down the core. The primary coolant mass gradually

decreased according to the break flow. Loop A and B begin to register different flow conditions after 

the AM action, flow stagnation in loop A and natural circulation flow in loop B, producing an 

asymmetric coolant mass flow rate in the primary loops. Different loop flow conditions are well 

predicted by TRACE5.

At 0.2 NT approximately, accumulators start to inject cold water in primary loops. When 

accumulators become completely empty, starts a continuous nitrogen injection in the system. 

Coincident with the nitrogen entrance, some fluctuations appear in the mass flow in loop B 

calculated by TRACE5. Furthermore, calculation time becomes lower when nitrogen is injected in 

the system. In general, all tendencies are well reproduced by TRACE5.
5-8

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

Primary Loop A masflow

TRACE5 A

Figure 8 Primary mass flow, loop A (0.0 to 1.0 NT)

0 0.02 0.04 0.06 0.08 0.1 0.12 0.14 0.16 0.18 0.2 0

0.2

0.4

0.6

0.8

1

Normalized time

Normalized values

Primary Loop A masflow

TRACE5 A

Figure 9 Primary mass flow, loop A (0.0 to 0.2 NT)
5-9

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

Primary LoopB massflow

TRACE5 B

Figure 10 Primary mass flow, loop B (0.0 to 1.0 NT)

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0

0.2

0.4

0.6

0.8

1

1.2

Normalized time

Normalized values

Primary LoopB massflow

TRACE5 B

Figure 11 Primary mass flow, loop B (0.0 to 0.4 NT)

In the previous figure two consecutive flow stagnations (in loop B at 0.44 NT and 0.47 NT in the 

experiment) are shown, during the gas inflow process. Due to the inadequate reproduction of 

nitrogen transport with TRACE, loop B flow stagnation during the gas inflow is not properly 

reproduced. 
5-10

5.6 Vessel collapsed liquid levels

The following figures show a comparison between the collapsed liquid levels in the upper plenum, 

core and downcomer, respectively for both experimental and TRACE 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 

AIS coolant inventory through the break. 

It is clearly shown that TRACE adequately reproduces all collapsed liquid levels during the first and 

second part of the transient (until gas inflow in the primary circuits). However, when nitrogen inflow 

starts, TRACE predicts a collapsed liquid drop in the upper plenum, which is not registered in the 

experiment. 

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

Measured Upper Plenum Collapsed level

TRACE5

Figure 12 Upper plenum collapsed liquid level (0.0 to 1.0 NT)
5-11

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5

0.4

0.5

0.6

0.7

0.8

0.9

1

1.1

Normalized values

Measured Upper Plenum Collapsed level

TRACE5

Figure 13 Upper plenum collapsed liquid level (0.0 to 0.5 NT)

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

Measured Core Collapsed level

TRACE5

Figure 14 Core collapsed liquid level (0.0 to 1.0 NT)

Simultaneously with the TRACE upper plenum drop, also downcomer starts to decrease its liquid 

level, underpredicting one meter during the rest of the transient.
5-12

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

Measured Downcomer Collapsed level

TRACE5

Figure 15 Downcomer collapsed liquid level (0.0 to 1.0 NT)

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

UH Liquid level

Upper Head collapsed liquid level

TRACE

Figure 16 Upper head collapsed liquid level

Upper head liquid level is shown in figure 16. Experimentally, collapsed liquid level in upper head is 

obtained by difference of pressures between the condensing pot level (9.653 m) and the upper 

support plate (6.170 m). This value is compared with the one obtained by TRACE, between the 

upper head top (8.600 m) and the core support plate (6.170 m). The different upper levels used for 

measurements in the experiment, justifies the apparent disagreement between both curves. In the 
5-13

experiment, liquid level decreases sharply at the beginning of the transient and stops at the level of 

spray nozzle top, indicating steam discharge from the upper head to downcomer. This effect is not 

well reproduced by TRACE5. 

5.7 Peak cladding temperature

TRACE correctly reproduces the evolution of the maximum fuel rod temperature in the core. The 

maximum temperature is reached at 0.88 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 5). TRACE fuel rod surface temperature is in good agreement with experimental values (figure 

17). 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0.7

0.8

0.9

1

1.1

1.2

1.3

1.4

1.5

1.6

1.7

Normalized Time

Normalized Temperature

Fuel rod surface 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

However, at the end of the experimental transient during the LPI actuation, a repeated core quench 

process is produced, which is not reproduced by TRACE due to the fact that the pressure 

conditions to the activation of LPIs is not reached in the simulation.
5-14

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

total core power

TRACE5 A

Figure 18 Core power (0.0 to 1.0 NT)

5.8 Hot and cold legs liquid levels

The following figures show the liquid level in hot and cold legs, respectively. Experimentally, liquid 

level was obtained with a three gamma ray beam densitometer. 

It is shown that hot leg liquid levels are similar in the two loops, irrespective of the different loop flow 

conditions (flow stagnation in loop A and natural circulation flow in loop B). TRACE5 reproduces 

this effect successfully. At 0.2 NT collapsed liquid level starts to increase due to accumulator’s 

water injection. Fluctuations in liquid level due to nitrogen entrance are registered mainly in loop A.

Collapsed liquid level has been calculated measuring the void fraction in different cells. For this 

calculation curvature of pipe was not taken into account. 
5-15

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

1.4

Normalized time

Normalized values

Hot leg A liquid level

TRACE5 A

Figure 19 Collapsed liquid level in hot leg A (0.0 to 1.0 NT)

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0

0.2

0.4

0.6

0.8

1

1.2

1.4

Normalized time

Normalized values

Hot leg A liquid level

TRACE5 A

Figure 20 Collapsed liquid level in hot leg A (0.0 to 0.4 NT)
5-16

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

1.4

Normalized time

Normalized values

Hot leg B liquid level

TRACE5 B

Figure 21 Collapsed liquid level in hot leg B (0.0 to 1.0 NT)

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0

0.2

0.4

0.6

0.8

1

1.2

1.4

Normalized time

Normalized values

Hot leg B liquid level

TRACE5 B

Figure 22 Collapsed liquid level in hot leg B (0.0 to 0.4 NT)

In the cold leg liquid level graphics, an increase of level between 0.2 NT and 0.4 NT due to the 

entrance of water from accumulators can be observed. In this case, differences between loop A and 

B are more important, especially at 0.4 NT, when the accumulators become empty. 
5-17

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

1.4

Normalized time

Normalized values

Cold leg A liquid level

TRACE5 A

Figure 23 Collapsed liquid level in cold leg A (0.0 to 1.0 NT)

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

1.4

Normalized time

Normalized values

Cold leg B liquid level

TRACE5 B

Figure 24 Collapsed liquid level in the cold leg B (0.0 to 1.0 NT)
5-18

5.9 Steam Generator main steam mass flow rate

The main steam isolation valves are fully closed when the scram signal is generated. A good 

reproduction of the mass flow through the MSIVs of both steam generators has been achieved, as it 

can be seen in figures 25, 26, 27 and 28.

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 0

0.5

1

1.5

Normalized time

Normalized values

SGA Steam Line

TRACE5 A

Figure 25 Steam mass flow through the main line (0.0 to 1.0 NT)

0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 0

0.5

1

1.5

Normalized time

Normalized values

SGA Steam Line

TRACE5 A

Figure 26 Steam mass flow through the main line (0.0 to 0.1 NT)
5-19

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

SGB Seam Line

TRACE5 B

Figure 27 Steam mass flow through the main line (0.0 to 1.0 NT)

0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.1 0

0.2

0.4

0.6

0.8

1

1.2

Normalized time

Normalized values

SGB Seam Line

TRACE5 B

Figure 28 Steam mass flow through the main line (0.0 to 0.1 NT)
5-20

5.10 Steam Generator relief valve flow rate

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

in figures 29, 30, 31 and 32. These figures perfectly show the periods corresponding to the relief 

valves actuation. An apparent disagreement between TRACE and experimental value can be seen 

in figure 31. This discrepancy is due to the location of the signal which measures the steam mass 

flow in the model. In the curve corresponding to TRACE, the mass flow through RV of SGB is 

represented and therefore flow through AM valve is not included.

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 0

1

2

3

4

5

6

Normalized time

mass flow-reliefvalveA

TRACE

Figure 29 SG A relief valve mass flow rate (0.0 to 1.0 NT)
5-21

0 0.02 0.04 0.06 0.08 0.1 0.12 0.14 0.16 0.18 0.2 0

1

2

3

4

5

6

Normalized time

mass flow-reliefvalveA

TRACE

Figure 30 SG A relief valve mass flow rate (0.0 to 0.2 NT)

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 0

0.5

1

1.5

Normalized time

mass flow-reliefvalveB

TRACE

Figure 31 SG B relief valve mass flow rate (0.0 to 1.0 NT)
5-22

0 0.02 0.04 0.06 0.08 0.1 0.12 0.14 0.16 0.18 0.2 0

0.5

1

1.5

Normalized time

mass flow-reliefvalveB

TRACE

Figure 32 SG B relief valve mass flow rate (0.0 to 0.2 NT)

5.11 Steam generator U-tubes collapsed liquid level

U-tubes have been grouped according to similar lengths. Experimentally 6 types were considered. 

In the TRACE simulation and due to the calculation time cost together with the obtained results, a 

3-group classification has been adopted. Figures 33, 34, 35, 36, 37 and 38 show a comparison 

between the experimental and the simulated U-tube liquid level of SG A and SG B.

The collapsed liquid level in the U-tubes of SG A and B show a completely different behavior. 

Collapsed liquid level in U-tubes (SG A) started to decrease after the inlet plenum fluid temperature 

reached the saturation condition and became completely empty at 0.16 NT, as can be seen in the 

following figures. In this moment the mass flow through the primary side of loop A almost 

disappears, only existing steam natural circulation. On the other hand, the behavior of the up-flow 

(inlet) and down-flow (outlet) sides is quite similar. The signal type used by TRACE, which allows 

measuring collapsed liquid level only in vertical cells, produces the difference in level during the first 

part of the transient. This is because the cells in the top of the U-tubes (not vertical) are not 

considered by the signal. After the top level is emptied, the values become almost the same for 

both TRACE5 simulation and the experiment.
5-23

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

SGA Tube1 Inlet

SGA Tube2 Inlet

SGA Tube3 Inlet

SGA Tube4 Inlet

SGA Tube5 Inlet

SGA Tube6 Inlet

TRACE5 group1

TRACE5 group2

TRACE5 group3

Figure 33 Collapsed liquid level in the U-tubes of the SG A. Inlet (0.0 to 1.0 NT)

0 0.05 0.1 0.15 0.2 0.25 0.3 -0.2

0

0.2

0.4

0.6

0.8

1

1.2

Normalized time

Normalized values

SGA Tube1 Inlet

SGA Tube2 Inlet

SGA Tube3 Inlet

SGA Tube4 Inlet

SGA Tube5 Inlet

SGA Tube6 Inlet

TRACE5 group1

TRACE5 group2

TRACE5 group3

Figure 34 Collapsed liquid level in the U-tubes of the SG A. Inlet (0.0 to 0.3 NT)
5-24

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

SGA Tube1 Outlet

SGA Tube2 Outlet

SGA Tube3 Outlet

SGA Tube4 Outlet

SGA Tube5 Outlet

SGA Tube6 Outlet

TRACE5 group1

TRACE5 group2

TRACE5 group3

Figure 35 Collapsed liquid level in the U-tubes of the SG A. Outlet (0.0 to 1.0 NT)

0 0.05 0.1 0.15 0.2 0.25 0.3 -0.2

0

0.2

0.4

0.6

0.8

1

1.2

Normalized time

Normalized values

SGA Tube1 Outlet

SGA Tube2 Outlet

SGA Tube3 Outlet

SGA Tube4 Outlet

SGA Tube5 Outlet

SGA Tube6 Outlet

TRACE5 group1

TRACE5 group2

TRACE5 group3

Figure 36 Collapsed liquid level in the U-tubes of the SG A. Outlet (0.0 to 0.3 NT)

Figures 37 and 38 show the collapsed liquid level in U-tubes of SG B. In both experimental and 

simulated cases, liquid level started to decrease at 0.16 NT indicating two-phase natural circulation 

conditions, recovering the top level during the auxiliar feedwater injection system actuation. 

Furthermore, the collapsed liquid level is slightly different depending on the longitude of the 

considered tube. TRACE also reproduces this situation successfully, as the following figures show.
5-25

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

SGB Tube1 Inlet

SGB Tube2 Inlet

SGB Tube3 Inlet

SGB Tube4 Inlet

SGB Tube5 Inlet

SGB Tube6 Inlet

TRACE5 group1

TRACE5 group2

TRACE5 group3

Figure 37 Collapsed liquid level in the U-tubes of the SG B. Inlet (0.0 to 1.0 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

SGB Tube1 Outlet

SGB Tube2 Outlet

SGB Tube3 Outlet

SGB Tube4 Outlet

SGB Tube5 Outlet

SGB Tube6 Outlet

TRACE5 group1

TRACE5 group2

TRACE5 group3

Figure 38 Collapsed liquid level in the U-tubes of the SG B. Outlet (0.0 to 1.0 NT)

It is important to remark that strong fluctuations are registered with TRACE, when gas nitrogen is 

introduced into the primary circuit.
5-26

5.12 Steam generators secondary-side liquid level

The following figures (39, 40, 41 and 42) show the collapsed liquid level of the secondary side of 

the steam generator. Experimentally, the liquid level is kept above the U-tubes. The behaviour of 

each steam generator is completely different. SG A liquid level slightly decreased after the RV 

closure at 0.108 NT. SG B secondary-side liquid level decreases after the start of the continuous 

steam discharge through the AM valve. In both cases, TRACE successfully reproduces the 

behaviour during the entire transient. 

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 0.8

0.85

0.9

0.95

1

Normalized time

Normalized values

SGA wide range

TRACE5 A

Figure 39 Steam generator A. Secondary-side collapsed liquid level (0.0 to 1.0 NT)
5-27

0 0.02 0.04 0.06 0.08 0.1 0.12 0.14 0.16 0.18 0.2 0.8

0.85

0.9

0.95

1

Normalized time

Normalized values

SGA wide range

TRACE5 A

Figure 40 Steam generator A. Secondary-side collapsed liquid level (0.0 to 0.2 NT)

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 0.75

0.8

0.85

0.9

0.95

1

Normalized time

Normalized values

SGB wide range

TRACE5 B

Figure 41 Steam generator B. Secondary-side collapsed liquid level (0.0 to 1.0 NT)
5-28

0 0.05 0.1 0.15 0.2 0.25 0.3 0.35 0.4 0.75

0.8

0.85

0.9

0.95

1

Normalized time

Normalized values

SGB wide range

TRACE5 B

Figure 42 Steam generator B. Secondary-side collapsed liquid level (0.0 to 0.4 NT)

5.13 Pressurizer liquid level

Figure 43 shows the PZR water level. The liquid level of the PZR decreases immediately after the 

break, becoming fully empty at 0.028 NT. Furthermore, heaters of the PZR automatically connect to 

the maximum power after the break, following the logic control of the transient depressurization. 

Heaters are disconnected at 0.013 NT by the signal of minimum level of the PZR. The liquid level of 

the PZR shows a temporal recovering at 0.023 NT, concurrent with the increment of pressure of the 

primary, due to the primary coolant expansion when the temperature of the secondary rises, after 

the closure of the MSIVs.
5-29

0 0.005 0.01 0.015 0.02 0.025 0.03 0.035 0.04 0.045 0.05

-0.2

0

0.2

0.4

0.6

0.8

Normalized Time

Normalized Liquid Level

Pressurizer Liquid Level

Measured PZR liquid level

TRACE5

Figure 43 PZR liquid level

5.14 Accumulator liquid level

Figures 44 and 45 show the accumulators collapsed liquid level. The AIS starts the injection of 

coolant from the accumulator tanks at 0.204 NT, when the primary pressure falls below 0.28 NV. 

Injection finishes at 0.4 NT, starting the continuous nitrogen-gas flowing through accumulators into 

both cold legs. In any case, a good agreement between experimental and TRACE predicted values 

can be seen. 

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

Accumulator A Liquid Level

Measured ACC liquid level

TRACE5

Figure 44 Accumulator liquid level in loop A
5-30

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

Accumulator B Liquid Level

Measured ACC liquid level

TRACE5

Figure 45 Accumulator liquid level in loop B

5.15 Hot leg temperature

Figure 46 shows the temperature of the fluid in the hot leg of loop A. The most remarkable 

discrepancy is registered during the period of injection of nitrogen, at which a slight overestimation 

with TRACE is observed.

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

0.75

0.8

0.85

0.9

0.95

1

Normalized Time

Normalized Temperature

Hot leg Liquid Temperature

Measured Hot Leg Temperature

TRACE5

Figure 46 Maximum hot leg fluid temperature
5-31

In Figure 10 it is shown the primary mass flow rate in loop B. When accumulators are empty and N2

injection starts (0.4-0.5 NT), it can be observed that the predicted mass flow rate is lower than the 

experimental one. Furthermore, collapsed liquid level in U-tubes of SG-B between 0.5 and 0.7 NT is 

higher than the experimental measurement (Figures 37 and 38). In summary, it seems that natural 

circulation is not properly simulated in this situation. Mass flow rate is underpredicted and 

consequently fluid temperature is increased (See Figure 46). This increase of temperature and 

pressure also contributes to not reach the LPI set point at the end of the transient.

5.16 Downcomer fluid temperature

Regarding the fluid temperature in the downcomer (Figure 47), the most remarkable discrepancy is 

observed at the level of 5.5 meters, in which the temperature of the fluid with TRACE is again 

slightly higher than the experimental during the injection of nitrogen. Similar tendencies are 

registered in the upper plenum and the lower plenum of the vessel. The accumulation of noncondensable gas in the U-tubes top, produces an increasing of pressure and therefore an 

increasing of the primary fluid temperature. This increase can be seen in the following figure at 0.44 

NT and 0.48 NT in the experimental curve. This effect has not been reproduced with TRACE5. 

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 temperature

Downcomer Temperature (at 5.5 m)

Measured Dowcomer Temperature

TRACE5

Figure 47 Downcomer fluid temperature (level 5.5 m)
5-32
6-1

6 CONCLUSIONS

This paper contains results obtained in the simulation of the OECD/NEA ROSA Project Test 6-2 

with the code TRACE5. One of the goals of the work is to investigate the effectiveness of the 

secondary-side depressurization operation during a SBLOCA of a PWR, followed by an unlikely 

total failure of HPI, using experimental data from the integral test facility LSTF together with 

TRACE5 code analyses. The secondary-side depressurization in such an accident scenario would 

have an effect on the primary-side, depressurizing it until the pressure at which the accumulator 

actuates. Despite this fact, it does not prevent the core from being overheated at the end of the 

transient, maybe due to the non-condensable inflow. A break at the reactor vessel bottom does not 

allow gas discharge until the whole vessel becomes empty of coolant. Non-condensable gas 

remained in the primary loops may cause significant degradation of the effects of AM actions. 

Results show that TRACE5 can successfully reproduce complicated conditions of natural circulation 

and fluid stagnation in different loops, when a break flow in the lower pressurized vessel is 

produced. TRACE5 adequately predicts the mass flow and the coolant inventory discharged 

through the break. In this particular case, no change to steam single-phase is produced. 

The main discrepancies have been found during the simulation of non-condensable gas transport, 

affecting to the pressurized vessel liquid levels (mainly to upper plenum and downcomer).
6-2
7-1

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on Core Cooling in PWR Vessel Bottom Small Small Break LOCA Experiments with HPI 

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2006.

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