NUREG/IA-0410 

International 

Agreement Report

Post-Test Calculation of the 

ROSA/LSTF Test 3-2 Using 

RELAP5/Mod3.3

 

Prepared by:

V. Martinez, F. Reventós, C. Pretel

Institute of Energy Technologies

Technical University of Catalonia

ETSEIB, Av. Diagonal 647, Pav. C

08028 Barcelona, SPAIN

 

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: January 2012 

Date Published: March 2012

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0410 

International 

Agreement Report

Post-Test Calculation of the 

ROSA/LSTF Test 3-2 Using 

RELAP5/Mod3.3

 

Prepared by:

V. Martinez, F. Reventós, C. Pretel

Institute of Energy Technologies

Technical University of Catalonia

ETSEIB, Av. Diagonal 647, Pav. C

08028 Barcelona, SPAIN

 

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: January 2012 

Date Published: March 2012

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

iii

ABSTRACT

The Thermalhydraulic Studies Group of Technical University of Catalonia (UPC) holds a large 

background in nuclear safety studies in the field of Nuclear Power Plant (NPP) code simulators. 

This report analyzes with RELAP5mod3.3 the LSTF Test 3-2, which simulates a loss-offeedwater transient (LOFW) without scram and with a total failure of the high pressure injection

system (HPIS). The simulation of several phenomena related with the high power natural 

circulation has been object of study as well as the impact of the environment heat losses in the 

primary coolant discharge across the pressurizer power-operated relief valve (PORV).

v

FOREWORD

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

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

of maturity through extensive qualification, assessment and validation processes. Best-estimate analysis methods are 

increasingly used in licensing, replacing the traditional conservative approaches. Such methods include an 

assessment of the uncertainty of their results that must be taken into account when the safety acceptance criteria for 

the licensing analysis are verified.

Traditional agreements between the Nuclear Regulatory Commission of the United States of America (USNRC) and 

the Consejo de Seguridad Nuclear of Spain (CSN) in the area of nuclear safety research have given access to CSN 

to the NRC-developed best estimate thermalhydraulic codes RELAP5, TRAC-P, TRAC-B, and currently TRACE. 

These complex tools, suitable state-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 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.

 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 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

CONTENTS .................................................................................................................................vii

FIGURES.....................................................................................................................................vii

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

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

ACKNOWLEDGMENTS............................................................................................................... x

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

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

1.1 High-Power Natural Circulation Events...............................................................1-1

1.2 The OECD/NEA ROSA Project..............................................................................1-2

2. FACILITY AND TEST DESCRIPTION.............................................................................2-1

2.1 LSTF Test Facility..................................................................................................2-1

2.2 Experimental Conditions ......................................................................................2-2

2.3 Initial Conditions ...................................................................................................2-2

2.4 Test Phase .............................................................................................................2-3

3. CODE INPUT MODEL DESCRIPTION............................................................................3-1

3.1 Nodalization ...........................................................................................................3-1

3.2 Preliminary Calculations ......................................................................................3-1

4. RESULTS.........................................................................................................................4-1

4.1 Test Phase .............................................................................................................4-1

4.2 Local Phenomena..................................................................................................4-5

4.2.1 U-tubes Liquid Level Oscillation During Natural Circulation ..............4-5

4.3 Sensitivity Analysis...............................................................................................4-7

4.3.1 Impact Of The Environment Heat Losses .............................................4-7

5. RUN STATISTICS............................................................................................................5-1

6. CONCLUSIONS ...............................................................................................................6-1

7. REFERENCES .................................................................................................................7-1

FIGURES

Figure 1 Thermal-hydraulic phenomena during LOFW without scram (Courtesy of the 

OECD/NEA ROSA group) .......................................................................................1-1

Page

Figure 2 OECD/NEA ROSA project experiments (Courtesy of the OECD/NEA ROSA 

group)......................................................................................................................1-2

Figure 3 LSTF Test Facility (Courtesy of the OECD/NEA ROSA group)...............................2-1

Figure 4 LSTF core power.....................................................................................................2-2

Figure 5 Primary and secondary pressure.............................................................................4-1

Figure 6 Maximum and minimum clad temperatures.............................................................4-2

Figure 7 SG riser collapsed liquid level .................................................................................4-2
viii

Figure 8 Auxiliary feedwater..................................................................................................4-3

Figure 9 U-tubes upflow side differential pressure ................................................................4-3

Figure 10 Primary mass flow rate............................................................................................4-4

Figure 11 PZR liquid level .......................................................................................................4-4

Figure 12 Differential pressures in the SG inlet and hot leg of loop A.....................................4-5

Figure 13 U-tube temperatures ...............................................................................................4-6

Figure 14 U-tubes differential pressure during natural circulation ...........................................4-6

Figure 15 U-tubes differential pressure during reflux condensation ........................................4-7

Figure 16 PRZ pressure during conditioning phase ................................................................4-7

Figure 17 Fluid temperature in the pressurizer........................................................................4-8

Figure 18 PZR collapsed liquid level .......................................................................................4-8

Figure 19 Time-integrated discharge flow through PORV.......................................................4-9

Figure 20 Maximum clad surface temperature ........................................................................4-9

TABLES

Page

Table 1 Chronology of the main events in Test 3-2..............................................................2-3

Table 2 Steady state conditions. ..........................................................................................3-1

Table 3 Comparison between experimental and simulated main events. ............................4-1

Table 4 Run Statistics. .........................................................................................................5-1
ix

EXECUTIVE SUMMARY

Experimental research activities are being performed in Japan by the OECD/NEA ROSA project 

with the aim of obtaining thermal-hydraulic data for the validation of computer codes and models 

for system integral analyses coupled with detailed analyses of local phenomena. These 

experiments are carried out at the LSTF test facility.

This report analyses Test 3-2, which simulates a loss-of-feedwater (LOFW) transient without

scram under the assumption of a total failure of the high pressure injection system (HPIS) and 

an actuation of the auxiliary feedwater (AFW). Liquid oscillation in the U-tubes during the natural 

circulation and coolant carryover from the hot leg into the pressurizer are objective of study in 

this test.

UPC-INTE LSTF model created for the simulation of LSTF/ROSA Test 3-1 [1] has been used 

and improved for this test. Calculations have been performed using RELAP5mod3.3 code. 

In order to improve the base case, a sensitivity analysis was performed to study the impact of 

the environment heat losses in the transient. Discrepancies were corrected adjusting the mass 

discharged through the power-operated relief valve (PORV) and the collapsed liquid level in the 

pressurizer.

Many other aspects related to the nodalization were adjusted and verified in order to improve 

results. 

The general agreement between model predictions and experimental data qualifies the UPCINTE LSTF model as an appropriate tool to simulate LSTF “High Power Natural Circulation” 

Tests. This level of qualification will certainly improve after a more accurate adjustment of the 

environment heat transfer for each component of the facility.
x

ACKNOWLEDGMENTS

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.
xi

ABBREVIATIONS

AFW auxiliary feed water

ANAV Asociación Nuclear Ascó-Vandellòs

ATWS anticipated transient without scram

CCFL counter current flow limitation

CNV II Vandellos II NPP

ECCS emergency core cooling system

HPIS high pressure injection system

INTE Institut de Tècniques Energètiques

LOFW loss of feed water

LPIS low pressure injection system

LSTF large scale test facility

MFW main feed water

MSIV main steam isolation valve

NEA Nuclear Energy Agency

NPP nuclear power plant

OECD Organization for Economic Cooperation and Development

PORV power operated relief valve

PZR pressurizer

PWR pressurized water reactor

RCP reactor coolant pumps

RELAP reactor excursion and leak analysis program

ROSA rig of safety assessment

SBLOCA small break loss of coolant accident

SG steam generator

UPC Universitat Politècnica de Catalunya (Technical University of Catalonia)

1-1

1. INTRODUCTION

Several safety activities have been performed during the last decades under the auspices of the 

OECD to develop and improve computer codes. They include several experiments at integral 

test facilities like the ROSA/LSTF Test 3-2, which forms part of the OECD/NEA ROSA project.

1.1 High-Power Natural Circulation Events

High-power events are transients with failure of scram in which core power decrease is due to 

negative reactivity feedback. Depending on the transient characteristics, this situation can lead 

to a relatively high core power during a long time.

Natural circulation occurs in transients with gradual loss of mass inventory (SBLOCA or LOFW

–losses across the pressurizer relief valve due to overpressure on the primary system-). While 

there is high core power and water in the loops, vapor and liquid with high velocity exit from the 

vessel inducing supercritical flow during natural circulation. This phenomenon can modify

significantly the coolant distribution affecting the core cooling. Particularly, for a LOFW-ATWS, 

the high pressure drop in the pressurizer PORV pulls in the supercritical flow in the hot leg at 

the surge line inlet nozzle, causing a counter-current flow limitation (CCFL) in the bottom of the 

pressurizer. It avoids that coolant returns to vessel for cooling the core –see figure one-.

Figure 1 Thermal-hydraulic phenomena during LOFW without scram (Courtesy of the 

OECD/NEA ROSA group)
1-2

1.2 The OECD/NEA ROSA Project

The OECD/NEA ROSA project includes several types of experiments (see figure two) on 

ROSA/LSTF test facility with the aim of providing a wide database for the validation of computer 

codes and models for system integral analyses coupled with detailed analyses of local 

phenomena. The main phenomena to study in these experiments are multi-dimensional mixing,

stratification, parallel flows, unstable flows, convection and influences of non-condensable gas. 

Test 3-2 is included in the group of “High power natural circulation experiments”.

Figure 2 OECD/NEA ROSA project experiments (Courtesy of the OECD/NEA ROSA 

group)
2-1

2. FACILITY AND TEST DESCRIPTION

2.1 LSTF Test Facility

LSTF (see figure three) is an experimental facility designed to simulate a Westinghouse-type 4-

loop 3,420 MWth PWR under accidental conditions. It is a full-height and 1/48 volumetricallyscaled two-loop system with a maximum core power of 10 MW (14 % of the scaled PWR 

nominal core power) and pressures scaled 1:1. Loops are sized to conserve volumetric factor 

(2/48) and to simulate the same flow regime transitions in the horizontal legs (respecting L/√D 

factor).

There is one steam generator (SG) for each loop respecting the same scaling factors. They 

have 141 full-size U-tubes, inlet and outlet plenum, steam separator, steam dome, steam dryer, 

main steam line, four downcomers and other internals.

All emergency systems are represented and have a big versatility referred to their functions and 

positions. Many break locations (20) are available too.

LSTF test facility has about 1,760 measurement points that allow an exhaustive analysis of the 

tests. There are two types of data or measurements of interest: directly measured quantities 

(temperature, pressure, differential pressure), and derived quantities (from the combination of 

two or more direct measured quantities –coolant density, mass flow rate...-).

Figure 3 LSTF Test Facility (Courtesy of the OECD/NEA ROSA group)
2-2

Test 3-2 simulates a LOFW transient without scram under the assumption of a total failure of the 

HPIS system and an actuation of the AFW. Liquid accumulation in the U-tubes during the 

natural circulation and coolant carryover from the hot leg into the pressurizer are objective of 

study in this test.

2.2 Experimental Conditions

The hardware configuration of LSTF is described in references [2] and [3]. Some important 

points are the following:

• ECCs: HPI and LPI unavailable simulating loss of off-site power.

• AFW: initiated when SG collapsed liquid level is less than its 3%.

• Core power curve: pre-determined from a previous volumetrically scaled analysis 

performed with RELAP5 code using one-point-kinetics model, which reproduces the 

transient in a commercial PWR. (see reference [3] and [4] for more detailed information). 

As LSTF core power is limited to the 14% of the scaled reference plant nominal power, the 

portion higher than 10 MW is cut-off (see first 200 s of figure four).

• LSTF core protection system: Core power is modified according to the maximum fuel rod 

surface temperature

• PZR heaters: shut off with the scram signal (proportional heater) and when the PZR level 

becomes lower than its 20% (backup heater).

Figure 4 LSTF core power

2.3 Initial Conditions

Initial steady-state conditions were fixed according to the reference PWR conditions. Because of 

the LSTF initial core power (14 % of the scaled PWR nominal core power), core flow rate was 

set to 14 % of the scaled nominal flow rate to obtain the same PWR temperatures, and 

secondary pressure was raised to limit the primary-to-secondary heat transfer rate to 10 MW.

0 5000 10000 15000 20000 25000 30000

2

4

6

8

10

Power (MW)

time (s)

 experimental

 UPC-INTE model
2-3

2.4 Test Phase

The transient starts at t=0 seconds with the loss of main feedwater in the secondary system. 

After 50 seconds the scram signal is generated causing the closure of the MSIV and the stop 

valve (turbine trip); pressurizer proportional heater is switched off and coastdown of the primary 

coolant pumps is initiated. Primary and secondary pressure raise to the set point of the PORV 

and steam generators (SG) relief valves initiating a continuous cycle opening at 330 seconds. 

Thirty seconds before approximately, pumps stop and natural circulation is started, changing 

from single to two phase flow about the 1,150 seconds. During this period, SG collapsed liquid 

level decreases below its 3 % switching on the AFW and achieving a semi-equilibrium energy 

balance in the test facility until the core dry out (core power and SG levels keep more or less 

constant until 26,000 seconds during PZR PORV and SG RV valves continuous opening).

Until 12,000 seconds, natural circulation increases gradually, time in which mass flow rate drops 

as a result of the whole core boiling and a large voiding in the SG U-tubes. Then, liquid levels of 

the U-tubes start to oscillate asymmetrically affecting the primary mass flow rate, the primary-tosecondary heat transfer and the discharge across the PORV valve. About 15,800 seconds, 

pressurizer becomes empty of liquid. Natural circulation keeps until 24,000 seconds 

approximately, time in which SG inlet liquid levels drops and reflux condensation begins.

About 25,700 seconds, significant increase occurs in the fuel rod surface temperature due to 

core dry out, initiating the LSTF core protection system about 26,000 seconds. It causes core 

quenching and the primary system depressurization. At 32,000 seconds approximately, SG are 

refilled, AFW is closed and pressures of both systems become constant, so transient is finished.

The main events are described in table one:

Table 1 Chronology of the main events in Test 3-2

Event Time [s]

Loss of feedwater 0

SCRAM signal: · Turbine trip and closure MSIV 50

 · PZR proportional heater off

 · Closure of MSIVs

 · Start coast-down of primary coolant pumps

Primary coolant pumps stop 300

Initiation of the AFW 1030

Natural circulation from single phase to two phase 1150

Significant level oscillation begin in SG U-tubes (whole core boiling) About 12000

PZR became empty of liquid 15800

Initiation of reflux and condensation phase 23500

Core power decrease by LSTF core protection system 25700

Max. fuel rod surface temperature 26220

Termination of auxiliary feedwater 32100

3-1

3. CODE INPUT MODEL DESCRIPTION

3.1 Nodalization

The RELAP5 version used to perform the post-test calculations is RELAP5/MOD3.3. The input 

deck used to simulate the transient was developed for the Pos-test calculation of the 

LSTF/ROSA Test 3-1 (reference [1]). Some improvements have been performed with the aim of 

obtaining a more realistic nodalization. So, five U-tube models have been designed to 

reproduce the nine LSTF U-tube types, one simulating the shortest one, and the others 

averaging of pair wise the rest of the lengths. Steam generators have been modified in the 

secondary side too, increasing the number of volumes in the riser for tuning the slope of 

temperature, and reproducing both annulus in the top and the bottom of the downcomer. 

Furthermore, surgeline has been re-nodalized, keeping lengths and heights as well as pressure 

drops, location and orientation of its inlet nozzle in the hot leg, and reproducing the multiple 

orifices in the inlet of the pressurizer. Finally, the line between the pressurizer and the storage 

tank has been modelled taking into account suggestions referred in [5] about the discharge lines 

nodalization. 

3.2 Preliminary Calculations

Preliminary calculations of the system energy balance showed an important disagreement in the 

environment heat losses between the experimental data and the simulation, particularly during 

the conditioning phase of the transient. A sensitivity analysis was performed to evaluate and to 

adjust the heat losses, showing a significant impact in the behavior of the PZR collapsed liquid 

level and the mass inventory discharged across the PORV valve. Results of this sensitivity 

analysis are developed in more detail in 4.2.1.

Table two shows the main parameters in steady state conditions for the UPC-INTE LSTF model. 

Values are normalized to the measured steady state conditions. 

Table 2 Steady state conditions

UPC-INTE LSTF

model

(loops w / wo PZR)

Core power 1.01

Hot leg temperature 1.003 /1.003

Cold leg temperature 1.001 / 1.001

Mass flow rate (x loop) 0.98 / 0.965

Downcomer-to-hot-leg bypass 1.001 / 1.001

Pressurizer pressure 1.0001

Pressurizer liquid level 0.999

Secondary-side pressure 0.998 / 0.998

Secondary-side liquid level 1.000 / 0.9998

Main feedwater temperature 1.0 / 1.0

Auxiliar feedwater temperature 1.0

Main feedwater flow rate 1.038 / 1.047

Accumulators pressure 1.0

Accumulators temperature 1.0 / 1.0

Steam flow rate 1.034 / 1.035

4-1

4. RESULTS

4.1 Test Phase

Table three shows the chronology of the main events occurred in Test 3-2, comparing the 

experimental values with the calculated ones.

Table 3 Comparison between experimental and simulated main events

Event Experimental 

[s]

UPC-INTE LSTF 

model [s]

Loss of feedwater 0 0

SCRAM signal: · Turbine trip and closure MSIV

 · PZR proportional heater off

 · Closure of MSIV valves

 · Start coast-down of coolant pumps

50 20

Primary coolant pumps stop 300 300

Initiation of the AFW 1030 1080

Natural circulation from single phase to two phase 1150 1100

Significant level oscillation begin in SG U-tubes (whole 

core boiling) 

About 12000 About 12500

PZR became empty of liquid 15800 17100

Initiation of reflux and condensation phase 23500 23030

Core power decrease by LSTF core protection system 25700 26090

Max. fuel rod surface temperature 26220 26320

Termination of auxiliary feedwater 32100 32100

As shown in figure five, primary and secondary pressure in the UPC-INTE LSTF model have 

good agreement with the experimental data until maximum fuel clad temperature is reached at 

26,220 seconds (see figure six), simulating correctly primary-to-secondary cooldown and SG 

collapsed liquid levels (see figure seven) during the actuation of the AFW (see figure eight).

Figure 5 Primary and secondary pressure

0 5000 10000 15000 20000 25000 30000

0,4

0,5

0,6

0,7

0,8

0,9

1,0

1,1

Pressure (Norm)

time (s)

 experimental PZR 

 experimental SGA

 experimental SGB

 UPC-INTE PZR

 UPC-INTE SGA

 UPC-INTE SGB
4-2

Figure 6 Maximum and minimum clad temperatures

Figure 7 SG riser collapsed liquid level

0 5000 10000 15000 20000 25000 30000

0,9

1,0

1,1

1,2

1,3

1,4

1,5

Temperature (Norm)

time (s)

 experimental 1st axial node

 experimental 7th axial node

 UPC-INTE 1st axial node

 UPC-INTE 7th axial node

0 5000 10000 15000 20000 25000 30000

0,0

0,1

0,2

0,3

0,4

0,5

0,6

0,7

0,8

0,9

1,0

1,1

1,2

1,3

1,4 experimental SG-A

 experimental SG-B

 UPC-INTE SG-A

 UPC-INTE SG-B

collapsed liquid level (Norm)

time (s)
4-3

Figure 8 Auxiliary feedwater

Figure six shows UPC-INTE LSTF model reproduces quite well the whole core boiling and so, 

the beginning of U-tube liquid levels oscillation (see figure nine).

Figure 9 U-tubes upflow side differential pressure

0 5000 10000 15000 20000 25000 30000

0,0

0,2

0,4

0,6

0,8

1,0

 experimental SG-A

 experimental SG-B

 UPC-INTE SG-A

 UPC-INTE SG-B

mass flow rate (Norm)

time (s)

0 5000 10000 15000 20000 25000 30000

0,00

0,25

0,50

0,75

1,00

Differential pressure (Norm)

time (s)

 experimental long U-tube upflow side

 UPC-INte long U-tube upflow side
4-4

However, although simulation reproduces primary mass flow rate drop related with the whole 

core boiling (see figure 10), it exists a relative delay between UPC-INTE LSTF model and 

experimental data that affects the mass flow rate into the pressurizer and its partial refilling (see 

figure 11). Its phenomenon holds up the emptying of the pressurizer, but even in this case, all 

these events are qualitatively well reproduced in the simulation.

Figure 10 Primary mass flow rate

Figure 11 PZR liquid level

Finally, the UPC-INTE LSTF model shows a quite good agreement simulating the partial 

emptying of the SG inlet and hot leg, and the initiation of reflux condensation (see figure 12).

0 5000 10000 15000 20000 25000 30000

-0,4

0,0

0,4

0,8

1,2

1,6

2,0

mass flow rate (Norm)

time (s)

 experimental loop A

 experimental loop B

 UPC-INTE loop A

 UPC-INTE loop B

0 5000 10000 15000 20000 25000 30000

0,0

0,2

0,4

0,6

0,8

1,0

1,2

1,4

1,6

collapsed liquid level (m)

time (s)

 experimental PZR

 UPC-INTE PZR
4-5

Figure 12 Differential pressures in the SG inlet and hot leg of loop A

4.2 Local Phenomena

4.2.1 U-tubes Liquid Level Oscillation During Natural Circulation

Significant liquid level oscillations start in the U-tubes when the whole core reaches the 

saturation temperature. Although UPC-INTE LSTF model reproduces this phenomenon, it is 

worth mentioning their amplitudes decrease faster than in the experimental data (Figure nine). 

Figure 13 compares U-tubes liquid temperatures with the saturation temperature in the core exit 

and in the SG riser. Comparison shows that in the simulation, U-tubes liquid temperatures 

oscillate less in this range, generating lower condensation in the steam generator. This fact 

justifies liquid level amplitudes in the U-tubes decrease faster in the UPC-INTE LSTF model.

Figures 14 and 15 show the frequency of the U-tubes liquid oscillations during natural circulation 

and reflux condensation respectively. Comparison with the experimental data and primary and 

secondary pressures shows UPC-INTE LSTF model reproduces oscillations of liquid in the Utubes during the natural circulation. Coupling between PORV valve continuous opening and Utubes liquid level oscillations becomes evident when reflux and condensation is achieved (see 

figure 15).

0 5000 10000 15000 20000 25000 30000

0,0

0,2

0,4

0,6

0,8

1,0

1,2

Differential pressure (Norm)

time (s)

 experimental HL loop A

 experimental SG inlet loop A

 UPC-INTE HL loop A

 UPC-INTE SG inlet loop A
4-6

Figure 13 U-tube temperatures

Figure 14 U-tubes differential pressure during natural circulation

14000 14500 15000 15500 16000

0,0

0,2

0,4

0,6

0,8

1,0

1,2

1,4

14000 14500 15000 15500 16000

0,4

0,6

0,8

1,0

1,2

Differential pressure (Norm)

time (s) experimental U-tube upflow side

 UPC-INTE U-tube upflow side

 UPC-INTE PZR 

 UPC-INTE SG-A

Pressure (Norm)

0 5000 10000 15000 20000 25000 30000

0,90

0,95

1,00

1,05

Temperature (Norm)

time (s)

 experimental UT level 7

 UPC-INTE UT level 7

 UPC-INTE SG riser level 7

 UPC-INTE core exit
4-7

Figure 15 U-tubes differential pressure during reflux condensation

4.3 Sensitivity Analysis

4.3.1 Impact Of The Environment Heat Losses

Preliminary simulations taking into account experimental procedures and boundary conditions of 

LSTF/ROSA Test 3-2 showed an unexpected increase of the pressurizer primary pressure 

during the conditioning phase (see figure 16). Liquid temperatures reached saturation quickly

suggesting an overheating in the pressurizer (see figure 17). Moreover, calculation of the 

system energy balance showed an important disagreement in the environment heat losses 

between the experimental data and the simulation during the conditioning phase of the 

transient.

Figure 16 PRZ pressure during conditioning phase

24000 24500 25000 25500 26000

0,140

0,145

0,150

0,155

0,160

24000 24500 25000 25500 26000

0,4

0,6

0,8

1,0

1,2

Differential pressure (Norm)

 UPC-INTE U-tube upflow side time (s) UPC-INTE PZR 

 UPC-INTE SG-A

Pressure (Norm)

-350 -300 -250 -200 -150 -100 -50 0

0,98

0,99

1,00

1,01

1,02

1,03

Pressure (Norm)

time (s)

 experimental PZR

 UPC-INTE PZR
4-8

Figure 17 Fluid temperature in the pressurizer

Sensitivity analysis was performed over preliminary simulations to analyze and fix the 

environment heat losses. A set of simulations were modeled increasing gradually heat losses, 

especially in the pressurizer. Results showed environment heat transfer affects significantly the 

pressurizer collapsed liquid level (see figure 18) and the PORV discharge mass flow rate (see 

figure 19), delaying the dryout of the core (see figure 20) and so, the depressurization of the 

primary system as a result of the initiation of the LSTF core protection system. The UPC-INTE 

LSTF model was taken as a base case.

Figure 18 PZR collapsed liquid level

0 5000 10000 15000 20000

-0,2

0,0

0,2

0,4

0,6

0,8

1,0

1,2

1,4

1,6

1,8 experimental 

 Preliminary calculation

 Case 1

 Case 2

 UPC-INTE model

 Case 4

collapsed liquid level (Norm)

X Axis Title

-350 -300 -250 -200 -150 -100 -50 0

0,998

0,999

1,000

1,001

1,002

Temperature (Norm)

time (s)

 UPC-INTE PZR fluid

 UPC-INTE PZR saturation
4-9

Figure 19 Time-integrated discharge flow through PORV

Figure 20 Maximum clad surface temperature

0 5000 10000 15000 20000 25000 30000

-0,15

0,00

0,15

0,30

0,45

0,60

0,75

0,90

1,05

Time integrated discharge flow (Norm)

time (s)

 experimental 

 Preliminary calculation

 Case 1

 Case 2

 UPC-INTE model

 Case 4

0 5000 10000 15000 20000 25000 30000

0,8

0,9

1,0

1,1

1,2

1,3

1,4

1,5 experimental 

 Preliminary calculation

 Case 1

 Case 2

 UPC-INTE model

 Case 4

Temperature (Norm)

time (s)

5-1

5. RUN STATISTICS

The calculations were performed on a Personal Computer with 3.4 GHz Pentium IV processor, 

512 MB of RAM and Windows XP Service Pack 2 SO.

Table four shows main run statistics for all calculations performed in this report:

Table 4 Run Statistics

Transient 

time

CPU time Mass error ratio

(emass/tmass)

Preliminary calculation 35900.0 5373.34 4.3515·10-5

Case 1 35900.0 5358.19 4.11722·10-5

Case 2 35900.0 5567.07 4.3036·10-5

UPC-INTE LSTF model 35900.0 5362.62 4.3789·10-5

Case 4 35900.0 5246.41 4.4801·10-5

6-1

6. CONCLUSIONS

UPC-INTE LSTF model developed for LSTF Test 3-1 [1] has been adjusted to the LSTF Test 3-

2 proving its suitability to simulate the behavior of this facility.

Model predictions for Test 3-2 were in quite good agreement with the available experimental 

data. Several conclusions have been obtained from the study of local phenomena and 

sensitivity analysis. First, that RELAP5/mod3.3 reproduces U-tubes liquid level oscillations 

during High-Power Natural Circulation. Although condensation in U-tubes, which affects to the 

amplitude of the oscillations, is slightly under predicted by the code, the main events of the 

transient don’t change significantly. On the other hand, sensitivity analysis shows the 

significance of the environment heat losses simulation. For long transients with a relatively highcore power, an incorrect implementation of the environment heat transfer causes discrepancies 

in the core power heat removal. Particularly, for Test 3-2, they modify wrongly the mass 

discharge through the PORV.

7-1

7. REFERENCES

1. Martínez, V; Pretel, C.; Reventós, F.; Post-Test Calculation of the ROSA/LSTF Test 3-1 

using RELAP5/mod3.3; Technical report for the “Proyecto de Análisis y simulación de los 

experimentos OECD PKL y ROSA OECD. Aplicación a las CCNN. españolas”. July 2008.

2. ROSA-V Large Scale Test Facility (LSTF) system description for the Third and Fourth

Simulated fuel assemblies. The ROSA group JAERI. March 2003.

3. OECD/NEA. Quick‐look Report of ROSA/LSTF Test 3‐1 (High Power Natural Circulation 

Experiment SB‐CL‐38 in JAEA). Report. OECD/NEA ROSA Project. March 20, 2007.

4. Japan Atomic Energy Agency, 2007. A Report on Pre-test Analyses to Define 

Experimental Conditions of ROSA/LSTF Test 3-2 on High Power Natural Circulation. 

Japan Atomic Energy Agency, Private Communication.

5. Kim K., Kim H. J; Assessment of the RELAP5/mod2 Critical Flow using Marviken Test 

Data 15 and 24; USNRC; NUREG/IA-0086, April 1992.

6. RELAP5/MOD3 Code manual. Volume IV: models and correlations. June, 1999. 

SCIENTECH, Inc.






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-0410 Post-Test Calculation of the ROSA/LSTF Test 3-2

Using RELAP5/Mod 3.3