NUREG/IA-0409 

International 

Agreement Report

Post-Test Calculation of the 

ROSA/LSTF Test 3-1 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-0409 

International 

Agreement Report

Post-Test Calculation of the 

ROSA/LSTF Test 3-1 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. 

RELAP5mod3.3 has been used in this study in order to analyze the LSTF Test 3-1, which 

simulates an anticipated transient without scram (ATWS) as a result of a small break LOCA 

without rods insertion (but with scram signal) and loss of off-site power. Two local phenomena 

have been object of study: U-tube liquid accumulation due to CCFL at the inlet of the SG Utubes, and loop seal behavior during the transient.

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.... ...............................................................................................................................viii

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 Boundary Conditions............................................................................................2-2

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

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

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

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

3.2 Preliminary Calculations ......................................................................................3-4

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

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

4.2 Local Phenomena..................................................................................................4-6

4.2.1 Liquid Accumulation Due To CCFL .......................................................4-6

4.2.2 Loop Seal Behavior...............................................................................4-10

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

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

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

FIGURES

Page

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

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

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 Original supplied input nodalization.........................................................................3-1

Figure 5 Break unit nodalization ............................................................................................3-2

Figure 6 Normalized pressure vs. distance on broken loop...................................................3-3

Figure 7 Normalized pressure vs. distance on intact loop.....................................................3-3

Figure 8 Core collapsed liquid level.......................................................................................3-5

Figure 9 Upper plenum collapsed liquid level........................................................................3-5

Figure 10 Primary and secondary pressure.............................................................................4-2

Figure 11 Break mass flow ......................................................................................................4-2

Figure 12 Cold leg flow rate.....................................................................................................4-3

Figure 13 Core level ................................................................................................................4-3

Figure 14 Steam generators level ...........................................................................................4-4

Figure 15 Maximum peak clad surface temperature ...............................................................4-4

Figure 16 LSTF core power.....................................................................................................4-5

Figure 17 Core void fraction ....................................................................................................4-5

Figure 18 Froude Number .......................................................................................................4-6

Figure 19 Flow regime number. Horizontal stratification associated to 12 value.....................4-7

Figure 20 Hot leg collapsed liquid level ...................................................................................4-7

Figure 21 Loop B liquid velocity...............................................................................................4-8

Figure 22 Loop B gas velocity .................................................................................................4-8

Figure 23 U-tube liquid level on broken loop ...........................................................................4-9

Figure 24 U-tube liquid level on intact loop..............................................................................4-9

Figure 25 Core collapsed liquid level.....................................................................................4-10

Figure 26 Time-integrated accumulator mass flow rate.........................................................4-10

Figure 27 Loop seal collapsed liquid level on broken loop ....................................................4-11

Figure 28 Loop seal collapsed liquid level on intact loop.......................................................4-11

TABLES

Page

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

Table 2 Differential pressures into the vessel. .....................................................................3-4

Table 3 Steady state conditions ...........................................................................................3-4

Table 4 Comparison between experimental and simulated main events. ............................4-1
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 3.1, which simulates an anticipated transient without scram (ATWS) 

as a result of a small break LOCA without rods insertion (but with scram signal) and loss of offsite power. One of the main aspects of this test is the supercritical natural circulation during 

initial high-core power.

Two local phenomena have been object of study: U-tube liquid accumulation due to counter 

current flow limitation at the inlet of the steam generator U-tubes, and loop seal behavior during 

the transient.

A nodalization has been created to simulate this test and the following ones. Calculations have 

been performed using RELAP5mod3.3 code. 

In order to improve the base case, differential pressures around loops and vessel have been 

adjusted using initial condition experimental data, and downcomer-to-hot-leg bypass has been 

corrected to reproduce its theoretical value of 1% of core flow.

Many other aspects related to the nodalization have been adjusted and verified in order to 

improve results. 

The general agreement between model predictions and experimental data qualifies the model 

as an appropriate tool to simulate this kind of LSTF tests. This level of qualification will certainly 

improve after testing the same model in other post-test calculations
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 originated by the OECD

to develop and improve computer codes. They include experiments performed at the LSTF test 

facility for the OECD/NEA ROSA project like Test 3-1 which is object of study in this report.

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 to the hot legs inducing supercritical flow during natural circulation. This phenomenon 

affects the coolant distribution due to counter-current flow limitation (CCFL) during condensing 

reflux at the inlet of the steam generators and U-tubes which causes liquid accumulation in them 

–see figure one-.

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

OECD/NEA ROSA group).

Break flow rate and liquid carryover into the pressurizer are affected by this phenomenon too.
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 3-1 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 SG for each loop respecting the same scaling factors. They have 141 full-size Utubes, inlet and oulet plena, 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)

Test 3-1 simulates a SBLOCA (break size of 1%) with scram failure and loss-of-offsite-power

(HPI and LPI are unavailable). Due to the high-core power, supercritical natural circulation 

exists in the hot legs until the loops become empty. CCFL at the inlet of the steam generators 

and U-tubes during the two-phase natural circulation are objective of study in this test.
2-2

2.2 Boundary Conditions

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

points are the following:

• Break assembly: small break in the cold leg of the loop without pressurizer (1 % of the 

scaled cross-sectional area of the reference PWR cold leg).

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

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

performed with SKETCH-INS/TRAC-PF, which reproduces the transient in a commercial 

PWR (see Appendix A of reference [3] 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 16).

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

surface temperature

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.

2.4 Test Phase

The transient is started opening the break at t = 0 s. After 20 seconds the scram signal is 

generated causing the closure of the MSIV and the stop valve (turbine trip); pressurizer heater 

is off, main feed water is closed and auxiliary feed water is started. Three seconds later, coastdown of the primary coolant pumps is initiated. Until 300 seconds while there is high core 

power, secondary pressure rises over the specified set-point causing the continuous opening of 

the SG relief valves and generating two-phase natural circulation in the primary loops. Between 

300 and 1,600 seconds of the transient, the primary system is coupled with the isolated 

secondary system, which is depressurized with the cool auxiliary feed water that condenses 

vapor of the steam generators.

About 1,100 seconds, core liquid level starts to decrease rising the average temperature of the 

system. Then, the LSTF core protection system actuates decreasing the core power until the 

maximum fuel rod surface temperature is achieved. As a result of the low power, the primary 

pressure falls down below the secondary. About 2,100 seconds after the start of the transient, 

the accumulator injection system initiates causing a loop seal clearing in the loop without 

pressurizer 100 seconds later. At 5,547 seconds the break is closed and the transient finished.

The main events are described in table one:
2-3

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

Event Time [s]

Break 0

SCRAM signal: · Turbine trip and closure MSIV 20

 · PZR heater off

 · End of main feedwater and begin

 of auxiliary feedwater 

Start of coast-down of primary coolant pumps 23

Primary coolant pumps stop 272

End of continuous opening of SG RVs. End of two-phase natural 

circulation. Break flow from single-phase liquid to two-phase flow 

About 300 

Core liquid level starts to decrease (core uncovery) About 1100 

Core power decrease by LSTF core protection system 1630

Max. fuel rod surface temperature 1825

Primary pressure lower than SG secondary-side pressure About 1900 

Initiation of accumulator injection system About 2100 

Loop seal clearing only in loop without PZR About 2200 

End of the transient 5547

3-1

3. CODE INPUT MODEL DESCRIPTION

3.1 Nodalization

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

Consequently, it has been necessary to adjust the original supplied input (see figure four) from 

RELAP5mod3.2 to this the newest version. 

Figure 4 Original supplied input nodalization

Other improvements have been performed with the aim of obtaining a more realistic 

nodalization:

• Re-nodalization of the cold leg in the broken loop respecting the distance between pump,

accumulator nozzle, break orifice and core inlet.

• Simulation of the bypass between the UH and the UP with an annulus around the upper 

core support plate and a multiple junction reproducing the orifices at the upperhead 

bottom.
3-2

• New break unit that reproduces pipes and junctions between the orifice assembly and the 

storage tank (see figure five).

Figure 5 Break unit nodalization

Finally, differential pressures around the loops (figures six and seven) and into the vessel (Table 

two) have been adjusted to improve the steady-state conditions. Figures six and seven show an 

important divergence in the pump inlet. Table two shows the deviation in kPa between predicted 

and experimental data. As RELAP code computes pressure in the middle of the volumes, 

differential pressures have been corrected to level the heights
3-3

Figure 6 Normalized pressure vs. distance on broken loop

Figure 7 Normalized pressure vs. distance on intact loop.

0 10 20 30 40

0,9996

0,9998

1,0000

1,0002

1,0004

1,0006

1,0008

1,0010

1,0012

normalized pressure (P/Pcore outlet)

distance (m)

 experimental

 UPC-INTE

CO

UT

PI

CI

0 10 20 30 40

0,9996

0,9998

1,0000

1,0002

1,0004

1,0006

1,0008

1,0010

1,0012

CI

PI

UT

CO

normalized pressure (P/Pcore outlet)

distance (m)

 experimental

 UPC-INTE
3-4

Table 2 Differential pressures into the vessel.

Vessel Elevation

(m)

Differential pressures

(kPa)

Differential pressures

-correcting elevation- (kPa)

Upper head -2.19 -9.58 -1.00

Upper plenum -2.13 -0.63 -0.30

Core -3.64 -1.39 0.54

Lower plenum total -1.42 -3.96 0.60

Downcomer -8.82 1.53 0.00

TOTAL -8.21 28.63 0.40

3.2 Preliminary Calculations

Preliminary calculations show an important divergence in the downcomer-to-hot leg bypass 

mass flow rate (see table three; values are normalized to the measured steady state 

conditions). 

Table 3 Steady state conditions

Preliminary RELAP 

simulation

(loops w / wo PZR)

UPC-INTE RELAP 

model

(loops w / wo PZR)

Core power 0.990 0.990

Hot leg temperatura 1.0 / 0.9997 1.0 /0.9997

Cold leg temperatura 1.001 / 1.0 1.001 / 1.0

Mass flow rate (x loop) 1.04 / 1.021 1.04 / 1.021

Downcomer-to-hot-leg bypass 7.042 / 7.021 1.001 / 1.001

Pressurizer pressure 1.003 1.003

Pressurizer liquid level 0.971 0.971

Secondary-side pressure 0.998 / 0.998 0.998 / 0.998

Secondary-side liquid level 1.003 / 0.998 1.003 / 0.998

Main feedwater temperatura 1.001 /0.999 1.001 / 0.999

Auxiliar feedwater temperatura 1.0 1.0

Main feedwater flow rate 1.008 / 1.031 1.008 / 1.031

Accumulators pressure 1.0 1.0

Accumulators temperatura 1.0 / 1.0 1.0 / 1.0

Steam flow rate 1.002/ 1.029 1.002 / 1.029

Correcting the bypass mass flow rate, both modifying the section (0.58·10-4 m2) and modifying 

the flow energy loss coefficients (345.0), the drop delay in the core collapsed liquid level of the 

preliminary simulation disappears (see figure eight).
3-5

Figure 8 Core collapsed liquid level

Although both cases improve it, only the second partly reproduces the upper plenum refilling 

(see figure nine). So the final nodalization incorporates the new flow energy loss coefficients.

The other main parameters of the transient are not meaningfully affected by the downcomer-tohot leg bypass mass flow.

Figure 9 Upper plenum collapsed liquid level

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

1,2

1,4

liquid level (Norm)

time (s)

 experimental

 UPC-INTE model

 UPC-INTE model bypass section =0.58·10-4 m2

 UPC-INTE model bypass flow energy loss coef.= 345.0

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

1,2 experimental

 UPC-INTE model

 UPC-INTE model bypass section =0.58·10-4 m2

 UPC-INTE model bypass charge losses coef.= 345.0

liquid level (Norm)

time (s)

4-1

4. RESULTS

4.1 Test Phase

Table four shows the chronology of the main events occurred in Test 3-1, comparing the 

experimental values with the calculated ones.

Table 4 Comparison between experimental and simulated main events

Event Experimental 

[s]

UPC-INTE model 

[s]

Break 0 0

SCRAM signal: · Turbine trip and closure MSIV

 · PZR heater off

 · End of main feedwater and begin

 of auxiliary feedwater

20 20

Start of coastdown of primary coolant pumps 23 23

Primary coolant pumps stop 272 272

End of continuous opening of SG RVs, End of two-phase 

natural circulation, break flow from single-phase liquid to 

two-phase flow 

About 300 300-400

Core liquid level starts to decrease (core uncovery) About 1100 About 1100

Core power decrease by LSTF core protection system 1630 1707

Max. fuel rod surface temperature 1825 1875

Primary pressure lower than SG secondary-side pressure About 1900 1875

Initiation of accumulator injection system About 2100 2180

Loop seal clearing only in loop without PZR About 2200 2898

End of the transient 5547 5547

As shown in figure 10, primary and secondary pressure have good agreement with the 

experimental data until 2,100 seconds, when the initiation of the accumulator injection system

causes some discrepancies on primary pressure and break mass flow (see figure 11). In the 

UPC-INTE model, accumulators refill cold legs increasing the mass flow across the break.
4-2

Figure 10 Primary and secondary pressure

Figure 11 Break mass flow

Figures 12 and 13 show how the UPC-INTE model reproduces natural circulation and emptying 

of the core.

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

Pressure (Norm)

time (s)

 experimental primary pressure

 experimental secondary pressure loop A

 experimental secondary pressure loop B

 UPC-INTE model primary pressure

 UPC-INTE model secondary pressure loop A

 UPC-INTE model secondary pressure loop B

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

1,2

1,4

0,0

0,2

0,4

0,6

0,8

1,0

time-integrated break flow (Norm)

 experimental break

 UPC-INTE model break

Flow rate (Norm)

time (s)

 experimental break 

 UPC-INTE model break
4-3

Figure 12 Cold leg flow rate

Figure 13 Core level

The UPC-INTE model performs a good secondary cooldown. It reproduces continuous opening

of the relief valves during natural circulation (see secondary pressure in figure 10) and adjusts 

correctly the steam generators collapsed liquid level (see figure 14).

0 200 400 600 800 1000

0,0

0,4

0,8

1,2

1,6

Flow rate (Norm)

time (s)

 experimental loop A

 experimental loop B

 UPC-INTE loop A

 UPC-INTE loop B

0 1000 2000 3000 4000

0,00

0,25

0,50

0,75

1,00

1,25

1,50

liquid level (Norm)

time (s)

 experimental core

 UPC-INTE model core
4-4

Figure 14 Steam generators level

Figure 15 shows rod surface temperature has a quite good agreement with experimental data 

until the initiation of accumulators because of a correct LSTF core protection system 

implementation (see figure 16).

Figure 15 Maximum peak clad surface temperature

0 1000 2000 3000 4000

0,6

0,8

1,0

1,2

1,4

1,6

1,8

2,0

2,2

liquid level (Norm)

time (s)

 experimental SG A

 experimental SG B

 UPC-INTE model SG A

 UPC-INTE model SG B

0 1000 2000 3000 4000

0,6

0,8

1,0

1,2

1,4

rod surface temperature (Norm)

time (s)

 experimental

 UPC-INTE model
4-5

Figure 16 LSTF core power

Finally, it is observed that with a pre-determined core power curve (figure 16) RELAP code 

predicts the increase in core void fraction (figure 17) that induces the negative reactivity 

feedback which is the main factor of core power drop.

Figure 17 Core void fraction

0 200 400 600 800 1000

0,0

0,2

0,4

0,6

0,8

1,0

void fraction

time (s)

 core void fraction UPC-INTE model

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

Power (Norm)

time (s)

 experimental core power

 UPC-INTE model core power
4-6

4.2 Local Phenomena

4.2.1 Liquid Accumulation Due To CCFL:

Supercritical flow during two-phase flow Natural Circulation induces liquid accumulation in the 

U-tubes during reflux condensation because of counter current flow limitation in the inlet U-tube 

and in the bottom of the inlet plenum (see figure one). Partial core drop is observed as a result 

of this accumulation.

RELAP5mod3.3 reproduces supercritical flow (Froude number > 1) during the two-phase flow 

natural circulation (see figure 18) and simulates related phenomena like horizontal stratification 

in the hot leg (figure 19) and a partial drop of its level during supercritical flow (figure 20 shows 

an asymmetrical drop of the UPC-INTE model collapsed liquid level during a 100-200 seconds 

interval which seems to be related to the Froude number values of figure 18).

Figure 18 Froude Number

0 100 200 300 400 500 600 700 800

0,0

0,5

1,0

1,5

2,0

2,5

3,0

Froude Number

time (s)

 UPC-INTE model liquid flow loop A

 UPC-INTE model liquid flow loop A
4-7

Figure 19 Flow regime number. Horizontal stratification associated to 12 value

Figure 20 Hot leg collapsed liquid level

RELAP5/mod3.3 simulates a small negative liquid velocity (figure 21) and a positive gas velocity

(figure 22) at the U-tube inlet during one-phase gas flow with high vapor velocity (from 400 

seconds to 800 seconds approximately). These velocities justify counter current limitation at the

U-tubes inlet and a possible liquid accumulation in them.

0 1000 2000 3000 4000 5000

0

2

4

6

8

10

12

Flow regim number

time (s)

 UPC-INTE model hot leg loop A

 UPC-INTE model hot leg loop B

0 100 200 300 400 500 600 700 800

0,00

0,25

0,50

0,75

1,00

liquid level (Norm)

time (s)

 experimental hot leg A

 exerimental hot leg B

 UPC-INTE model hot leg A

 UPC-INTE model hot leg B
4-8

Figure 21 Loop B liquid velocity

Figure 22 Loop B gas velocity

As shown in figures 23 and 24, although asymmetrical effect is not reproduced, the UPC-INTE 

model simulates the U-tube liquid accumulation phenomenon and its related partial core level 

drop (see figure 25).

0 200 400 600 800 1000

-2

0

2

velocity (Norm)

time (s)

 UPC-INTE model hot leg

 UPC-INTE model UT intlet

0 200 400 600 800 1000

-1

0

1

2

3 UPC-INTE model hot leg

 UPC-INTE model UT intlet

velocity (Norm)

time (s)
4-9

Figure 23 U-tube liquid level on broken loop

Figure 24 U-tube liquid level on intact loop

0 200 400 600 800 1000

0,0

0,2

0,4

0,6

0,8

1,0

liquid level (Norm)

time (s)

 experimental U-tube A

 UPC-INTE model U-tube A 

0 200 400 600 800 1000

0,0

0,2

0,4

0,6

0,8

1,0

1,2

liquid level (Norm)

time (s)

 experimental U-tube B

 UPC-INTE U-tube B
4-10

Figure 25 Core collapsed liquid level

4.2.2 Loop Seal Behavior

The UPC-INTE model shows discrepancies in primary pressure and break mass flow rate after 

the initiation of the accumulator injection system. In the simulation there is an important delay 

between the accumulators entrance (2180 seconds –see figure 26- ) and the loop seal clearing 

in the broken loop (2898 seconds –see figure 27; black- green lines show the level between the 

SG outlet and the loop seal bottom, and red-blue lines show the level between the loop seal 

bottom and the pump inlet-).

Figure 26 Time-integrated accumulator mass flow rate

0 1000 2000 3000 4000

0,00

0,25

0,50

0,75

1,00

1,25

1,50

collapsed level (Norm)

time (s)

 experimental core

 UPC-INTE model core

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

1,2

Time-integrated accumulator flow (Norm)

time (s)

 experimental loop A

 experimental loop B

 UPC-INTE model loop A

 UPC-INTE model loop B
4-11

Figure 27 Loop seal collapsed liquid level on broken loop

Figure 28 Loop seal collapsed liquid level on intact loop

As shown in figure 26, there are important discrepancies in the accumulator injection. This could 

be one of the factors why water distribution around the loops is different between the UPC-INTE

model and the experimental data (see in figures 27 and 28 the opposite loop seal behavior after 

the initiation of accumulators system).

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

collapsed level (Norm)

time (s)

 experimental loop B SG outlet- Loop seal bottom

 experimental loop B Loop seal bottom- pump inlet

 UPC-INTE model loop B SG outlet- Loop seal bottom

 UPC-INTE model loop B Loop seal bottom- pump inlet

0 1000 2000 3000 4000

0,0

0,2

0,4

0,6

0,8

1,0

 experimental loop A SG outlet- Loop seal bottom

 experimental loop A Loop seal bottom- pump inlet

 UPC-INTE model loop A SG outlet- Loop seal bottom

 UPC-INTE model loop A Loop seal bottom- pump inlet

collapsed level (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.

The 5379.02 seconds long transient consumed 4545.58 seconds CPU time. It means the CPU 

time / transient time ratio was 0.84506. 

The mass error ratio (emass / tmass) was reasonably low: 3.10425 kg / 231,939 kg = 

1.338391E-05.

6-1

6. CONCLUSIONS

A developed model of the LSTF Test Facility has been adjusted and has proved to be a suitable 

tool to simulate the behavior of this facility.

The post-test calculation for Test 3-1 has been performed. Model predictions were in quite good 

agreement with the available experimental data. Several conclusions have been obtained from 

the study of local phenomena (U-tube liquid accumulation due to CCFL and loop seal behavior) 

and preliminary calculations:

• RELAP5/mod3.3 reproduces supercritical flow and liquid accumulation in the U-tubes 

during High-Power Natural Circulation.

• A correct implementation of the vessel bypass is strongly related with a satisfactory 

simulation of the transient.

• There is an important delay in the loop seal clearing as a result of a different accumulator 

injection and an incorrect water distribution around the primary system after their 

actuation.

• Differences in the core refilling and fluctuations in the core level suggest vessel bypasses 

and upper head could be a source of uncertainty.

7-1

7. REFERENCES

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

SCIENTECH, Inc.

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.




UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

--------------------

OFFICIAL BUSINESS
NUREG/IA

-0409 Post-Test Calculation of the ROSA/LSTF Test 3-1 

Using RELAP5/Mod 3.3