NUREG/IA-0432 

International 

 Agreement Report

Analysis of the Test OECD-PKL2 G7.1 with

the Thermal-Hydraulic System Code TRACE 

Prepared by:

S. Belaïd, J. Freixa, O. Zerkak

Paul Scherrer Institut

5232 Villigen PSI

Switzerland

K. Tien, NRC Project Manager

 Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2013

Date Published: September 2013

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-0432

International 

 Agreement Report

Analysis of the Test OECD-PKL2 G7.1 with

the Thermal-Hydraulic System Code TRACE

Prepared by:

S. Belaïd, J. Freixa, O. Zerkak

Paul Scherrer Institut

5232 Villigen PSI

Switzerland

K. Tien, NRC Project Manager

 Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: August 2013

Date Published: September 2013

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 test OECD-PKL2 G7.1 is a counterpart of the OECD/NEA ROSA-2 Test 3. These tests

consisted of a small break loss-of-coolant-accident experiment concurrent with additional 

system failures, namely the total failure of the High Pressure Safety Injection system combined 

with the absence of an early manual secondary side cool down. To prevent a core meltdown a 

manual secondary depressurization (cooldown) is needed to restore the heat sink through fast 

reduction of the secondary pressure inducing similar primary pressure behavior, and thus

leading to passive accumulator injection and low pressure safety injection. The core exit 

temperature parameter can be used to detect a core heat up and trigger such accident 

management procedures.

The test OECD-PKL2 G7.1 was performed to assess the reliability of the core exit temperature 

measurement and its correlation to the peak cladding temperature, and to investigate the

physical processes affecting the performance of the core exit temperature measurement.

In this report, the analysis of test OECD-PKL2 G7.1 using the US-NRC thermal-hydraulics best 

estimate system code TRACE and based on an existing PKL Mark III model is presented and

the obtained results are compared with the experimental data. A good agreement with the 

experimental data is observed, especially for relevant parameters such as the primary pressure.

The predicted core exit and peak cladding temperatures evolutions could also capture the 

experimental data, although some discrepancies in the shapes of the evolution curves are

observed.

v

CONTENTS

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

LIST OF FIGURES ...................................................................................................................... vii

LIST OF TABLES......................................................................................................................... ix

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

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

1 INTRODUCTION .................................................................................................................. 1

2 OBJECTIVES......................................................................................................................... 3

3 TEST FACILITY DESCRIPTION......................................................................................... 5

3.1 Overview......................................................................................................................... 5

3.2 Reactor Coolant System (RCS) ...................................................................................... 6

3.3 Rod Bundle Vessel.......................................................................................................... 6

3.4 Upper Head and Upper Plenum...................................................................................... 6

3.5 Upper Head Bypass......................................................................................................... 7

3.6 Test Bundle ..................................................................................................................... 7

4 TRACE MODEL .................................................................................................................... 9

5 BOUNDARY CONDITIONS OF TEST G7.1..................................................................... 11

5.1 Test Procedure .............................................................................................................. 11

5.2 Core Power.................................................................................................................... 12

5.3 Secondary Pressurization.............................................................................................. 12

5.4 Break Conditions .......................................................................................................... 12

5.5 Steam Generator Depressurization ............................................................................... 12

5.6 Emergency Core Cooling Systems, ECCS ................................................................... 13

6 ANALYSIS RESULTS ........................................................................................................ 15

6.1 Overview....................................................................................................................... 15

6.2 Situation at the Beginning of the Conditioning Phase .................................................. 19

6.3 Initial Test Conditions................................................................................................... 20

6.4 Situation Prior to Secondary-Side Depressurization..................................................... 21

6.5 Acc Injection................................................................................................................. 23

6.6 Situation After Acc Injection........................................................................................ 26

7 CONCLUSIONS................................................................................................................... 27

8 REFERENCES ..................................................................................................................... 29

vii

LIST OF FIGURES

Figure 1 PKL integral test facility layout 5

Figure 2 Description of the CET measurements 6

Figure 3 TRACE nodalization of the PKL test facility (only two loops are shown) 10

Figure 4 Test procedure including the conditioning phase 11

Figure 5 PKL G7.1 evolution of the RPV (blue), PZR (red) and SG (black) levels 15

Figure 6 PKL G7.1 Test pressures evolution 16

Figure 7 PKL G7.1 small break flow rate evolution 16

Figure 8 PKL G7.1 mass inventory evolution in the TRACE calculation 17

Figure 9 PKL G7.1 core mass flows evolution as calculated by TRACE 17

Figure 10 PKL G7.1 loops flow rate evolution 19

Figure 11 Initial test conditions for Test G7.1 20

Figure 12 PKL G7.1 break gas and liquid mass flows evolution 21

Figure 13 PKL G7.1 PCT & CET evolution 22

Figure 14 PKL G7.1 MS-RCVs steam flow rate evolution 22

Figure 15 PKL G7.1 Accs pressure evolution 24

Figure 16 PKL G7.1 Accs injection evolution 24

Figure 17 PKL G7.1 core void fraction evolution (calculated by TRACE) 25

Figure 18 PKL G7.1 core density evolution (calculated by TRACE) 25

Figure 19 PKL G7.1 core gas and liquid velocities evolution (calculated by TRACE) 26

Figure 20 PKL G7.1 MS-LPSI injections evolution 26

ix

LIST OF TABLES

Table 1 PKL G7.1 test - sequence of events.......................................................................... 18

Table 2 Results at the beginning of the conditioning phase .................................................. 19

Table 3 Results at the start of the test .................................................................................... 20

Table 4 Results prior to the secondary side depressurization phase ...................................... 23

xi

ABBREVIATIONS

Acc Accumulator

AM Accident Management

CET Core Exit Temperature

CL Cold Leg

DC Downcomer

ECCS Emergency Core Cooling System

FW Feed Water

HL Hot Leg

HPSI High Pressure Safety Injection

JAEA Japan Atomic Energy Agency

LP Lower Plenum

LPSI Low Pressure Safety Injection

MS-RCV Main Steam Relief Control Valve

NC Natural Circulation

PCT Peak Cladding Temperature

PS Primary System

PWR Pressurized Water Reactor

PZR Pressurizer

RC Reflux Condensation

RCL Reactor Coolant Line

RCP Reactor Coolant Pump

RPV Reactor Pressure Vessel

SAMG Severe Accident Management Guidelines

SBLOCA Small Break Loss Of Coolant Accident

SEPD Separator

SG Steam Generator

SOT Start Of Test

SS Steady State

TC Thermocouple

UH Upper Head

UP Upper Plenum

xiii

ACKNOWLEDGEMENTS

This work was partly funded by the Swiss Federal Nuclear Safety Inspectorate ENSI 

(Eidgenössisches Nuklearsicherheitsinspektorat) and the Swiss Federal Office of Energy 

(Bundesamt für Energie), and the Spanish Nuclear Regulatory Body (Consejo de Seguridad 

Nuclear).

This report contains findings that were produced within the OECD-PKL2 Project. The authors 

are grateful to the participants and the Management Board of the OECD-PKL2 Project for their 

consent to this publication.

1

1 INTRODUCTION

The aim of this work is to evaluate the capability of the US-NRC thermal hydraulic best estimate 

system code TRACE [4] in properly simulating a SBLOCA in the hot leg of a PWR integral test 

facility, with special focus on the prediction of the Core Exit Temperature (CET). In various 

countries, the CET reading is used as a criterion for the initiation of the AM procedures involving 

emergency operation procedures and/or Severe Accident Management Guidelines (SAMG), 

since the CET can provide a timely indication of core heat up. The value of the CET set point to 

be used in the AM procedures may vary among different PWR types, because the performance 

of the CET depends on the installation position of the thermocouples, the radial and axial power 

distribution in the core, the scenario conditions (pressure, break size), and the geometry of the 

upper plenum.

The test OECD-PKL2 G7.1 (Test G7.1) was performed to assess the reliability of the CET 

measurement and its correlation to the Peak Cladding Temperature (PCT), and more generally 

to provide information on the physical phenomena responsible for the CET performance. The 

PKL integral test facility replicates the entire primary system and most of the secondary system 

of a 1300 MWe Pressurized Water Reactor (PWR) with elevations scaled to 1:1 and diameters 

reduced by a factor 12. It models the nuclear power plant on a scale of 1:145. The detailed 

design of the test facility was based to the largest possible extent on specific data of the 

Philippsburg nuclear power plant, unit2, a Siemens 4-loop design [1].

Test G7.1 [2] is a hot leg (HL) SBLOCA scenario concurrent with additional system failures, 

namely the total failure of the High Pressure Safety Injection (HPSI) system with no early 

secondary side cool down. This scenario necessitates Accident Management (AM) measures to 

prevent a core melt down. The secondary depressurization is intended to restore the heat sink 

through a fast reduction of the primary pressure, thus leading to passive Accumulators (Acc) 

injection followed by Low Pressure Safety Injection (LPSI). 

The test OECD-PKL2 G7.1 is a counterpart of the OECD/NEA ROSA-2 Test 3, a HL SBLOCA 

transient conducted at the ROSA/LSTF facility. ROSA is a full-height and 1/48 volumetrically 

scaled test facility operated by the Japan Atomic Energy Agency (JAEA) for system integral 

experiments simulating the thermal-hydraulic responses at full pressure conditions of a 1100 

MWe-class PWR during SBLOCA accidents and other transients [3]. The working primary-side 

pressure of the ROSA facility is the same as in a PWR. The ROSA-2 Test 3 is run in two 

sequences. During the first sequence, a first core uncovery takes place around 80 bars thus 

allowing for a comparison of the CET and PCT at high pressure. Then, after a transition phase 

where direct HPSI injection into the Reactor Pressure Vessel (RPV) is used to replenish the 

water level up to the Steam Generator (SG) inlet plenum, a secondary side cool down is 

conducted further down to below 45 bars which is the PKL test facility pressure range. A second 

core uncovery and core heat up sequence is initiated at this pressure and is followed by a 

secondary side depressurization and Cold Leg (CL) Accs injection. The second phase of the 

ROSA-2 Test 3 can be compared with the conditions of Test G7.1.

The initial conditions of Test G7.1 are reached following a conditioning phase up to the Start Of 

Test (SOT). Before the SOT, the inventory in the reactor coolant system is reduced down to 

bring about a mixture level in the SG inlet chambers thus allowing for stationary reflux 

condenser conditions. At the SOT, an upwardly oriented small break with a size equal to 1.5% 

of the CL cross section is opened on the HL. The heaters in the core simulator generate a 

volumetrically scaled residual heating power equivalent to 1.8% of the reactor total power, and 

the power level is held constant throughout the conditioning phase and the duration of the test.

The four SGs are depressurized through the full opening of two of the four Main Steam Relief 

Control Valves (MS-RCV) as an AM measure as soon as the CET reaches the safety limit value 

set in the AM procedure. The maximum steam flow rate in the PKL facility is limited by nozzles 

of approximately 20 mm orifice diameters, installed in the main steam lines.
2

The CET parameter is used in many countries as one of the main signals to initiate emergency 

operating procedures and to switch to severe accident management procedures. The CET 

measurements are known to have some limitations in detecting core cooling and core uncovery. 

In fact, if the CET reading indicates superheating, it is in all cases with certain time delay and it 

is always lower than the actual maximum cladding temperature. It is proposed to evaluate in this 

study to what extent a best-estimate system code like TRACE is able to capture the time delay 

and the differences between rod surface temperatures and CET readings under core heat up 

conditions.
3

2 OBJECTIVES

The general objectives of the test G7.1 concern the investigation of relevant phenomena 

occurring under the AM conditions as well as the investigation of scaling effects. Thus, the 

important phenomena to simulate in this test are:

• core uncovery due to boil-off with generation of superheated steam

• primary-side pressure behavior before and after occurrence of core uncovery

• relation between the maximum cladding temperature and the CET during the core uncovery 

period

• SG depressurization based on the CET measurement as a core protection action

• interaction between primary pressure and Acc injection

• Accs injection after SG depressurization and influence on core cooling

• loop seal clearing occurring soon after the Acc’s injection and effect on the RPV water 

levels

The calculations are performed using the TRACE code version V5.0rc3 and a PKL facility model 

that has been previously used in the simulation of various tests ([5], [6]). For the present study, 

first the test initial and boundary conditions have been introduced based on the boundary 

descriptions provided in ref. [2] and [7]. The quality of the results obtained is derived from 

comparison with the available experimental data.

5

3 TEST FACILITY DESCRIPTION

3.1 Overview

The PKL test facility replicates the entire primary system and most of the secondary system of a 

1300 MWe 4-loop PWR plant with elevations scaled to 1:1 and diameters reduced by a factor 

12. A layout of the facility where two loops are displayed can be seen in Figure 1. The detailed 

design of the test facility was based to the largest possible extent on specific data of the 

Philippsburg nuclear power plant, unit2. The scaling concept aims at simulating the overall 

thermal-hydraulic behavior of the full-scale power plant. One of the main conditions to meet 

these requirements is to respect a full scale hydrostatic head correspondence with the reference 

PWR, in order to properly capture gravitational and natural circulation (NC) conditions.

The power, volume and cross-sectional area scaling factor is 1:145. A full-scale correspondence 

of the frictional pressure loss distribution in the system is reproduced. The four primary loops 

are simulated independently to enable all possible primary loop coolant asymmetries during 

transient conditions. Full height fuel rods, spacers and internal structures are used in the RPV 

simulator, and the rods in the core are electrically heated. Also a full height U-tube bundle with a 

scaled number of tubes is used to simulate the primary side of each SG.

Figure 1 PKL integral test facility layout [1]

The next subsections provide a brief description of the main components of the PKL test facility.

Further information can be found in ref. [1].
6

3.2 Reactor Coolant System (RCS)

The RCS is mainly composed of the RPV containing a heater bundle simulating the PWR core; 

the downcomer (DC) model; the four loops equipped with a Reactor Circulation Pump (RCP) a

SG and the connecting pipes, mainly the HL, the CL and the pump seal (PS). The pressurizer is 

connected to the HL of the first loop by the surge line.

3.3 Rod Bundle Vessel

The vessel models the Upper Head (UH) plenum, the Upper Plenum (UP) both above and 

below the Reactor Coolant Line (RCL), the reactor core, the reflector gap and the Lower Plenum 

(LP). The maximum allowable working pressure and temperature of this component are 50 bars

and 573 K, respectively.

For the assessment of the CET performance, the vessel is equipped with multiple 

thermocouples (TC) positioned at the core exit, on different radial locations and installation 

configurations, as a replication of a typical PWR instrumentation (TC with and without shield 

tube). See Figure 2 for more details.

Figure 2 Description of the CET measurements [7]

3.4 Upper Head and Upper Plenum

The upper head and upper plenum are modeled at full scale in height and 1:145 in volume with 

respect to the reference PWR. The internals in the UP are simulated by means of seal welded 

tubes and a RPV level detector is installed in the center. The top of the sensor is located at the 

mid-loop level of the RCL.
7

3.5 Upper Head Bypass

The upper head bypass is modeled with four parallel bypass lines associated with their 

respective primary loop to enable asymmetric flow distribution in the RCS; for example a single 

loop operation. The bypass pressure loss is controlled via an orifice in each by-pass line, in 

order to reproduce a core bypass flow equal to 0.5% of the total primary mass flow rate.

3.6 Test Bundle

The PKL simulated core consists of 314 electrically heated fuel rods. The core also contains 26 

control rod guide thimbles. These are held within an assembly by a top end fitting and the 

spacer grids. The fuel rods are arranged into three concentric zones powered independently. 

This enables the simulation of a radial power profile. However, for Test 7.1 all zones are equally 

powered.

Sixteen rods are equipped with chromel-alumel-sheathed thermocouples. The six sheathed 

thermocouples per rod are brazed into slots distributed over the heater rod length. The thermocouples have an outside diameter of about half a millimeter and the measuring junctions are 

insulated from the outer sheath.

9

4 TRACE MODEL

The TRACE nodalization of the PKL test facility used in this study has been converted and 

further developed from a well tested RELAP5 model developed at the Technical University of 

Catalonia [8]. The model used in this work is consistent with the latest configuration Mark III 

(three) of the facility, which was utilized to conduct Test G7.1. A diagram of the nodalization is 

shown in Figure 3. The discretization used in the nodalization consists of 438 hydraulic volumes 

and 1356 heat structures mesh points. All the walls and surfaces are modeled by heat 

structures to take into account any heat loss, or heat exchange between the structures and the 

coolant. The core fuel rods and other heaters (upper head and pressurizer) are modeled by 

means of heat structures, as well. Most auxiliary, safety and control systems are modeled by 

means of FILL and BREAK components.

The reactor pressure vessel is represented by PIPE components. The core region is nodalized 

by a single PIPE with 7 axial cells. The core bypass is modeled by a single PIPE. The DC is 

modeled with two parallel PIPEs in order to allow capturing recirculation flow in the DC (this 

pattern has been observed in some experiments). On the other hand, the downcomer of the 

SGs are simulated with a single PIPE as the required level of detail in the secondary system is 

not as high as in the primary side. The four bypass pipes from the DC to the UH are described

with two different PIPEs. In each PIPE, the flow area is equivalent to the cross section of two 

PKL facility by-passes. The PIPE components reproduce the bypass orifice using a hydraulic 

diameter restriction.

The four PKL primary loops are modeled with PIPE components. Only one PIPE component is 

used to model the U-tube bundle. The hydraulic diameter of this PIPE component is equal to the 

diameter of a single U-tube while a multiplication factor is used to set the right total heat 

exchange area of the U-tubes. The Emergency Core Cooling System (ECCS) of each loop is 

connected to the cold leg. Each ECCS is composed of an Acc, a HPSI and a LPSI. The Acc is 

represented by a single PIPE component and a valve component. The HPSI and LPSI are 

represented by FILL components and the injection characteristics are imposed as a mass flow 

as function of the primary system pressure. For Test G7.1, only the four Accs and the four LPSI 

systems are available.

The PKL integral test rig pressurizer (PZR) is represented in the TRACE model by a single PIPE 

component. VALVE and BREAK components are used to represent the PZR relief valve and 

relief line boundary conditions. A FILL component is connected at the bottom of the PZR to 

control the water level. The PRZ heaters are linked to the bottom cell of the PZR and are used 

to control the system pressure during the steady state (SS). The PZR is connected to Loop 1 by 

the surge line.

The SG secondary side includes a separator TRACE component (SEPD), a downcomer and 

riser PIPE component and a steam line PIPE component. The four steam lines are connected to 

a common steam header. A constant emergency Feed Water (FW) flow rate is imposed after

the CET limit trip is activated and this flow is kept constant until the end of the transient.

The SG water level during the conditioning phase is controlled using four TRACE FILL 

components and their associated control system. The FILL components are located in

secondary downcomer of the SGs.
10

Figure 3 TRACE nodalization of the PKL test facility (only two loops are shown)
11

5 BOUNDARY CONDITIONS OF TEST G7.1

5.1 Test Procedure

The conditioning phase aims at adjusting the required initial conditions prior to the SOT.

The first phase consists of establishing a primary side full of coolant at a pressure slightly above

40 bars and operating a stable SS under NC conditions. The scaled core power is set to 1.8% 

and the RPV heat losses are compensated. The secondary side initial working conditions are 

around 25 bars.

The second phase consists of bringing the PKL facility to the counterpart OECD/NEA ROSA-2 

Test 3 conditions. First, the steam dump valve is closed to enable a pressurization of the 

secondary side up to more than 40 bars and to decrease the subcooling in the primary side. 

Meanwhile, the break is opened to achieve the desired collapsed levels in the primary side of 

the SG U-tubes. Then, the break is closed to enable SS conditions prior to the SOT. The 

required secondary side pressure is obtained by controlling the operation of the MS-RCVs.

The third phase corresponds to the case study transient of a HL SBLOCA concurrent with

additional failures: total failure of the HPSI combined to no secondary side manual cool down. 

At the end of the conditioning phase the break is opened at the SOT and is kept open until the 

end of the transient phase.

An overview of the test procedure including the conditioning phase is shown in Figure 4.

Figure 4 Test procedure including the conditioning phase [7]
12

5.2 Core Power

The power of the core is 1.8% of the scaled PWR power. In addition, power is generated in the 

system to compensate for the PKL facility heat losses [9]. The core simulator contains 314 

electrically heated elements subdivided into three core zones of 63, 118 and 133 fuel elements 

respectively with a flat axial core power profile. These zones are modeled independently by 

three different heat structures in the TRACE model. However, they are all connected to the 

same hydraulic channel.

5.3 Secondary Pressurization

The first action in bringing the PKL facility to the counterpart OECD/NEA ROSA-2 Test 3 

conditions is the secondary pressurization up to more than 40 bars by closing the steam dump 

valve (VALVE component in the TRACE model). This action is done after reaching the initial 

conditions of a full primary side under SS NC conditions. The secondary pressure is then 

controlled with MS-RCVs. The new SS obtained allows SG reflux condenser (RC) cooling 

conditions.

5.4 Break Conditions

The small break is located on the HL of Loop 1 and is oriented upwards. The size of the break is 

1.5% of the CL cross section area. The break is first used to bring the facility to the specified

initial inventory prior to the SOT. The break is opened for a fixed time duration to bring the 

collapsed level in the SG U-tubes down to the entrance chambers. The HL SBLOCA transient 

itself is started with the break opening after stationary RC conditions have been reached.

5.5 Steam Generator Depressurization

At the end of the conditioning phase, the water level in the SG secondary side is reduced down 

to a value similar to the one obtained in the OECD/NEA ROSA-2 Test 3 (see the SG water 

levels evolution in the bottom part of Figure 5).

The fast cool down AM measure is executed using two of the four MS-RCVs. The maximum 

MS-RCVs steam flow rates are controlled by elliptically shaped nozzles to limit the flow 

according to the specifications of the reference PWR. After the SOT, the secondary 

depressurization is activated by the CET signal when it reaches the safety limit value in the AM 

procedure.

The four SGs of the facility are connected via the main steam header. The steam lines are 

controlled by 4 VALVE components. Since only two of the four MS-RCVs are operated during 

the test, the four valves are kept open in order to balance the pressures in the four SGs during 

the initial phase, the conditioning phase and the transient.

During the transient phase, only two of the four available steam generator relief valves are 

available. For this reason, two of the MS-RCVs of the TRACE model are associated to the CET 

trip signal to enable the secondary depressurization as an AM measure, whereas the two other

MS-RCVs are reserved for secondary pressurization pressure control during the conditioning 

phase. This choice is dictated by the different maximum valve opening rates needed for the 

control of the secondary pressure. When controlling the secondary pressure, the valves open 

only to one third of their total flow area with a relatively slow opening rate compared to the 

safety relief function after the CET limit has been reached (then the valves can open much 

faster). This modeling configuration enabled the MS-RCVs results to better reproduce the 

experiment procedure.
13

5.6 Emergency Core Cooling Systems, ECCS

In Test G7.1, all the HPSI systems are assumed to fail. Both the Accs and the LPSI systems 

inject water in the cold legs of the four PKL primary loops. The initial pressure in the Accs is 

above 25 bars. The LPSI can inject water in the system only after the primary pressure has 

fallen well below 10 bars.

15

6 ANALYSIS RESULTS

6.1 Overview

As previously described, Test G7.1 presents three distinct phases: the steady state, the 

conditioning phase and the transient test phase. In Figure 5 the RPV, PZR and SG secondary 

collapsed water levels are shown for the three phases. The evolution of the pressure in the 

primary and secondary circuits is shown in Figure 6 along with indications of the main events 

occurring during the test. The actuation of the break valve is shown in Figure 7 and the 

calculated mass inventory can be seen in Figure 8. The steam and total core mass flows are 

shown in Figure 9 and the mass flow in the loops is presented in Figure 10. These figures 

provide an overview of the actions performed during all the phases. In Table 1, the chronology 

of the main events is shown.

For convenience, the obtained results are presented in five sub-sections according to the most 

relevant situations during the whole test.

0

2

4

6

8

10

12

0 2000 4000 6000 8000 10000

4

5

6

7

8

9

10

11

12

13

RPV & PRZ levels (m)

Conditioning

RPV

PRZ

SS

Test phase

SGs Secondary Levels (m)

Time(s)

 TRACE

 Experimental

SOT

Secondary depressurization

Figure 5 PKL G7.1 evolution of the RPV (blue), PZR (red) and SG (black) levels
16

0 2000 4000 6000 8000 10000

0

10

20

30

40

50

60

Pressure (bar)

Time (s)

 Primary Experimental

 Primary TRACE

 secondary Experimental

 secondary TRACE

Steam dump closure

Break opening SOT

Core boiloff

AM measure:

Secondary depressurization

ACCs injection

Loop seals clearance

LPSIs injection

Primary flow establishment

Figure 6 PKL G7.1 Test pressures evolution

0 2000 4000 6000 8000 10000

0.0

0.5

1.0

1.5

2.0

2.5

Break Mass Flow Rate (kg/s)

Time (s)

 TRACE

 Experimental

Conditioning primary level

SOT

Low void

High void

Loop seals clearing

Figure 7 PKL G7.1 small break flow rate evolution
17

0 2000 4000 6000 8000 10000

0

500

1000

1500

2000

2500

Break high void mixture flow

Primary Mass Inventory (kg)

Time(s)

PKL initial mass inventory

PKL level prior to SOT

SOT

Break low void mixture flow Start of Accs

 injection

Figure 8 PKL G7.1 mass inventory evolution in the TRACE calculation

0 2000 4000 6000 8000 10000

0

5

10

15

Core Mass Flows (kg/s)

Time(s)

Gas

Total

Liquid

Liquid deficiency&

core boil-off:

6870-7450s

Figure 9 PKL G7.1 core mass flows evolution as calculated by TRACE
18

Table 1 PKL G7.1 test - sequence of events

(Deviation from experiment is indicated for relevant events)

Accident event Calculation (s) Experiment (s)

SS ends (Steam dump closure) 991 528

Mass inventory reduction starts 1591 -

Mass inventory reduction ends 3012 -

SG level reduction starts 4915 4915

SG level reduction ends 5231 5231

SOT (break opens) 6000 6000

Low void break flow (end off-take) 6320 6177

Start of core uncovery 7050 6990

Core boil off (p-prim < p-sec) 7150 (-0.1 bar) 7010

CET limit is reached 7380 7302

Secondary depressurization 7385 (-1.2 bar) 7317

PCT peak maximum value 7410 (-83 K) 7389

CET peak maximum value 7420 (-43 K) 7457

Core vg increase/early mitigation 7400 (0.7 to 3.0 

m/s)

-

Accs injections start 7490 7483

CET peak end 7450 7520

Minimum core level 7560 (-1.3 m) 7503

End of void break flow 7570 -

Accs injection, end 8020 7842

LPSI 8030 7866

Loop seal clearing, end 9221 (-0.9 bar) 9022

Primary flow establishment 9431 (+0.8 bar) 9635
19

6.2 Situation at the Beginning of the Conditioning Phase

The conditioning phase starts after the TRACE model has reproduced the PKL facility initial 

conditions the experiment. The four loops are totally filled with primary coolant and the core 

residual power is evacuated to the secondary system through NC. The electrically heated rod 

bundle in the RPV is supplied with a constant power. Compensation to the PRZ and UH heat 

losses is also locally supplied.

Figure 10 PKL G7.1 loops flow rate evolution

The main steam dump is open and the water levels in the PRZ and SG are stable. The loops 

primary flow rates are in the order of 1 kg/s. The pump impeller rotational speed is null. Table 2 

summarizes the main obtained results. These are in close agreement with the experiment 

except for a slightly higher CET.

Table 2 Results at the beginning of the conditioning phase

Primary side Deviation from 

experiment

Core power 0 kW

CET +3.2 K

Subcooling at core outlet -2 K

Pressurizer level -0.3 m

Primary pressure -0.2 bar

Loop flow rate 0 kg/s

Secondary side

Main steam pressure +1 bar

Secondary temperature -0.5 K

SG water level 0 m

FW temperature +4 K
20

6.3 Initial Test Conditions

The initial test conditions are reached after the pressurization of the secondary side is 

completed (Figure 6). The collapsed water level in the primary system is reduced by temporarily 

opening the break (Figure 7) down to the intended primary coolant mass inventory at the SOT

(Figure 8). The water level in the secondary side of the SGs is also brought down (Figure 5) to 

match initial conditions consistent with the ones of the OECD/NEA ROSA-2 Test 3. The 

obtained SS conditions prior to the SOT (Table 3) correspond to stable RC conditions as 

displayed in Figure 11. As shown in Table 3, the SOT conditions were close to the experiment.

Table 3 Results at the start of the test

Primary side Deviation from 

experiment

Power 0 kW

Pressure -0.7 bar

CET -1.3 K

Loop seal flow rate 0 kg/s

Secondary side

SG pressure -0.9 bar

SG water level 0 m

Temperature -2 K

Feed water temperature 0 K

psec ≈ 45 bars psec ≈ 45 bars psec ≈ 45 bars psec ≈ 45 bars

pprim ≈ 45+ bars

TCore out ≈ 530+ K

Pprim ≈ 1.8%

Figure 11 Initial test conditions for Test G7.1
21

6.4 Situation Prior to Secondary-Side Depressurization

The test is started by opening the break valve. The entire duration of the conditioning phase 

was 4000 s.

One of the specifics of this test is the particularly small size of the break. This leads to a 

relatively higher steam generation rate in the reactor core compared to the steam flow rate 

pulled through the break prior to core uncovery (Figure 7 and Figure 8). This keeps the primary 

pressure to a value close to the secondary pressure thus establishing a heat sink to the 

secondary system through RC. The residual core power is partly evacuated in the form of steam 

flow at the break and the other part is transferred to the SGs. This enables the primary pressure 

to be governed by the secondary system until the core starts boiling-off (Figure 6).

At the start of the test transient phase, the liquid coolant phase dominates the break leak flow 

rate. Phase separation in the HL connected to the upward oriented break (controlled by the offtake model in TRACE) essentially sets the duration of this phase. Figure 12 shows the total and 

steam flows at the break. More power of the generated steam rate is transferred to the heat sink 

as less steam is pulled through the break. 

The next sequence corresponds to the break leak flow rate dominated by vapor phase as the 

liquid level collapses due to the continuous decrease of the mass inventory (no HPSI is 

available). The choked flow conditions at the break govern the duration of this sequence. The 

steam generated in the core is removed through the break steam leak flow rate and the heat 

sink RC conditions. The secondary pressure evolution gets more stable and the pressure 

evolution is fairly constant up to the secondary depressurization. The primary pressure stays 

above the secondary pressure as long as the steam mass flow rate at the break is not sufficient 

to overcome the steam generation rate in the core.

Figure 12 PKL G7.1 break gas and liquid mass flows evolution

As a result of the continuous reduction of the primary mass inventory, the core uncovery 

eventually starts as a result of the decrease of the water level in the core. The dried upper core 

structures heats up and the corresponding power is no longer used for steam generation, 
22

instead a growing part of core power is used to superheat the generated steam. The 

corresponding loss of steam generation is not compensated by the volumetric expansion of the 

superheated steam and this affects the pressure equilibrium, thus causing the primary pressure

to drop below the secondary pressure following (Figure 6). Further core uncovery leads to more 

structure heat up and brings about an excursion of the PCT high enough to produce more 

superheated steam and causing in the end the CET to reach the safety limit (Figure 13). As an 

immediate consequence of the AM procedure, the two available MS-RCVs are fully opened and 

initiate the secondary side depressurization (Figure 5). The total steam mass flow rate in the 

operated MS-RCVs is shown in Figure 14, along with the previous flow rates during the 

controlled conditioning phase.

Figure 13 PKL G7.1 PCT & CET evolution

Figure 14 PKL G7.1 MS-RCVs steam flow rate evolution
23

As can be seen in Figure 6, experimental data show a nearly horizontal time evolution of the 

secondary pressure compared to the slightly more inclined pressure predicted by from 6000 s to 

7000 s. This might be an effect of the secondary heat losses compensation system in the PKL 

facility, which injects water through a special bypass and is suppressed at the start of the 

secondary depressurization event. The TRACE model does not introduce this power 

compensation due to lack of detailed information on this system. Actually, an effort in minimizing 

the simulation secondary heat losses helped shaping the primary pressure curve closer to the 

experimental one after the SOT, but also in the period from 7000 s to 7500 s, up to the initiation 

of the secondary depressurization.

After the secondary depressurization has been initiated, the Accs injection (Figure 15 for the 

pressure and Figure 16 for the mass flow rate) is affected by the slight discrepancy in the 

primary pressure evolution. The injection phase is longer in the calculation and the final 

pressure in the Accs is lower compared to the experimental data.

Table 4 presents the main TRACE calculation results compared to the experimental data.

Table 4 Results prior to the secondary side depressurization phase

Primary side Deviation from 

experiment

Power 0 kW

Pressure -0.4 bar

CET peak +14.6 K

Loop seal flow rate 0 kg/s

Secondary side

Pressure 0 bar

SG water level 0 m

6.5 Acc Injection

The obtained Accs pressure at the end of injection is lower than in the experiment. During the 

injection phase, the behavior of the Accs is affected by the steam condensation rate in the cold 

leg, and by the expansion of the nitrogen dome (the Accs are initially pressurized with nitrogen). 

The Accs reach the end point of operation when the liquid level has been depleted down to a 

specified value. Then the Accs valves close and the pressure remains equal to the 

corresponding primary pressure at the end of injection. In Figure 16, the injected mass flow is 

more irregular than in the experiment. This may hint at the need to improve the condensation 

heat and mass transfer models used in TRACE under Accs injection conditions. Moreover, the 

minimization of the heat losses mentioned in the previous sub-section partly contributed to the 

longer Acc injection time predicted by the TRACE model.
24

Figure 15 PKL G7.1 Accs pressure evolution

Figure 16 PKL G7.1 Accs injection evolution

The simulated loop seal clearing sequence takes place over one minute after the Accs injection 

have started and is followed by a sudden core level increase through an agitated phase. 

However, the TRACE simulation anticipates the mitigation of the CET peak over one minute 

before the start of the Accs injection (see Figure 13 where the dotted vertical line indicates the 

time of CET inversion in TRACE, which is sharp and happens more than a minute before Accs

injection). This discrepancy can be explained by a different steam generation rate in the lower 

part of the core, resulting from the fast depressurization in the primary system. Steam

production reduces the average coolant density in the lower part of the core while raising the 

mixture level higher up, which provides an earlier mitigation of the heat up in the hottest section 

of the core. Figures 17 and 18 show for each core elevation the void fraction and the density,

respectively.
25

Figure 17 PKL G7.1 core void fraction evolution (calculated by TRACE)

Figure 18 PKL G7.1 core density evolution (calculated by TRACE)

The early mitigation phenomenon can also be explained by looking at the calculation results in 

Figure 19 where the core gas and liquid velocities are shown. A sudden increase in the core exit 

vapor flow can be clearly observed after 7400 s, when the core exit vapor velocity increases by 

a factor 4. These higher steam velocities can also explain for a higher entrainment of water up 

into the core. This is not observed in the experimental data, where the mitigation of the CET and 

PCT excursions is not as immediate as in the TRACE simulation. This different behavior results 

in a relatively larger and flatter maximum experimental peaks compared to the sharp edged 
26

calculated peaks (Figure 13). Yet, these more rounded shapes of the experimental temperature 

peaks could also be seen as an early mitigation effect prior to Accs injection and only due to the 

secondary depressurization effect, although not to the same extent as in the calculation results.

Early core mitigation: 7400 s

Figure 19 PKL G7.1 core gas and liquid velocities evolution (calculated by TRACE)

6.6 Situation After Acc Injection

The core has been successfully reflooded with the injection of the Accs and the LS clearance, 

while the secondary depressurization has been completed. As the MS-SRVs remain open, the 

primary system keeps depressurizing following the secondary system and the LPSI set-point is 

reached. The simulated LPSI injection is compared to the experimental data in Figure 20. The 

LPSI injection successfully replenishes the primary system and NC conditions are met again as 

shown in Figure 10.

Figure 20 PKL G7.1 MS-LPSI injections evolution
27

7 CONCLUSIONS

A post-test analysis using TRACE of the test OECD-PKL2 G7.1 has been conducted. The test 

was based on a 1.5% SBLOCA scenario without HPSI and without operator action to initiate the 

secondary depressurization. The break was located on the top of the HL and vertically oriented.

The PKL TRACE model, using the version V5.0rc3 of the code, could reproduce well the main 

physical phenomena involved in Test G7.1, and in good agreement with the experimental data. 

The calculation showed however some discrepancies with the experimental data, such as in the 

shapes of the CET and PCT time evolutions, the Accs injection and the phase separation 

upstream of the break (off-take model), the overall pressure evolution and the progression of the 

different phenomena could be captured with a good level of accuracy. Another important 

conclusion is that the PKL TRACE model could replicate both qualitatively and quantitatively the 

effectiveness of the secondary side depressurization as AM measure to mitigate the core heatup after boil-off.

Finally, in respect to the performance of the TRACE models in simulating the important 

phenomena exhibited by Test G7.1, one can draw the following conclusions:

• The break model could accurately capture the early maximum liquid mass flow rate at the 

time of the SOT, while a slight over-prediction of the break mass flow was observed at low 

void. Very good agreement was obtained at high void.

• The primary-side and secondary-side pressure behaviors could be particularly well 

reproduced. To allow for such results, slight adaptations to the heat losses of the system 

had to be made (within the bounds of the specifications from the test facility). 

• The time evolution of the water level in the core after the SOT could be captured with good 

accuracy, and throughout the defining sequences of the transient, such as the 

depressurization phase, the Accs injection and the loop seal clearance (e.g. the time of 

core uncovery could be captured within a minute accuracy).

• During the core uncovery period, the maximum values of the CET and PCT calculated by 

the TRACE model were relatively lower compared to the corresponding experimental 

values, as a result of an early mitigation of the core heat up, prior to the Accs injection. This

early mitigation effect as a direct consequence of secondary depressurization will need to 

be further evaluated and is possible confirmed by more related experimental data.

• During the Accs injection phase, the injected mass flow proved less regular than in the 

experiment. This may hint at the need to improve the condensation heat and mass transfer 

models used in TRACE under Accs injection conditions, and more sensitivity studies should 

be conducted to better characterize the issue.

29

8 REFERENCES

1. H. Kremin, H. Limprecht, R. Guneysu, K. Umminger, Description of the PKL III test 

facility, FANP NT31/01/e30, Technical center of Framatome ANP, Erlangen, Germany, 

July 2001.

2. S. Bernhard, S. P. Schollengerger and K. Umminger. PKL III G7.1: SBLOCA with total 

failure of HPSI (Counterpart testing with ROSA/LSTF) Quick look report. PTCTPG/2011/en/0008. March 2012.

3. The ROSA-V Group. ROSA-V Large Scale test facility (LSTF) system description for the 

third and fourth simulated fuel assemblies. Technical Report JAERI-Tech 2003-037, 

Japan Atomic Energy Agency, 2003.

4. TRACE V5.0 user’s manual.

5. J. Freixa, A. Manera and F. Reventós. TRACE and RELAP5 Thermal-Hydraulic Analysis 

on Boron Dilution Tests at the PKL Facility. In proceedings of the NURETH-13 

conference. Kanazawa city, Ishikawa, Japan. October 2009.

6. J. Freixa and A. Manera. PSI contribution on the PKL G3.1 semi-blind calculation. Paul 

Scherrer Institut. OECD/NEA PKL-2 4th PRG meeting, Erlangen, Germany, October 

2009

7. B. Schoen, PKL Test G7.1 SB-LOCA with Total Failure of HPSI Counterpart Testing with 

ROSA/LSTF. AREVA NP. Presentation at the OECD/NEA PKL 2 PRG/MB 8th meeting. 

Erlangen, November 8-9, 2011.

8. J. Freixa, “SBLOCA with Boron Dilution in Pressurized Water Reactors, Impact to the 

Operation and Safety”, Technical University of Catalonia, PhD Thesis, 2008.

9. Determination of heat losses in PKL III test facility, NTT1-G/2006/en/0067






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