NUREG/IA-0469 

Development of a Coupled 

TRACE/PARCS Model for KKL and 

Benchmark Against the Turbine Trip 

Test

Prepared by:

P. Papadopoulos, A. Sekhri

Kernkraftwerk Leibstadt AG

CH-5325 Leibstadt, 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: June 2016

Date Published: April 2017

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

NUREG/IA-0469

Development of a Coupled 

TRACE/PARCS Model for KKL and 

Benchmark Against the Turbine Trip 

Test

Prepared by:

P. Papadopoulos, A. Sekhri

Kernkraftwerk Leibstadt AG

CH-5325 Leibstadt, 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: June 2016 

Date Published: April 2017

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 Nuclear Power Plant of Leibstadt (KKL) is a participating member of the Code Applications and 

Maintenance Program (CAMP) of the U.S. Nuclear Regulatory Commission (USNRC) to validate 

the TRACE code for BWR/6 transient analysis. The application of TRACE for the safety assessment 

of BWRs requires a throughout verification and validation using experimental data from tests but 

also plant data for the modelling. The purpose of this work is the review of the KKL TRACE/PARCS 

model, the benchmark of the model against plant data recorded during a turbine trip test and an 

investigation of the core lumping effect on the turbine trip test.

A coupled TRACE/PARCS model has been developed to analyze fast transients in KKL. The first 

benchmark against a turbine trip test has shown differences between the test data and the results 

predicted by TRACE/PARCS such as the total core power and the dome pressure. This is mainly 

due to unstable steady-state conditions during the initialization process and modelling issues. The 

improvements introduced in this work to the TRACE model are including but not limited to the 

geometry of the reactor internals, the redesign of the main steam lines and the implementation of a 

rudimentary control system. Furthermore, the PARCS input model has been updated with the 

turbine trip test corresponding cross sections. The new designed coupled TRACE/PARCS model 

was eventually benchmarked against the same turbine trip plant data.

v 

TABLE OF CONTENTS

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

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

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

EXECUTIVE SUMMARY .....................................................................................................................xi

ACKNOWLEDGMENTS....................................................................................................................xiii

ABBREVIATIONS...............................................................................................................................xv

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

2 PLANT DESCRIPTION...................................................................................................................... 3

3 TRACE/PARCS INPUT MODEL DESCRIPTION............................................................................. 5

3.1 Geometry Review of the TRACE Model ......................................................................................6

3.2 Control Design ..............................................................................................................................9

3.2.1 Feedwater Controller...........................................................................................................9

3.2.2 Implementation of a Pressure CSS.................................................................................. 10

3.3 PARCS Neutronics Data.............................................................................................................10

3.4 Final TRACE/PARCS Model......................................................................................................12

4 BENCHMARK AGAINST TURBINE TRIP PLANT DATA .............................................................15

4.1 Steady-State Calculations ..........................................................................................................15

4.2 Sequence of Event .....................................................................................................................16

5 RESULTS.........................................................................................................................................21

5.1 Lumped Core Model...................................................................................................................21

5.2 Full Core Model...........................................................................................................................26

5.3 Difference Between Full Core and Lumped Core Model...........................................................31

6 RUN STATISTICS............................................................................................................................33

7 CONCLUSION ................................................................................................................................. 35

8 REFERENCES................................................................................................................................. 37

vii

LIST OF FIGURES

Figure 1 Original TRAC-BF1 nodalization scheme......................................................................5

Figure 2 Initial KKL TRACE model converted from TRAC-BF1 – SNAP [7] 

hydraulics component view...........................................................................................6

Figure 3 Free volume inside the RPV [6].....................................................................................7

Figure 4 Volume inside the reactor internals ...............................................................................7

Figure 5 Main steam line geometry – SNAP hydraulics component view ...................................8

Figure 6 Bypass line geometry – SNAP hydraulics component view..........................................8

Figure 7 Level measurement instrumentation .............................................................................9

Figure 8 Axial power distribution in SIMULATE3 and PARCS..................................................11

Figure 9 Control rod pattern in SNAP........................................................................................12

Figure 10 KL TRACE model with modified nodalization scheme – SNAP 

hydraulics component view.........................................................................................13

Figure 11 Core Mapping between TRACE and PARCS .............................................................14

Figure 12 Closing sequence of the flow control valves ...............................................................17

Figure 13 Closing sequence of the turbine control valves...........................................................18

Figure 14 Opening of the bypass valves .....................................................................................19

Figure 15 Core power behavior during the turbine trip test – lumped core model.......................21

Figure 16 Dome pressure evolution during the turbine trip test – lumped core model ................22

Figure 17 Main steam flow rate during the turbine trip test – lumped core model.......................23

Figure 18 Core flow rate during the turbine trip test – lumped core model..................................23

Figure 19 Recirculation flow rate during the turbine trip test – lumped core model.....................24

Figure 20 Jet pump flow ratio performance during the turbine trip test – lumped 

core model ..................................................................................................................25

Figure 21 Water level determination during the turbine trip test – lumped core model ...............25

Figure 22 Core power behavior during the turbine trip test – full core model..............................27

Figure 23 Dome pressure evolution during the turbine trip test – full core model .......................27

Figure 24 Main steam flow rate during the turbine trip test – full core model ..............................28

Figure 25 Recirculation flow rate during the turbine trip test – full core model............................29

Figure 26 Core flow rate during the turbine trip test – full core model.........................................29

Figure 27 Jet pump flow ratio performance during the turbine trip test – full core model............30

Figure 28 Water level determination during the turbine trip test – full core model.......................31

Figure 29 Pressure drop lower plenum – mixing plenum during the turbine trip test...................32

ix

LIST OF TABLES

Table 1 keff comparison between SIMULATE3 and PARCS ........................................................ 10

Table 2 Initial conditions of the lumped core model before the turbine trip test............................ 15

Table 3 Initial conditions of the full core model before the turbine trip test................................... 16

Table 4 Pressure drop inside the RPV ......................................................................................... 16

Table 5 Sequence of event for the turbine trip test....................................................................... 17

Table 6 Final conditions after the turbine trip test for the lumped core model.............................. 26

Table 7 Final conditions after the turbine trip test for the full core model ..................................... 31

Table 8 Run statistics for the lumped core model ........................................................................ 33

Table 9 Run statistics for the full core model................................................................................ 34

xi

EXECUTIVE SUMMARY

The Nuclear Power Plant of Leibstadt (KKL) is a participating member of the Code Applications and 

Maintenance Program (CAMP) of the U.S. Nuclear Regulatory Commission (USNRC) to validate 

the TRACE code for BWR/6 transient analysis. The application of TRACE for the safety assessment 

of BWRs requires a throughout verification and validation using experimental data from tests but 

also plant data for the modelling. The purpose of this work is the review of the KKL TRACE/PARCS 

model, the benchmark of the model against plant data recorded during a turbine trip test and an 

investigation of the core lumping effect on the turbine trip test.

A coupled TRACE/PARCS model has been developed to analyze fast transients in KKL. The first 

benchmark against a turbine trip test has shown differences between the test data and the results 

predicted by TRACE/PARCS such as the total core power and the dome pressure. This is mainly 

due to unstable steady-state conditions during the initialization process and modelling issues. The 

improvements introduced in this work to the TRACE model are including but not limited to the 

geometry of the reactor internals, the redesign of the main steam lines and the implementation of a 

rudimentary control system. Furthermore, the PARCS input model has been updated with the 

turbine trip test corresponding cross sections. The new designed coupled TRACE/PARCS model 

was eventually benchmarked against the same turbine trip plant data. 

The comparison between the results of the TRACE/PARCS simulation and the recorded data of the 

turbine trip test has shown a very good agreement. The TRACE/PARCS model of KKL could 

reproduce the turbine trip sequence with a very good accuracy. Especially the power oscillations, 

indicating instable conditions in the core, could be eliminated. The only big discrepancies left result 

from the valve modeling in TRACE. The non-linear behavior of the closing and opening could not 

be reproduced so far and would need more investigation in the valve design.

Furthermore, the effect of the core lumping and mapping has been studied on the same turbine trip 

test. The same boundary conditions were used and revealed only small differences which mainly 

result from the different core pressure drop. As long as spatial radial effects are not of major 

importance, the lumped core can be used for a first estimate of the output.

In general, the new coupled TRACE/PARCS model performs very well. There are though, as the 

outlook states, further investigations needed in the design of the recirculation lines, including the jet 

pumps, the water level measurement and the valve design. 

xiii

ACKNOWLEDGMENTS

The authors acknowledge the support for the PARCS input data from the Universitat Politécnica de 

València, namely P. Hidalga, R. Miro, T. Barrachina, P. Hidalga and G. Verdú. Special thanks go 

also to F. Giust from Axpo Power AG who helped preparing the CASMO input deck for the specific 

cycle of the turbine trip for the further process into the PARCS NEMTAB files. 

xv

ABBREVIATIONS

BPV bypass valve

BWR boiling water reactor

CPU central processing unit

CSS constrained steady-state

ECCS emergency core cooling system

FCV flow control valve

HPCS high-pressure core spray

KKL nuclear power plant Leibstadt (in German: Kernkraftwerk Leibstadt)

LOCA loss of coolant accident

MSL main steam line

MSIV main steam isolation valve

NK neutron kinetics

NPP nuclear power plant

PARCS Purdue Advanced Reactor Core Simulator

RCIC reactor core isolation cooling

RCS reactor coolant system

RLS reload licensing

RPV reactor pressure vessel

SNAP Symbolic Nuclear Analysis Package

SRI selected rod insertion

SRV safety relief valve

TCV turbine control valve

TH thermal-hydraulic

TT turbine trip

TRACE TRAC/RELAP Advanced Computational Engine

1 

1 INTRODUCTION

System codes are widely used in nowadays nuclear industry for the evaluation of occurrences 

and incidents. The transient analysis with simulation tools is a very important subject as real 

experiments are very expensive and mostly not possible to perform without damaging the 

infrastructure and the integrity of the power plant.

The TRAC/RELAP Advanced Computational Engine (TRACE) is a modernized thermalhydraulics system code designed to merge the capabilities of the legacy codes TRAC-P, TRACB and RELAP. The code is used to analyze different system transients and accident scenarios 

in pressurized (PWR) and boiling water reactors (BWR). TRACE has the capability to model 

thermal-hydraulic phenomena in the one- and three-dimensional space. The Purdue Advanced 

Reactor Core Simulator (PARCS) on the other hand is a computer code that solves the timedependent two-group neutron diffusion equation in three-dimensional Cartesian geometry. The 

code is more commonly used in the analysis of reactivity-initiated accidents in light water reactors 

where spatial effects play an important role. Both system codes may be run in stand-alone or in

coupled mode depending on the type of transient that is being investigated.

In coupled mode, they allow the user to design a model of a nuclear reactor and take all thermalhydraulic and neutronic effects into account. The setup of such a model for the safety analysis is 

a highly time consuming process that is driven by an iterative procedure between modelling and 

validation. Each modification that is applied on an existing coupled model needs verification with 

multiple experimental data sets if possible. The result of these efforts is a multi-disciplinary model 

that is able to predict design and beyond-design transients. The analysis and study of those is 

an important part of the reload licensing. Despite the time and efforts that a numerically robust 

model needs, different accident scenarios have been studied with the coupled system code 

TRACE/PARCS and been published already all over the world. Depending on the type of 

accident, either the thermal-hydraulic or the neutronics part is investigated more thoroughly.

TRACE has for example been used to analyze the spurious opening of the valves of the automatic 

depressurization system (ADS) in a BWR/6 [1]. The safety relief valves which are responsible for 

the ADS, accidently opened during normal operation in a BWR/6. The recorded plant data is 

compared with a TRAC-BF1 and a TRACE simulation. Both models were able to reproduce the 

plant data of the incident. For certain main characteristic parameters like the dome pressure, the 

TRACE code was even better performing than the TRAC-BF1 code. In fact, «the TRACE code can 

satisfactorily predict the system behavior during the ADS transient in a BWR/6» [1].

Beside pure thermal-hydraulic scenarios, TRACE has also been coupled to study reactivity 

induced occurrences. In [2], a control rod withdrawal transient is simulated with TRACE/PARCS 

with a simplified model of the Cofrentes nuclear power plant core, also BWR/6. The model is 

validated with a TRAC-BF1/VALKIN and SIMULATE model of the same core geometry. As the 

results point out, this fast transient analysis is strongly dependent on the numerical solver and 

the time step size. However, the TRACE/PACRS model managed to show the same steadystate behavior as the other system codes.

These examples give a small overview of the capabilities of the system codes TRACE and 

PARCS. Since U.S. NRC declared TRACE to be their flagship thermal-hydraulic code, the 

Leibstadt Nuclear Power Plant (KKL) started with the design of a TRACE/PARCS model and the 

investigation of different design basis transients. The resulting models of this development are 

used to review and reproduce the outcome of the transient analysis of other institutes. The long 
2 

term aim is the ability to conduct own sensitivity studies on different transients to estimate most 

conservative and limiting state conditions of the reactor design. 
3 

2 PLANT DESCRIPTION

KKL is a GE design BWR/6 with a Mark III containment. The plant has been in commercial operation 

since 1984. The initial power rating of KKL was 3012 MWth which was increased in several power 

uprates and modernization steps to a nuclear thermal power of 3600 MWth. The reactor coolant 

system (RCS) is arranged by two recirculation loops, each of them containing one recirculation 

pump, feeding in total 20 internal jet pumps inside the RPV. The reactor core itself is composed of 

648 fuel elements and 146 control rods. The water level in the downcomer is measured with two 

different systems, the narrow (NR) and the wide range (WR) instrumentation. After the steam 

separator and steam dryer stage, four main steam lines lead the way to the high pressure turbine 

and the three low pressure turbines. At the exit of the low pressure turbine, the steam is condensed 

to its initial state. The heat sink is provided by the nearby river water which absorbs the heat in a 

passive cooling tower.

5 

3 TRACE/PARCS INPUT MODEL DESCRIPTION

The development of a thermal-hydraulic model for KKL started with the design of a TRACG model

for more sophisticated transient analysis. The model was later on converted into a TRAC-BF1 input 

model [3] to make the comparison of the results with other facilities easier since the TRAC-BF1 

license was publicly available through the code application and maintenance program (CAMP) 

agreement. The TRAC-BF1 input model, on contrary to the TRACG input, could also be used for 

the publication of benchmarks against plant data of existing occurrences. With the release of 

TRACE V5.0 ([8 - 9]), the thermal-hydraulic model was once again adapted into the new input 

syntax. The two models were both benchmarked against KKL plant data of the event “Inadvertent 

Opening of ADS Valves” [4]. The results lead to the conclusion to focus the further development on 

the TRACE input deck rather than the TRAC-BF1 model, also because U.S. NRC declared TRACE 

as their future flagship in terms of thermal-hydraulic system codes.

The original nodalization and geometry of the RPV is depicted in Figure 1. The geometry was not 

changed, neither during the first conversion step from TRACG into TRAC-BF1 nor from TRAC-BF1 

into the new TRACE model. The scheme shows 15 axial levels and four radial nodes, corresponding 

to three inner shroud rings and one downcomer.

STEAM LINES

Level 1

Level 2

Level 3

Level 4

Level 5

Level 6

Level 7

Level 8

Level 9

Level 10

Level 11

Level 12

Level 13

Level 14

Level 15

0.0

Ring 1 Ring 2 Ring 3 Ring 4

RECIRC 

SUCTION

FEEDWATER

Normal Water 

Level

CHANCHAN

CHAN

JETPUMP

SEPARATORSEPARATORSEPARATOR

Upper 

Plenum

Lower

Plenum

HPCS Ring 1 Ring 2 Ring 3

Figure 1 Original TRAC-BF1 nodalization scheme
6 

Each of the inner core rings holds one lumped PIPE component for the guide tubes, one lumped 

CHAN component for the fuel elements and one lumped SEPD component for the steam 

separators. The connections for the two recirculation lines, the HPCS, the feedwater line and the 

steam lines are also shown at the proper axial level. The design of the piping from and to the RPV 

is illustrated with a SNAP [7] screenshot in Figure 2. The feedwater pump is represented by a FILL 

component as are the RCIC and HPCS safety systems. The main steam lines are split into two 

different pipes. The cross section ratio of this two steam lines is three to one. The division into two 

different sized cross section areas gave certain modularity for the transient simulations. With this 

configuration, the failure of one, three or all four main steam pipes or valves could be investigated.

Figure 2 Initial KKL TRACE model converted from TRAC-BF1 – SNAP [7] hydraulics 

component view

In a further step, this KKL TRACE model was coupled for the first time to a generic core

configuration to test the neutronic feedback during a turbine trip test [5]. The conclusion stated a lot 

of space for improvements in both, the thermal-hydraulic and the neutronic input deck.

3.1 Geometry Review of the TRACE Model

The revision of the input deck started with the TRACE part of the input. Plant documentation was 

accumulated to get all the updated information of the reactor vessel, the reactor core and internals. 

These data was condensed and inserted in the KKL TRACE input model. The only components 

that remained unchanged at the end were the two recirculation pumps. In Figure 3, the evolution of 

the free volume inside the vessel is displayed as it was also presented in [6]. From the fourth axial 

level onwards, the TRACE model is no longer over predicting the free volume. Especially for LOCA 

analysis, the correct free volume of the RPV will be essential.
7 

Figure 3 Free volume inside the RPV [6] 

The geometries of the reactor internals underwent also some modifications. As displayed in Figure 

4, except for the jet pumps, all other values could be improved. Especially the error in the core 

bypass was reduced from over 50% to less than 10%. The correct size of the bypass is of high 

importance for the heat losses from the fuel elements and the neutronic coupling, e.g. the 

temperature correction of the moderator. The redesigned jet pumps lost in volumetric but gained in 

geometric accuracy of the contraction ratio. The guide tubes are still overrated due to the different 

lengths of the tubes. The tubes are mounted at the bottom of the reactor pressure vessel, thus their 

length depends on their radial position. This was not taken into account due to the small influence 

of the guide tubes in the whole system dynamics.

Figure 4 Volume inside the reactor internals

0

50

100

150

200

250

300

350

400

1 2 4 7 8 9 11 13 14 15

Volume [m3]

Axial Level [-]

Reference

TRAC-BF1

TRACE

0.00

0.20

0.40

0.60

0.80

1.00

1.20

1.40

1.60

Separators Guide Tubes Jet Pumps Core Bypass

Relative Volume [-]

TRAC-BF1 TRACE
8 

Despite the RPV and the reactor internals, one of the biggest modifications was done at the main 

steam lines. The lumped steam line models were replaced by four appropriate full length sized 

steam lines according to the plant documentation. Also the amount of valves was adjusted on each 

main steam line to the proper number, e.g. the S/R valves or MSIVs. In Figure 5, a SNAP screenshot 

is depicted of the main steam lines, including the position of the S/R valves, the bypass connection 

and the four turbine control valves.

Figure 5 Main steam line geometry – SNAP hydraulics component view

The bypass lines lead to a manifold, the main steam header, as illustrated in Figure 6. This steam 

header supplies all the auxiliary steam demands of the power plant out of which two are of big 

importance for the normal operation. There is the connection to the reheater, which is running during 

normal operation, and the four bypass lines. These lines are used to reject the heat directly to the 

condenser and thereby bypassing the turbines. These lines were all introduced for the first time to 

the TRACE model of KKL.

Figure 6 Bypass line geometry – SNAP hydraulics component view

The adjustments on the geometry of the input model were not enough for a converging steady-state 

calculation. The design of the controls of the plant model needed some additional attention.
9 

3.2 Control Design

The plant balance was eventually achieved using two main control systems to keep the dome 

pressure and the water level stable in the TRACE calculation. The feedwater controller turned out 

to be one of the crucial controllers for the stability at the water-steam phase boundary. The pressure 

in the dome was contributing a lot on the flow stability inside the main steam lines.

3.2.1 Feedwater Controller

The estimation and control of the water level is one of the most challenging tasks inside a BWR 

model. The level was determined with the hydrostatic pressure equation (1), using the parameters 

in Figure 7 for the calculation.

∆pp = ρρgggg(h2 − hll) + ρρllgg(hll − h1) (1)

Beside the pressure difference over the phase boundary, the water and the vapor density (ρg and 

ρl) are needed. Two PIPE components were therefore attached to the RPV at two different heights. 

One pipe is set above, the other one below the regular water level.

Figure 7 Level measurement instrumentation

The downcomer level was further on used as an input for the TRACE level controller (ICBN = 202, 

[9]) along with the feedwater flow, the main steam flow and the desired water level. The output of 
10

this controller is the new feedwater flow rate which serves as an input for the feedwater FILL

component.

3.2.2 Implementation of a Pressure CSS

The first steady-state calculations showed a very sensitive behavior of the steam flow to power 

oscillations. This sensitivity was a result of the fixed turbine pressure at the end of the main steam 

lines. Steam flow waves were propagating through the steam lines back and forward, thus heavily 

influencing the dome pressure. The implementation of a pressure controller at the turbine control 

valve was able to dampen these wave oscillations and to help keeping the dome pressure at a fixed 

value.

The pressure regulators for the turbine control valves were introduced with the constrained steadystate (CSS) option. The CSS calculation runs the model with the given initial conditions until all 

significant parameters reach the desired and initially set convergence criterion, e.g. 1E-4. It also

adds a stated number of controllers, such as pressure or velocity controllers for valve components

which are only active during the steady-state calculation. Starting a transient simulation leads to the 

deactivation of all CSS controllers. The other possibility would be the implementation of the built-in 

pressure controller (ICBN = 20 [9]). Unfortunately, several attemps did not perform as described in 

[9], which probably comes from a programming error.

3.3 PARCS Neutronics Data

The results from [5] were obtained with a generic core configuration. The turbine trip test though 

was performed during cycle 18. In order to reproduce the core behavior properly, an update of the 

cross section data was necessary. The cross section input was generated by Axpo Power AG in 

CASMO-4. The PARCS input deck was created by the Institute for Industrial, Radiophysical and 

Environmental Safety (ISIRYM) of the Universitat Politècnica de València (UPV) with the SIMTAB 

methodology [11]. The CASMO files were used in SIMULATE3 to determine the neutron kinetics. 

In a second step, the SIMULATE3 [12] output was transformed into PACRS NEMTAB files.

The comparison of the keff between the SIMULATE3 and PARCS in Table 1 and Figure 8 and the 

axial power profile showed small deviations between the two codes. The root mean square error for 

the total axial power is 1.03%.

Table 1 keff comparison between SIMULATE3 and PARCS

System Code keff [-] Difference [pcm] 

SIMULATE3 1.00519 - 

PARCS v3.1 1.00251 267.7

The comparison of the radial power profile of the two different codes showed a good agreement for 

a wide range of the core. However, the relative error at the periphery and the rodded section 

increased to above 5%. This effect is still under investigation at the Universitat Politècnica de 

València. Nonetheless, the cross section data were good enough for the use in the KKL 

TRACE/PARCS model development.
11

As mentioned in the plant description, the reactor core has a total number of 146 control rods. The 

assignment of the control rods are depicted in Figure 9 in a SNAP scheme. This control rod pattern 

was used during the turbine trip test in cycle 18.

Figure 8 Axial power distribution in SIMULATE3 and PARCS

0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1 3 5 7 9 11 13 15 17 19 21 23 25 27

Power Distribution [-]

Node [-]

SIMULATE3

PARCS v3.1
12

Figure 9 Control rod pattern in SNAP

3.4 Final TRACE/PARCS Model

With the modifications on the TRACE geometry and the update on the PARCS cross section input deck 

done, the KKL TRACE/PARCS model was ready for the benchmark against the turbine trip test. Figure 

10 shows the new scheme for the KKL TRACE model in SNAP. The four radial rings of the RPV were 

collapsed to one inner core and one downcomer segment. As a consequence, also the separators and 

the guide tubes were condensed to one component. The lumping showed no significant influence on 

the system dynamics since no radial effects were of interest for the fast transient analysis. The biggest 

improvements were as described before the four proper main steam lines.
13

Figure 10 KKL TRACE model with modified nodalization scheme – SNAP hydraulics 

component view

The reactor core is divided in three parts. One outer peripheral ring, an inner hot core and the 

intermediate zone are holding 648 fuel elements. A second model was designed with 648 CHAN 

components where each CHAN represented one single fuel element. Thus, the effect of a lumped 

core during a fast transient could be investigated.
14

Figure 11 Core Mapping between TRACE and PARCS

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32

1 * * * * * * * * * * * * 0 0 0 0 0 0 0 0 * * * * * * * * * * * *

2 * * * * * * * * * * 0 0 0 26 26 26 26 26 26 0 0 0 * * * * * * * * * *

3 * * * * * * * * * 0 0 26 26 25 25 25 25 25 25 26 26 0 0 * * * * * * * * *

4 * * * * * * * 0 0 0 26 26 25 25 25 25 25 25 25 25 25 26 0 0 0 * * * * * * *

5 * * * * * * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 26 26 0 0 * * * * * *

6 * * * * * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 0 * * * * *

7 * * * * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 0 * * * *

8 * * * 0 0 26 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 0 * * *

9 * * * 0 26 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 * * *

10 * * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 0 0 * *

11 * 0 0 26 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 0 *

12 * 0 26 26 25 25 25 25 25 25 25 25 25 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 25 26 0 *

13 0 0 26 25 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 25 0 0

14 0 26 26 25 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 25 26 0

15 0 26 26 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 26 0

16 0 26 26 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 26 26 0

17 0 26 25 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 26 0

18 0 26 26 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 26 0

19 0 26 26 25 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 25 26 0

20 0 0 26 25 25 25 25 25 25 25 25 25 24 24 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 26 0 0

21 * 0 26 25 25 25 25 25 25 25 25 25 25 24 24 24 24 24 24 25 25 25 25 25 25 25 25 25 25 26 0 *

22 * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 0 *

23 * * 0 0 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 0 * *

24 * * * 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 26 0 * * *

25 * * * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 0 0 * * *

26 * * * * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 26 0 0 * * * *

27 * * * * * 0 0 26 25 25 25 25 25 25 25 25 25 25 25 25 25 25 25 26 26 0 0 * * * * *

28 * * * * * * 0 0 25 26 25 25 25 25 25 25 25 25 25 25 25 25 26 26 0 0 * * * * * *

29 * * * * * * * 0 0 0 26 25 25 25 25 25 25 25 25 25 26 26 0 0 0 * * * * * * *

30 * * * * * * * * * 0 0 26 25 25 25 25 25 25 25 26 26 0 0 * * * * * * * * *

31 * * * * * * * * * * 0 0 0 26 26 26 26 26 26 0 0 0 * * * * * * * * * *

32 * * * * * * * * * * * * 0 0 0 0 0 0 0 0 * * * * * * * * * * * *
15

4 BENCHMARK AGAINST TURBINE TRIP PLANT DATA 

The new model needed as a first step to be validated against plant data. The turbine trip test is a 

fast transient during which all reactivity removal mechanisms are acting, i.e. selected control rod 

insertion and core flow reduction. The transient is also acting in the high pressure region where all 

turbine control and bypass valves need to be activated to stabilize the reactor pressure. The turbine 

trip test during cycle 18 was therefore chosen due to the diverse reactor protection measures to 

validate the modified KKL TRACE/PARCS model.

4.1 Steady-State Calculations

In order to start the transient analysis, the TRACE/PARCS model needs first to run two steady-state 

calculations. In the first calculation, the TRACE model runs with no coupling to PARCS to find the 

steady-state conditions of the initialized model. This independent steady-state calculation speeds 

up the process since there is no neutronic feedback during this unstable initial state. The converged 

TRACE input is coupled in a second step to the PARCS input for another steady-state calculation. 

The stable outcome of this run holds the initial conditions prior to the transient. As mentioned before, 

two different models were designed for the TT test benchmark.

The quality of the lumped model design can be seen in Table 2. The deviations of the different 

characteristic parameters between the plant data and the TRACE/PARCS calculation are less than 

1%. The water level, feedwater flow rate, the dome pressure and the steam flow rate are of particular 

interest. These four values are controlled by the level controller and the CSS respectively while the 

other values are basically fixed by the initial conditions like the recirculation flow rate and the core 

power or a result of the steady-state calculation, i.e. the core flow rate.

Table 2 Initial conditions of the lumped core model before the turbine trip test

 Plant Data TRACE/PARCS Relative Error [%] 

Core Power [MW] 3557.63 3540.03 -0.495

Core Flow Rate [kg/s] 10170.88 10239.54 0.675

Dome Pressure [bar] 72.96 72.92 -0.055

Recirculation Flow Rate [kg/s] 3370.85 3391.46 -0.611

Steam Flow Rate [kg/s] 1955.02 1955.57 0.028

Feedwater Flow Rate [kg/s] 1952.99 1951.85 -0.058

Relative Water Level [-] 1.00 0.999 -0.007

The results of the full core model look slightly different. Once more, the only change in this model 

is the replacement of the three lumped channels with 648 channel components. Most of the values 

in Table 3 are still in the same order of magnitude as before except for the core flow rate. The 

relative error for the recirculation flow rate did also slightly increase. The core flow rate is determined 
16

by the two jet pump models that represent all 20 internal jet pumps. The sum of these flow rates 

corresponds to the total core flow. 

Table 3 Initial conditions of the full core model before the turbine trip test

 Plant Data TRACE/PARCS Relative Error [%] 

Core Power [MW] 3557.63 3540.34 -0.486

Core Flow Rate [kg/s] 10170.88 10450.13 2.746

Dome Pressure [bar] 72.96 73.07 0.151

Recirculation Flow Rate [kg/s] 3370.85 3400.54 0.881

Steam Flow Rate [kg/s] 1955.02 1957.06 0.104

Feedwater Flow Rate [kg/s] 1952.99 1953.51 0.027

Relative Water Level [-] 1.00 0.999 -0.007

According to the results in Table 2 and Table 3, there is a clear difference between the two core 

flow rates which cannot be explained only by the changed recirculation flow rates. As mentioned, 

the core flow rate is estimated by the jet pump flow rates. The jet pumps are on the one hand driven 

by the forced flow from the recirculation flow rate but also by the pressure drop alongside the 

downcomer. Since all initial conditions are set exactly the same for both models, the effect has to 

come from the different modeling of the core. In Table 4, the pressure drop between the mixing 

plenum and the lower plenum and the absolute pressure in the upper plenum is shown. 

Table 4 Pressure drop inside the RPV

 Lumped Core Full Core Difference [bar] 

Lower Plenum – Mixing Plenum 

Pressure Drop [bar] 1.38 1.25 0.13

Upper Plenum Pressure [bar] 72.92 73.07 0.15

The full core model has a clear influence in the pressure distribution in the core region. The core 

flow rate is established for each channel individually while the lumped core distributes the coolant 

among three channels. The pressure is hence also affected in the upper plenum and as a 

consequence in the downcomer. This pressure difference will influence the flow conditions in the 

jet pumps. The full core model has therefore already shown an effect on the results of the transient 

analysis. 

4.2 Sequence of Event

As mentioned before, the event is followed by multiple system measures. In Table 5, the four events 

are chronologically summarized and compared with the TRACE/PARCS trip initiation. The turbine 
17

control valves (TCVs) start the fast closing 200 ms after the initiation of the transient. As a 

consequence, the bypass valves (BPVs) open with a delay of 40 ms to reduce the pressure in the 

RPV. The rod insertion is initiated another 60 ms later to compensate the added reactivity due to the 

pressure induced void collapse in the core. The second measure is activated 10 ms after the rod 

insertion which is called recirculation runback. The flow control valves of the recirculation line are 

closed to a predefined position to reduce the core flow and thus also reduce the reactivity in the core.

Table 5 Sequence of event for the turbine trip test

Event Initiation after Trip TRACE/PARCS 

Turbine Control Valve Closing 0.20 0.20

Bypass Valve Opening 0.24 0.24

Selected Rod Insertion 0.30 0.30

Recirculation Runback 0.31 0.20

The delay of the recirculation runback is the only quantity that does not match with the test protocol 

which will be discussed in the results section. The closing sequence of the flow control valve 

illustrated in Figure 12 starts as described in Table 5 100 ms earlier than the reference and reaches 

the final state accordingly faster. The offset of the valve opening ratio at the beginning and the end 

comes from the modeling of the valve. The flow areas of all valves are designed in a way to agree 

with the properties like pressure or mass flow. The offset has only little effect on the behavior of the 

system as long as the closing times are matching.

Figure 12 Closing sequence of the flow control valves 

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0 5 10 15 20

Valve Opening Ratio [-]

Time [s]

Flow Control Valve A

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0 5 10 15 20

Valve Opening Ratio [-]

Time [s]

Flow Control Valve B

TRACE/PARCS

Model

Reference Plant

Data
18

The opening ratio of the TCVs is as described earlier estimated by the CSS controller. The design 

of the valves was analyzed more thoroughly which made it possible to meet the opening ratio with 

the plant data during steady-state. As a result, the flow area did not correspond to the real 

geometrical TCV area but from a modeling point of view, the closing sequence could be represented 

very well except for the final state. The valve signal in the closed state always has an offset by the 

design of the measurement technique. Even for the TCV A which seems to be stuck open, the TCV 

indicator in the control room showed a closed position. Besides the big offset of the TCV A, there is 

also a small deviation of the TCV D at 0.3 seconds with an insignificant effect on the system 

response. 

Figure 13 Closing sequence of the turbine control valves

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

0.0 0.5 1.0

Valve Opening Ratio [-]

Time [s]

Turbine Control Valve A

0.0

0.1

0.2

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0.5

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0.7

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0.9

1.0

0.0 0.5 1.0

Valve Opening Ratio [-]

Time [s]

Turbine Control Valve B

TRACE/PARCS

Model

Reference Plant

Data

0.0

0.1

0.2

0.3

0.4

0.5

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Valve Opening Ratio [-]

Time [s]

Turbine control Valve C

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Valve Opening Ratio

Time [s]

Turbine Control Valve D
19

The first action after the closing of the TCVs is the opening of the bypass valves. In the real plant, 

the bypass is regulated by a pressure controller. The attempt to model a pressure controller with 

the built-in version in TRACE unfortunately failed. The controller does not respond at all which 

probably comes from a programming error. Therefore, the signal was set manually with the help of 

the reference data. After several iterations, a stable configuration was found. In Figure 14, the valve 

opening ratio of the BPVs is pictured. The sequence maxima and minima were adjusted to be 

synchronous with the plant data. The value of the valve opening ratio though varied due to the same 

reason as the flow control valve opening.

Figure 14 Opening of the bypass valves

0.0

0.1

0.2

0.3

0.4

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1.0

0 5 10 15 20

Valve Opening Ratio [-]

Time [s]

Bypass Valve N

0.0

0.1

0.2

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1.0

0 5 10 15 20

Valve Opening Ratio [-]

Time [s]

Bypass Valve N/M

TRACE/PARCS Model

Reference Plant Data

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

0 5 10 15 20

Valve Opening Ratio [-]

Time [s]

Bypass Valve S/M

0.0

0.1

0.2

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0.4

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1.0

0 5 10 15 20

Valve Opening Ratio [-]

Time [s]

Bypass Valve S
20

In general, the model responded as expected with the given initiation sequence. The selected rod 

insertion was modeled in the PARCS part of the input. The different positions of the control rods 

were set in the PARCS input deck as it was documented in the test report.
21

5 RESULTS

The turbine trip test scenario was performed on the two different TRACE/PARCS inputs – the 

lumped core and the full core model. The problem time was set to 20 seconds with a maximal time 

step size of 10 milliseconds. The results were compared with the plant data of the turbine trip. Five 

characteristic parameters were chosen for comparison: core power, dome pressure, core flow, main 

steam flow and the water level. The final conditions of the transient after the 20 seconds are also 

summed up to conclude the results of the calculation.

5.1 Lumped Core Model

The results of the lumped core model shall be discussed first. The evolution of the core power is 

the most difficult value to control due to its strong coupling to many other parameters like the dome 

pressure, core flow or coolant temperature. In Figure 15, the core power during the turbine trip test 

is depicted.

Figure 15 Core power behavior during the turbine trip test – lumped core model

As it can be seen, the deviation of the simulation results from the plant data is during most of the 

transient time very small. An exception is the time period between 1.5 and 6 seconds where 

TRACE/PARCS is clearly overestimating. The power excursion at 5 seconds can be explained with 

the evolution of the dome pressure, illustrated in Figure 16. At 5 seconds, the dome pressure 

experiences a peak caused by the pressure regulator of the bypass valves. This pressure collapses 

the void in the channels and induces a power response. In the plant data, this peak is not measured. 

The explanation can be found in the pressure distribution in the RPV. The dome pressure in the 

model is apparently propagating through the separators to the core while in the real plant a pressure 

wave is dampened by the separators. Therefore, some further improvements would be necessary 

for the reactor internals.

The estimated dome pressure by TRACE in Figure 16 is like the core power deviating only very 

little. As described before, the pressure was controlled manually with the bypass opening so it was 

1000

1500

2000

2500

3000

3500

4000

0 5 10 15 20

Thermal Power [MW]

Time [s]

TRACE/PARCS Model

Reference Plant Data
22

very difficult to keep the dome pressure on a constant value. Besides this drawback, the calculation 

results were very good.

Figure 16 Dome pressure evolution during the turbine trip test – lumped core model

The values of the main steam flow in Figure 17 were the first to show a major discrepancy from the 

reference plant data. While the values for the maxima did match the reference one, the minima 

were always far underestimated. Also the timings of the peaks were always shifted by around 500 

milliseconds. The design of the bypass valves can be one explanation for the different behavior. 

The modeling of the valve characteristic of the bypass valves is a difficult task. During normal 

operation, these valves are kept closed so there is little data available to benchmark the model 

against. This would explain the time shift in the response. The choked flow model could also have 

its contribution to the deviation. Both explanations would need further investigation to improve the 

behavior of the main steam flow.

68

69

70

71

72

73

74

75

0 5 10 15 20

Pressure [bar]

Time [s]

TRACE/PARCS Model

Reference Plant Model
23

Figure 17 Main steam flow rate during the turbine trip test – lumped core model

The evolution of the core flow is shown in Figure 18. As mentioned in the section above, the 

recirculation runback is initiated 100ms earlier. This anticipation allows the TRACE/PARCS model 

to meet the linear part of the core flow drop between 1 and 4 seconds where most of the reactivity 

is decreased. The input was also here not able to predict the non-linear behavior of the real valves. 

Beside these side effects, the core flow follows the plant data with a very good agreement.

Figure 18 Core flow rate during the turbine trip test – lumped core model

The recirculation flow in Figure 19 on the other hand did show a good agreement. The discrepancy 

of the two recirculation lines is by far the biggest of all the presented parameters. As mentioned in 

the section above, the jet pump characteristic was found to be the reason. The flow control valve 

1000

1200

1400

1600

1800

2000

2200

2400

0 5 10 15 20

Mass Flow [kg/s]

Time [s]

TRACE/PARCS Model

Reference Plant Data

6000

6500

7000

7500

8000

8500

9000

9500

10000

10500

0 5 10 15 20

Mass Flow [kg/s]

Time [s]

TRACE/PARCS Model

Reference Plant Data
24

was adjusted in a way to meet the conditions for the core flow rate. As a consequence, the results 

of the recirculation flow rate suffered.

Figure 19 Recirculation flow rate during the turbine trip test – lumped core model

The jet pump flow ratio is pictured in Figure 20. The two different behaviors of the input model and 

the plant data drift clearly apart. The plant data shows first a decrease, followed by a recovery to an 

even higher value than the initial one. The jet pump flow ratio for the TRACE/PARCS model starts 

immediately to increase but stabilizes at about the same time as the reference plant data on a lower 

value. This big discrepancy originates probably from the pressure distribution in the RPV. Compared 

to the real plant, as it could be seen in the core power in Figure 15, the pressure drop from the dome 

to the lower region is too small and the internals respond too sensitive on dome pressure changes. 

Thus, the flow in the jet pump is accelerated by the higher pressure in the downcomer while the 

plant data indicates this response after 4 seconds. Concluding these results, the recirculation lines 

and the jet pumps will need some modeling work in the near future. 

800

900

1000

1100

1200

1300

1400

1500

1600

1700

1800

0 5 10 15 20

Mass Flow [kg/s]

Time [s]

TRACE/PARCS Model Line A

TRACE/PARCS Model Line B

Reference Plant Line A

Reference Plant Line B
25

Figure 20 Jet pump flow ratio performance during the turbine trip test – lumped core 

model

The new level water level instrumentation was also benchmarked against the plant data. In Figure 

21, the comparison of the calculation and the reference is depicted. The level measurement reacts 

very sensitive on the pressure difference which explains the noisy behavior of the estimated water 

level. The preassigned method in (1) to determine the water level is thus only working for the steadystate calculation where the pressure difference is also converging towards a stable value. During a 

transient, where the internal pressure is rapidly changing, a different method has to be found to get 

a reliable water level during transient calculations.

Figure 21 Water level determination during the turbine trip test – lumped core model

1.5

1.6

1.7

1.8

1.9

2.0

2.1

2.2

2.3

2.4

2.5

0 5 10 15 20

Jet Pump Flow Ratio [-]

Time [s]

TRACE/PARCS Model

Reference Plant Data

0.970

0.975

0.980

0.985

0.990

0.995

1.000

1.005

1.010

1.015

1.020

0 5 10 15 20

Relative Water Level [-]

Time [s]

TRACE/PARCS Model

Reference Plant Data
26

The final values of the transient were again summarized and can be found in Table 6. As it can be 

seen, the relative errors increased for some parameters, i.e. core power, recirculation and steam 

flow rate. The core power is a very difficult parameter to control since it is influenced by almost all 

other values. Therefore, a relative error of within 2 % is a good result.

Table 6 Final conditions after the turbine trip test for the lumped core model

 Plant Data TRACE/PARCS Relative Error [%] 

Core Power [MW] 2110.00 2069.19 -1.934

Core Flow [kg/s] 6360.62 6399.39 0.610

Dome Pressure [bar] 68.97 69.24 0.391

Recirculation Flow Rate [kg/s] 1911.15 1996.54 4.468

Steam Flow Rate [kg/s] 1146.94 1096.12 -4.431

Feedwater Flow Rate [kg/s] 1102.78 1103.61 0.075

Relative Water Level [-] 1.002 1.008 0.599

As mentioned above, the recirculation and steam flow rate suffer both of compromises that had to 

be made. The recirculation flow rate had to be increased to keep the core flow at the power plant 

level. Again, the recirculation line needs another revision to widen the working range of the jet 

pumps. The steam flow rate has been already discussed at Figure 17. 

In principal, the lumped core model performed very well, especially comparing with the previous 

state of the model in [4]. There is still some space for improvements, e.g. the recirculation lines, the 

pressure controller for the bypass valves, the water level instrumentation or the pressure distribution 

in the RPV. The TRACE/PARCS model is though able to reach coupled steady-state conditions 

with a good convergence criterion in order to run some transient calculations.

5.2 Full Core Model

The full core model has been setup according to the initial conditions listed in Table 3. The model was 

running the same transient with the same trips, closing and opening times and control rod insertion. 

The aim of this calculation was to investigate the influence of a thermal-hydraulic lumped core. 

The first illustration in Figure 22 shows the core power evolution of the full core model. The 

differences between the core power in Figure 15 and Figure 22 appear to be very small. The 

selected control rod insertion is the only measure that shows a better response at 1.5 seconds than 

in the lumped core case. The fuel elements around the control rods are in this case more sensitive 

to inserted control rods than the lumped fuel element where the effect is averaged over an assembly 

of fuel elements.
27

Figure 22 Core power behavior during the turbine trip test – full core model

The pressure in the dome, pictured in Figure 23, is following the plant data at beginning with a high 

accuracy. The discrepancy between the simulation and the reference data starts to increase after 

3 seconds of problem time. The pressure remains from there on overestimated for the rest of the 

transient. The explanation for this behavior has already been given in the previous section. Since 

the core pressure drop is smaller in the full core model than in the lumped core, it is not surprising 

to see an influence also in the behavior of the dome pressure.

Figure 23 Dome pressure evolution during the turbine trip test – full core model

The main steam flow rate in Figure 24 looks very similar to the flow rate in Figure 17. Also a closer 

look does not reveal any significant difference. The root-mean-square deviation between the results 

1,000

1,500

2,000

2,500

3,000

3,500

4,000

0 5 10 15 20

Thermal Power [MW]

Time [s]

TRACE/PARCS Model

Reference Plant Model

68

69

70

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75

0 5 10 15 20

Pressure [Pa]

Time [s]

TRACE/PARCS Model

Reference Plant Data
28

of the lumped core model and the full core model is less than 1%. It is no coincidence to find basically 

the same results concerning the steam flow rate. The mass flow rate is influenced by the activation of 

the TCVs and BPVs. The steam boiling in the core is compared to the acting pressure waves in the 

main steam lines relatively slow so the mass flow will be determined rather by the valve action than 

the power excursion. The opening respectively closing sequence of the valves did not change for the 

two TRACE/PARCS model so the difference between the two is expected to be small.

Figure 24 Main steam flow rate during the turbine trip test – full core model

The comparison between the recirculation flow rate of the full core model in Figure 25 and the 

lumped core model in Figure 19 shows even fewer differences than the main steam flow rate. The 

root-mean-square error of the two flow rates is actually less than 0.2 %.The similarity is again no 

surprise, since the recirculation pumps and the FCV closing sequences are exactly the same for 

both models, which is eventually the driving force of the recirculation flow rate.

1000

1200

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1600

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2000

2200

2400

0 5 10 15 20

Mass Flow [kg/s]

Time [s]

TRACE/PARCS Model

Reference Plant Data
29

Figure 25 Recirculation flow rate during the turbine trip test – full core model

The core flow rate on the other hand is not only influenced by the forced flow of the recirculation 

lines but also by the pressure difference between the downcomer and the lower plenum. As 

mentioned before, the core pressure drop of the full core model is slightly smaller than in the lumped 

core model. Therefore, the dome pressure was also higher which has a direct impact on the inlet 

pressure of the jet pump. In Figure 26, the total core flow rate is illustrated for the full core model. 

The flow rate is overestimated for most of the time except for the missing non-linearity at around 5 

seconds.

Figure 26 Core flow rate during the turbine trip test – full core model

800

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0 5 10 15 20

Mass Flow [kg/s]

Time [s]

TRACE/PARCS Model Line A

TRACE/PARCS Model Line B

Reference Plant Line A

Reference Plant Line B

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Mass Flow [kg/s]

Time [s]

TRACE/PARCS Model

Reference Plant Data
30

The effect of the pressure on the jet pump is also visible in the jet pump flow ratio in Figure 27. 

Comparing the results with Figure 20 shows the same behavior during the transient but shifted on 

the ordinate. The jet pump flow ratio is in this case around 2 % higher.

Figure 27 Jet pump flow ratio performance during the turbine trip test – full core model

In Figure 28, the evolution of the water level is depicted. Once more, the deviations to the results of 

the lumped core model seem to be very small. The water level is as described above estimated via 

the hydraulic pressure difference at two different points in the downcomer. A higher pressure in the 

dome will increase the pressure everywhere, from the downcomer to the upper plenum. Thus, the 

difference will remain approximately constant and the water level determination will not be strongly 

affected.

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2.0

2.1

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Jet Pump Flow Ratio [-]

Time [s]

TRACE Model

Reference Plant Data
31

Figure 28 Water level determination during the turbine trip test – full core model

Last but not least, the final conditions after the turbine trip test are summarized in Table 7. 

Comparing the relative errors of the simulation with the lumped core model in Table 6, the only 

significant difference can be found for the core power, the dome pressure and the core flow rate. 

As also seen in the results, the core pressure drop affected only these three values.

Table 7 Final conditions after the turbine trip test for the full core model

 Plant Data TRACE/PARCS Relative Error [%] 

Core Power [MW] 2110.00 2061.41 -2.303

Core Flow Rate [kg/s] 6360.62 6491.61 2.059

Dome Pressure [bar] 68.97 69.56 0.855

Recirculation Flow Rate [kg/s] 1911.15 1996.71 4.477

Steam Flow Rate [kg/s] 1146.94 1100.00 -4.093

Feedwater Flow Rate [kg/s] 1102.78 1106.60 0.346

Relative Water Level [-] 1.002 1.009 0.699

5.3 Difference Between Full Core and Lumped Core Model

The differences between the fully mapped core and the lumped three channel model appeared to 

be very small and insignificant. The lumped core seemed even to be slightly more conservative in 

terms of core power and pressure distribution inside the core. The question arises of whether the 

full core model is necessary for the transient analysis or if the lumped core would be sufficient. This 

0.970

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0.995

1.000

1.005

1.010

1.015

1.020

0 5 10 15 20

Relative Water Level [-]

Time [s]

TRACE/PARCS Model

Reference Plant Data
32

question cannot be fully answered with only one transient calculation, but the results give a small 

hint. In Figure 29, the pressure difference between the lower and the mixing plenum is shown. The 

pressure drop for the lumped core stays above the pressure drop of the full core model.

Figure 29 Pressure drop lower plenum – mixing plenum during the turbine trip test

Therefore, the pressure distribution will not change inside the RPV, relatively speaking. This is what 

the results of the full core model have shown as well. The only differences originated in the different 

pressure drop inside the core. Eventually, the deviations remained very small. So in terms of fast 

transients where the neutronic feedback has not too much time to respond, the lumped core model 

can enough for modeling the transient. If more accurate results are needed, e.g. hottest rod position, 

exact pressure distribution in a particular fuel element or if the margins are small, the full core should 

be represented in the TRACE/PARCS model.

0.6

0.7

0.8

0.9

1.0

1.1

1.2

1.3

1.4

0 5 10 15 20

Pressure Drop [bar]

Time [s]

Lumped Core Model

Full Core Model
33

6 RUN STATISTICS

The calculations on the lumped core model with the TRACE/PARCS computer code (TRACE V5.0 

p3, PARCS 2.7) were performed on a work station which is built on an Intel Xeon 5450 CPU (based 

on the Intel Core microarchitecture @ 3.00 GHz), having 4 cores and 4 threads in total. The 

operating system is Windows 7 Service Pack 1 (x86_64) with 16 GB of RAM memory.

Table 8 Run statistics for the lumped core model

Characteristic Value Stand-Alone SteadyState 

Coupled SteadyState Turbine Trip 

Transient 

Problem Time [s] 176.786 78.170 20.002

CPU Time [s] 478.642 157.436 805.717

Number of Time Steps [-] 9505 810 2024

Maximum Time Step Size [s] 1e-2 1e-1 1e-2 

CPU Time / Problem Time [-] 2.707 2.014 40.282

Number of Thermal-Hydraulic Cells [-

] 81

Number of Heat Structures [-] 225

In Table 8, the run statistics for the transient turbine trip benchmark are listed. The total number of 

thermal-hydraulic cells was 81 and 225 heat structures. The transient calculations run due to the 

small time step size 40 times slower than the real time. Stand-Alone steady-state calculations lasted 

176.786 seconds of problem and 478.642 seconds of CPU time, requiring 9505 time steps to reach 

the convergence criterion while the coupled steady-state needed 78.17 seconds of problem time, 

157.436 seconds of real time and 810 steps. 

The full core model was running on a work station with two Intel Xeon E5-2643 CPUs (based on 

the Sandy Bridge microarchitecture @ 3.30 GHz), having 8 cores and 16 threads in total. The 

operating system is Windows 7 Service Pack 1 (x86_64) with 32 GB of RAM memory.

The run statistics for the full core model are presented in Table 9. The problem times remained in 

the same order of magnitude while the CPU times increased significantly. The number of thermalhydraulic cells went up to 17496 and the number of heat structures to 48600. Therefore, more 

computations were needed per time step which increased the CPU time to 20812.748 seconds for 

the stand-alone case, to 5866.339 seconds for the coupled steady-state calculation and 611.453 

for the transient. The worst CPU / problem time ratio turned out to be during the transient calculation 

with 1 second of problem time being more than 10 minutes of CPU time. The number of time steps 

though stayed approximately on the same level for both models.
34

Table 9 Run statistics for the full core model

Characteristic Value Stand-Alone SteadyState 

Coupled SteadyState Turbine Trip 

Transient 

Problem Time [s] 199.886 126.805 20.009

CPU Time [s] 20812.748 5866.339 12234.565

Number of Time Steps [-] 11850 1310 2049

Maximum Time Step Size [s] 1e-2 1e-1 1e-2 

CPU Time / Problem Time [-] 104.123 46.263 611.453

Number of Thermal-Hydraulic Cells [-

] 17496

Number of Heat Structures [-] 48600
35

7 CONCLUSION

This work continued the improvement of a TRACE/PARCS model which has been started some 

time ago ([3 - 5]). The TRACE model has been reviewed with existing technical documentation and 

improved in terms of control systems and cell volume terms and a second model has been 

developed where the full core has been modeled with single fuel element channels. Both TRACE 

models have been coupled with the neutronics system code PARCS and validated with existing 

plant data resulting from a turbine trip test. For the analysis of further fast transients, the validation 

required to be done with another fast transient. The turbine trip is a very favorable transient, since 

all the measures to maintain the reactor in operation are taken, i.e. selected rod insertion, 

recirculation runback and the activation of the turbine bypass.

The results of the benchmarks against the plant data were very promising. The numerical 

instabilities were tremendously reduced and the characteristic parameters, i.e. dome pressure, core 

flow, core power and steam flow, were able to follow and to a certain degree also match the existing 

data. The initial and final conditions are consistent within 5 % accuracy. The discrepancy of the 

valve opening and closing behavior with the real valves could not be resolved yet and will need 

future work. The built-in pressure controller in TRACE for the bypass valves did not perform as 

described in the TRACE manual which made manual adjustments of the bypass opening 

necessary. Last but not least, the jet pump characteristic was not properly modeled and will require 

further research and modification.

The difference of the lumped core and the full core model revealed a different pressure drop inside 

the core. In the full core model, the core inlet flow is established elementwise in each fuel assembly 

and thus also the pressure drop. In the lumped core model, the core flow is distributed within three 

channels which lead to different flow conditions in the core and hence a different pressure drop.

In general, the TRACE/PARCS coupled models showed a very good performance. Not only that 

the results in the benchmark against the turbine trip data were met with a good accuracy, the results 

in the fast transient analysis could also be reasonably explained. First steps have been done in 

understanding the effect of lumping for fast transients and will be expanded in the future work. 

However, the nuclear power plant of Leibstadt is now one step further in their development of a 

TRACE/PARCS model that can already be used for the validation of existing transient analysis.

37

8 REFERENCES

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of the TRACE code”, Nuclear Engineering and Design, 241, 6 (2011), pp. 2240-2447.

2. Miro, R., Verdu, G., Sanchez, A. M., Barrachina, T., Gomez, A., “Analysis of a rod

withdrawal accident in a BWR with the neutronic-thermalhydraulic coupled code TRACBF1/VALKIN and TRACE/PARCS”, Proceedings of American Nuclear Society’s Topical

Meeting of Reactor Physics PHYSOR, Vancouver BC, Canada, September 10 – 14 2006.

3. Coddington, P., Zimmermann, M. A., „TRAC-BF1 Input Model for KKL and Conversion

from TRACG“, TM-41-96-11, Paul Scherrer Institut, July 19 1996.

4. Sekhri, A., Baumann, P., Wicaksono, D., Miró, R., Barrachina, T., Verdú, G., “TRACE and

TRAC-BF1 Benchmark against Leibstadt Plant Data during the Event Inadvertent Opening

of ADS Valves”, Proceedings of the International Congress on Advances in Nuclear Power

Plants, Chicago Illinois, USA, June 24 – 28 2012.

5. Hidalga, P., Abdelkrim, S., Baumann, P., Morera, D., Miró, R., Barrachina, T., Verdú, G.,

“TRACE coupled with PARCS Benchmark against Leibstadt Plant Data during the Turbine

Trip Test”, Proceedings of the International Nuclear Atlantic Conference, Recife, Brazil,

November 24 – 29 2013.

6. Papadopoulos, P., Sekhri, A., Giust, F., Hidalga, P., Miró, R., Barrachina, T., Verdú, G.,

“Assessment of a TRACE/PARCS Benchmark against Leibstadt Plant Data during the

Turbine Trip Test”, Proceedings of the International Topical Meeting on Nuclear Reactor

Thermalhydraulics, Chicago Illinois, USA, August 31 – September 4 2015.

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User’s Manual Version 2.2.1, Bloomsburg PA, U.S.A (2012).

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Physical Models, Office of Nuclear Regulatory Research, Washington, U.S.A. (2013).

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Research, Washington, U.S.A. (2013).

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School of Nuclear Engineering, Purdue University.

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12.02.

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Reactor Analysis Code. Studsvik/SOA-95/15.

Development of a Coupled TRACE/PARCS Model for KKL and Benchmark 

Against the Turbine Trip Test

NUREG/IA-0469

Petros Papadopoulos, Abdelkrim Sekhri

Kernkraftwerk Leibstadt AG

CH-5325 Leibstadt, Switzerland

K.Tien, NRC Project Manager

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

The Nuclear Power Plant of Leibstadt (KKL) is a participating member of the Code Applications and Maintenance Program 

(CAMP) of the U.S. Nuclear Regulatory Commission (USNRC) to validate the TRACE code for BWR/6 transient analysis. 

The application of TRACE for the safety assessment of BWRs requires a throughout verification and validation using 

experimental data from tests but also plant data for the modelling. The purpose of this work is the review of the KKL 

TRACE/PARCS model, the benchmark of the model against plant data recorded during a turbine trip test and an investigation 

of the core lumping effect on the turbine trip test.

A coupled TRACE/PARCS model has been developed to analyze fast transients in KKL. The first benchmark against a 

turbine trip test has shown differences between the test data and the results predicted by TRACE/PARCS such as the total 

core power and the dome pressure. This is mainly due to unstable steady-state conditions during the initialization process 

and modelling issues. The improvements introduced in this work to the TRACE model are including but not limited to the 

geometry of the reactor internals, the redesign of the main steam lines and the implementation of a rudimentary control 

system. Furthermore, the PARCS input model has been updated with the turbine trip test corresponding cross sections. The 

new designed coupled TRACE/PARCS model was eventually benchmarked against the same turbine trip plant data.

NPP Leibstadt (in German: Kernkraftwerk Leibstadt)(KKL)

Main Steam Isolation Valve (MSIV)

Reload Licensing (RLS)

Selected Rod Insertion (SRI)

GE Boiling Water Reactor (BWR/6)

Automatic Depressurization System (ADS)

Reactor Core Isolation Cooling (RCIC)

SIMULATE3

April 2017

Technical



NUREG/IA-0469 Development of a Coupled TRACE/PARCS Model for KKL and Benchmark Against the Turbine 

Trip Test

April 2017