NUREG/IA-0255

International 

Agreement Report

Coupled RELAP/PARCS Full Plant Model – Assessment of a 

Cooling Transient in Trillo Nuclear Power Plant

Prepared by:

J. C. Martínez-Murillo1, M. Novo

1

, R. Miró2, T. Barrachina2, and G. Verdú2

 1 Almaraz-Trillo AIE

Av Manoteras 46Bis, 28050 Madrid

SPAIN

 2

Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM)

Universitat Politècnica de València

Camí de Vera, s/n

46022 Valencia, SPAIN

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: November 2010

Date Published: August 2011

Prepared as part of the agreement on technical exchange and cooperation between the Consejo de Seguridad Nuclear (CSN) of 

Spain and the U.S. Nuclear Regulatory Commission (NRC) in the field of nuclear safety research.

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0255

International 

Agreement Report

Coupled RELAP/PARCS Full Plant Model – Assessment of a 

Cooling Transient in Trillo Nuclear Power Plant

Prepared by:

J. C. Martínez-Murillo1, M. Novo

1

, R. Miró2, T. Barrachina2, and G. Verdú2

 1 Almaraz-Trillo AIE

Av Manoteras 46Bis, 28050 Madrid

SPAIN

 2

Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM)

Universitat Politècnica de València

Camí de Vera, s/n

46022 Valencia, SPAIN

A. Calvo, NRC Project Manager

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: November 2010

Date Published: August 2011

Prepared as part of the agreement on technical exchange and cooperation between the Consejo de Seguridad Nuclear (CSN) of 

Spain and the U.S. Nuclear Regulatory Commission (NRC) in the field of nuclear safety research.

Published by

U.S. Nuclear Regulatory Commission
iii

ABSTRACT

This report presents the methodology to change from a 0-D kinetics core in a RELAP5/Mod3.3 full 

plant model of Trillo NPP to a coupled RELAP5/PARCS 3-D core. The coupled plant model is 

assessed against a real cooling transient that deformed the core power axial profile.

The coupled steady state core model was adjusted to a CASMO/SIMULATE calculation by 

condensing the cross-sections with the SIMTAB methodology. The need of information exchange 

between RELAP and PARCS during the transient was the reason for two modifications of the 

coupling between both codes.

The CASMO/SIMULATE calculation of the initial core conditions correspondent to the date 

of the transient was performed by Iberdrola Ingeniería.

The SIMTAB methodology used to build the PARCS input file, the automated tools for core 

input files mapping and the RELAP/PARCS modifications of control rod and boron content 

were performed by the Nuclear Engineering Group belonging to the Institute for Industrial, 

Radiophysical and Environmental Safety (ISIRYM) at the Universitat Politècnica de València 

(UPV).

The modifications to the RELAP5 full plant model to couple the 3-D PARCS core, the 

steady-state adjustments of PDDs, and the transient calculation and analysis were 

performed by the Thermo-Hydraulic group of Almaraz-Trillo NPPs.

The results of the transient show an almost perfect agreement with plant data for all the compared 

variables. The comparison of in-core parameters evolution with plant data is also very good and the 

3-D simulation allows to perform a more detailed analysis of core behaviour.

This report was prepared by the Thermo-Hydraulic group of Almaraz-Trillo NPPs (CNAT), with the 

help of the Polytechnic University of Valencia.

The Asociación Española de la Industria Eléctrica (UNESA, Electric Industry Association of Spain)

and Almaraz-Trillo NPPs AIE sponsored this work. 
iv

FOREWORD

This report represents one of the assessment or application calculations submitted to fulfil the 

bilateral agreement for cooperation in thermal-hydraulic activities between the Consejo de 

Seguridad Nuclear (CSN) and the U.S. Nuclear Regulatory Commission (NRC) in the form of a 

Spanish contribution to the NRC’s Code Assessment and Management Program (CAMP), the main 

purpose of which is to validate the TRAC/RELAP Advanced Computational Engine (TRACE) code.

CSN and the Asociación Española de la Industria Eléctrica (Electric Industry Association of Spain), 

together with some relevant universities, have established a coordinated framework (CAMP-Spain) 

with two main objectives: to fulfil the formal CAMP requirements and to improve the quality of the 

technical support groups that provide services to the Spanish utilities, CSN, research centers, and 

engineering companies.

The AP-28 Project Coordination Committee has reviewed this report, the contribution of one of the 

Spanish utilities to the above-mentioned CAMP-Spain program, for submission to CSN.

UNESA

December 2009
v

CONTENTS

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

FOREWORD................................................................................................................................ iv

CONTENTS .................................................................................................................................. v

LIST OF FIGURES ...................................................................................................................... vi

LIST OF TABLES......................................................................................................................... vi

EXECUTIVE SUMMARY ............................................................................................................ vii

ABBREVIATIONS.......................................................................................................................viii

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

2. PLANT DESCRIPTION ......................................................................................................2-1

3. TRILLO PLANT ANALYZER (APT)....................................................................................3-1

4. METHODOLOGY FOR 3-D PLANT MODEL CONFIGURATION ......................................4-1

4.1 CASMO / SIMULATE .................................................................................................4-2

4.2 SIMTAB......................................................................................................................4-2

4.3 PARCS.......................................................................................................................4-3

4.4 RELAP........................................................................................................................4-4

4.5 MAPTAB.....................................................................................................................4-6

4.6 PARCS Steady-State Alone calculation (P_SSA). .....................................................4-7

4.7 RELAP Steady-State Alone calculation (R_SSA). .....................................................4-8

4.8 Coupled steady-state simulation. (_CSS). .................................................................4-8

4.9 Coupled transient calculation. (_CTR). ......................................................................4-8

5. PLANT TRANSIENT DESCRIPTION.................................................................................5-1

6. TRANSIENT SIMULATION................................................................................................6-1

6.1 Plant Results ..............................................................................................................6-1

6.2 Core Results...............................................................................................................6-6

7. EXECUTION STATISTICS ................................................................................................7-1

8. CONCLUSIONS.................................................................................................................8-1

9. REFERENCES...................................................................................................................9-1
vi

LIST OF FIGURES

Fig. 1. Trillo Core Layout. Control Rods and PDDs location......................................................2-2

Fig. 2. Trillo Primary Circuit Model with RELAP5.......................................................................3-1

Fig. 3. Trillo Secondary Circuit Model with RELAP5..................................................................3-2

Fig. 4. APT Screen. Primary Circuit. ..........................................................................................3-3

Fig. 5. APT Screen. Secondary Circuit. .....................................................................................3-3

Fig. 6. Parallel execution scheme RELAP/PARCS....................................................................4-1

Fig. 7. Modified parallel execution scheme RELAP/PARCS......................................................4-1

Fig. 8. PARCS Neutronic Radial Map. “radial.mapN” ................................................................4-4

Fig. 9. RELAP Hydraulic Radial Map. “radial.mapH” .................................................................4-5

Fig. 10. RELAP Thermal Radial Map. “radial.mapT” .................................................................4-5

Fig. 11. PARCS Control Rod Simulation....................................................................................4-6

Fig. 12. PARCS Steady-state Alone Comparison......................................................................4-7

Fig. 13. High Pressure Heating Train Isolation. ........................................................................5-1

Fig. 14. Main Steam Pressure (Loop-1).....................................................................................6-1

Fig. 15. Cold Leg Temperature (Loop-1). ..................................................................................6-2

Fig. 16. Control rod Banks Insertion Depth (1D, 2D). ................................................................6-3

Fig. 17. Reactor Power (PKG) and Allowed Power (PERL).......................................................6-4

Fig. 18. Boron and Demineralised water injection. ....................................................................6-5

Fig. 19. In-core peaks. Bottom (PU) and top (PO).....................................................................6-6

Fig. 20. Core Reactivities (PARCS). ..........................................................................................6-7

Fig. 20. Moderator Temperature Evolution. Core Middle Axial Plane........................................6-8

Fig. 21. Moderator Temperature Evolution. Core 3D Visualization............................................6-9

Fig. 22. Fuel Linear Power. Core Middle Axial Plane. .............................................................6-10

Fig. 23. Fuel Linear Power. Core 3D Visualization. .................................................................6-11

Fig. 24. Fuel Linear Power. Core 3D Visualization. .................................................................6-12

Fig. 25. RELAP Steady-State Execution Statistics. ...................................................................7-1

Fig. 26. RELAP Transient Execution Statistics. .........................................................................7-2

LIST OF TABLES

Table 1. Main characteristics of Almaraz and Trillo NPPs. ........................................................2-1

Table 2. Execution Statistics......................................................................................................7-2
vii

EXECUTIVE SUMMARY

In this report is presented the methodology developed to improve a RELAP5 full plant model of 

Trillo NPP (Trillo Plant Analyzer - APT) with a coupled PARCS 3-D core.

The driving force for this development was a plant cooling transient occurred in year 2000. That 

transient resulted in an axial power profile deformation that was not possible to model with the 

RELAP point-kinetic (0-D) core. The improved RELAP/PARCS plant model behaviour is validated 

against those real cooling transient records.

The Trillo NPP is the only German design reactor in Spain. It is a three loop PWR of former KWU 

design, with a nominal core power of 3010 megawatts thermal, in commercial operation since 1988. 

Trillo NPP, as the rest of the German plants, has sophisticated automatic systems, including the 

control, protection, but specially the limitation system. This limitation system, that takes automatic 

actions with priority over operators, has a circuit of in-core surveillance that limits the reactor power 

by changes of the power axial profile.

All these automatic systems are simulated in detail with RELAP modules in APT. To reproduce a 

plant transient with RELAP/PARCS, it was necessary to modify the coupling capabilities in order to 

transfer the control rod position and the boron content from RELAP to PARCS, and the in-core 

measurements from PARCS to RELAP. 

As a result of this assessment, the methodology, modifications and tools to run coupled 

RELAP/PARCS full plant model calculations on Trillo NPP is well established. The almost perfect 

comparison with the real plant data, gives us the confidence of the adequacy of the model for future 

applications.

It is worthy to mention that this first long term simulation (1000 sec), with the very complete plant 

model of RELAP coupled to a relatively large core model with PARCS, required almost 50 times the 

CPU time required for the point kinetics model.
viii

ABBREVIATIONS

PWR Pressurized Water Reactor

CAMP Code Assessment and Management Program

APT Trillo Plant Analyzer

APT-3D Trillo Plant Analyzer with PARCS 3D core. 

CSN Consejo de Seguridad Nuclear (Spanish Nuclear Regulatory Council)

RELAP Reaction Excursion and Leak Analysis Program

TRACE TRAC/RELAP Advanced Computational Engine

CNAT Almaraz-Trillo NPPs AIE

UPV Universitat Politècnica de València

IB Iberdrola Ingeniería

PDD Power Density Detector.

FA Fuel Assembly

PARCS Purdue Advance Reactor Core Simulator

RCP Reactor Coolant Pump

YR Control System

YT Limitation System

YZ Protection System

FW Feedwater

PO Maximum Linear Heat Generation Rate Top of the Core

PU Maximum Linear Heat Generation Rate Bottom of the Core

CPU central processing unit

PIPE Main Hydraulic component of RELAP 

HS Heat Structure component of RELAP

MTC Moderator Temperature Coefficient

FTC Fuel Temperature Coefficient (Doppler)
1-1

1. INTRODUCTION

One of the most important goals, if not the main objective, which focuses the work of ThermoHydraulic (TH) section within the Fuel Department of Almaraz-Trillo NPPs (CNAT), is to understand 

and reproduce the dynamic behaviour of Almaraz and Trillo NPPs. This objective has led us to the 

development and daily use of advanced thermal-hydraulic simulation tools. These tools, with more 

than 20 years of development work, include numerical simulation models and graphical interfaces 

for both plants. 

The base tool for the developments of Almaraz Plant Analyzer (APA) and Trillo Plant Analyzer 

(APT[3]) is the RELAP5 code [1], currently at version mod.3.3. 

Among the lines of work opened to expand the simulation capabilities in CNAT, it is worth to 

highlight the improvement in core simulation using more advanced neutron kinetics tools (PARCS 

code [2]) than the "0-D” model included by default in RELAP5. 

The driving force for this development line, was a Trillo NPP cooling transient occurred in year 

2000. That transient resulted in an axial power profile deformation that was not possible to model 

with the RELAP point-kinetic (0-D) core. The improved RELAP/PARCS plant model (APT-3D) 

behavior is validated in this report against this real cooling transient.

To build a full plant model (APT) with a coupled core RELAP/PARCS (APT-3D), there were needed

several steps done in collaboration with the UPV and IB: 

Transfer CASMO/SIMULATE model calculations of Trillo to a PARCS model (SIMTAB 

method).

Partial core model with RELAP/PARCS.

Modifications in RELAP / PARCS to allow movement of the control rods in PARCS from 

RELAP.

Modifications in RELAP / PARCS to allow boron changes in PARCS from RELAP.

Replacement of a simplified core (0-D) in the full plant model (APT) for a coupled 3-D core.

Adjustments of steady-state RELAP/PARCS to CASMO/SIMULATE.

Power density detectors (PDDs) calibration.

The Trillo NPP is particularly conducive to the use of advanced methods for core simulation 

because it has two characteristics that make it unique within the Spanish nuclear power fleet:

1. It has in-core on-line instrumentation, Power Density Detectors (PDDs) that generates 

automatic actions through the system of limitation (YT).

2. The limitation system design prevents the reactor trip during relevant transients like a single 

RCP trip or a turbine trip, stabilizing the plant to a partial power in a few seconds.

This report summarizes the first assessment of the Trillo plant APT-3D model against a real cooling 

transient occurred in year 2000. The initial reproduction of this transient with RELAP model (APT0D) led to develop in 2001 of the simplified module [yq], to approximate the deformation of the axial 

power profile that was the cause of the activation of the peak power limits (PU) system and the 

power reduction to 90%. This coupled RELAP / PARCS simulation reproduces the 3-D neutron flux 
1-2

deformation in PARCS with the control rod insertion and the boron injection ordered by RELAP.
2-1

2. PLANT DESCRIPTION

The Almaraz and Trillo NPPs are owned by different shareholders (Iberdrola Generación SA, 

Endesa Generación SA, Gas Natural SDG, Hidroeléctrica del Cantábrico, and Nuclenor S.A.), and 

operated by Almaraz-Trillo NPPs AIE (CNAT).

The current relevant characteristics of Almaraz and Trillo NPPs are summarized in the following 

table:

Table 1. Main characteristics of Almaraz and Trillo NPPs.

PLANT PARAMETERS ALMARAZ TRILLO

Number of Units 2 1

Design NSSS Westinghouse KWU-Siemens

Coolant Loops 3 3

Thermal Power (Mwt) 2947 | 2729 3010

Gross Electrical Power (Mwe) 1050 | 980 1066

Net Electrical Power (Mwe) 1013 | 946 1002

Cycle Length (months) 18 12

Refrigeration System Cooling Pond Cooling Towers

Commercial Operation Since 1981 | 1983 1988

The Trillo NPP is the only German design reactor in Spain. It is a three loop PWR of former KWU 

design, with a nominal core power of 3010 megawatts thermal, in commercial operation since 1988. 

There are two main differences between Trillo and Almaraz design. The first one is the great 

number of systems / equipment of Trillo compared with Almaraz. Despite of the fact than both are 

three loop plants, Trillo has four redundancies in safeguard systems (safety injection, emergency 

power, emergency feedwater). 

The second difference is that Trillo, as the rest of the German design plants, has sophisticated 

automatic systems, including the control (YR), protection (YZ), but specially the limitation system

(YT). This limitation system, that takes automatic actions with priority over operators, has a specific 

circuit of in-core surveillance that can reduce the reactor power, for example by deformations of the 

power axial profile.

The core layout of Trillo (177 FAs) is represented in the next figure. The Power Density Detectors

(PDD or LVD in German) performs the surveillance of the neutron flux in 6 positions of the core with

6 different axial levels (36 detectors). The maximum of the three upper levels (PO) and the three 

lower levels (PU) are compared against calculated limit values.
2-2

Fig. 1. Trillo Core Layout. Control Rods and PDDs location.

MAX PO

MAX PU
3-1

3. TRILLO PLANT ANALYZER (APT)

The Trillo NPP simulation with RELAP5 started in 1988 with the start-up tests simulation. The 

RELAP5 numerical model has grown over the years with the different plant applications (i.e. 

NUREG/IA-0177 [7], operator training, real plant analysis, safety analysis, PSA scenarios, etc.) up to 

the current model scope.

The thermo-hydraulic and I&C simulations includes:

The primary circuit (Fig. 2):

o Vessel with a triple downcomer and two channel core, point kinetic core.

o Detailed loops and RCPs

o Pressurizer with pressurizer relief tank, spray lines, heaters.

o Auxiliary Systems: Volumetric and chemistry control

o Safeguard Systems: Safety Injection (High Pressure, Low Pressure, Accumulators), 

Extra Borating System.

o Detailed Steam Generators (tube-side)

Fig. 2. Trillo Primary Circuit Model with RELAP5.
3-2

The secondary circuit: (Fig. 3)

o Detailed Steam Generators (shell side)

o Feedwater system (from the feedwater tank to SGs, including both heating trains 

and the start-up feedwater system).

o Main Steam System (relief and safety valve stations)

o Main turbine and turbine bypass systems

o Safeguard Systems: Emergency feedwater system.

Fig. 3. Trillo Secondary Circuit Model with RELAP5.

The I&C simulation includes:

o Control, Protection and Limitation Systems

o Steam Generators Level control.

o Control loops and functional groups.

The RELAP5 numerical model has been improved with a graphical and interactive simulation

performed by means of CSV [5]/ NPA [6]software. Based on our simulation needs, several 

modifications were performed to this numeric / graphic coupling, allowed us to spread the use of the 

Plant Analyzers among the organization.
3-3

Fig. 4. APT Screen. Primary Circuit.

Fig. 5. APT Screen. Secondary Circuit.
4-1

4. METHODOLOGY FOR 3-D PLANT MODEL CONFIGURATION

The calculation methodology is based on running parallel simulations, thermo-hydraulics (RELAP5) 

and neutronics (PARCS). For this parallel runs, with coordination of time steps of both codes, the 

"Parallel Virtual Machine" PVM [4]is used.

PVM

dens

tfuel

RELAP

(TH)

PARCS

(N)

power

Fig. 6. Parallel execution scheme RELAP/PARCS.

The diagram above indicates the standard variables sent from RELAP each time step (moderator 

density and fuel temperature), for PARCS to calculate the distribution of netron flux with the cross 

sections tabulated in terms of these variables. The power of each node in the core is transferred to 

the heat structures simulating the fuel elements in RELAP.

However for the application on whole plant models it has been necessary to modify the calculation 

scheme to introduce the effect in PARCS of the control rod movement and the boron content 

calculated in RELAP:

dens

tfuel

cr_pos

c_bor

RELAP

(TH)

PARCS

(N)

power

PVM

Fig. 7. Modified parallel execution scheme RELAP/PARCS.

The first modification of the coupling scheme allows the transfer of RELAP control variables that 

specify in PARCS the positions of the control rods. This change is essential to reproduce plant 

transients, where the RELAP model simulates these positions based on orders from the automatic 

systems of control, limitation and protection.

The second modification allows the transfer of RELAP boron density in each of the core volumes, to 

be used in PARCS as a value of interpolation between two sets of cross sections at different boron 

concentrations.

Before running a coupled calculation of a plant transient, it is necessary to perform a series of steps 
4-2

to get a properly configuration of PARCS and RELAP models and their link:

Models Configuration Steps:

1. CASMO / SIMULATE: Core 3D simulation with representative cross sections of the burning 

conditions of the day of the event.

2. SIMTAB: Collapsation of the cross sections for use in PARCS.

3. PARCS: Generation of input file to PARCS - steady state / transient. Neutron model.

4. RELAP: Generation of input file to RELAP - steady state / transient. TH model.

5. MAPTAB: Generation of tabular map of correspondence between TH channels (RELAP)

and neutronic compositions (PARCS).

Simulation Steps:

6. PARCS steady-state alone calculation. (P_SSA).

7. RELAP steady-state alone calculation. (R_SSA).

8. Coupled steady-state simulation. (_CSS).

9. Coupled transient calculation. (_CTR).

Three tools have been developed for automatic input file generation (RELAP5, PARCS) and 

mapping method between both codes (MAPTAB).

4.1 CASMO / SIMULATE

In collaboration with Iberdrola Ingeniería (IB), the Fuel Department of CNAT has developed 3-D 

neutronic models of Almaraz and Trillo cores with CASMO-4/SIMULATE-3. These models are 

validated every cycle against the core neutron flux maps in both plants, through the simulation of

burnup history of the core.

From a geometric mesh of the core, CASMO-4/SIMULATE-3 models calculate the neutron cross 

sections and three-dimensional neutronic parameters that characterize the steady state with a given 

position of the control rods.

The CASMO / SIMULATE output files, with a high number of "neutronic compositions" or types of 

cells with different cross sections due to fission products obtained as a result of burning the fuel, are

used as starting point for the development of PARCS neutronic 3-D model.

The CASMO / SIMULATE model of Trillo used for this application includes 178 channels (177 FAs

+ 1 radial reflector), with 32 axial levels in the active zone plus two axial levels for the lower and 

upper axial reflectors.

Besides the cross-sections, the output files of CASMO / SIMULATE contain geometric information 

and boundary conditions that are used in the automatic generating of the RELAP and PARCS 

models.

4.2 SIMTAB

The objective of SIMTAB methodology (developed by UPV) is to reduce the number of 
4-3

"compositions" calculated with CASMO / SIMULATE and transform them in tabular format for use in 

PARCS or other neutronic codes.

The method of reducing the number of compositions is based on grouping them in different burnup 

intervals ( B), i.e. the compositions of CASMO / SIMULATE differentiated less than B in steady 

state are represented by a single composition in PARCS model.

In this application a burning discrimination of B = 0.1 GWd / TU have been used. The 177 

FAs have been reduced in PARCS to 36 different FAs.

Other important parameters for the generation of the cross sections files in PARCS, are the 

expected ranges of variation of their basic parameters (moderator density, fuel temperature) during 

the transient of interest. If during the transient these ranges of variation are exceeded, it would be 

necessary to recalculate the collapsed cross sections.

For this application wide ranges of density (222 to 914 kg/m3) and fuel temperature (422 to 

1300 o C) have been used.

The outcomes of this method are the tabulated cross section files NEMTAB (without control rod 

“unroded”) and NEMTABR (with control rod “roded”).

For this application two sets of tables at two the boron concentrations that cover the range 

of variation during the transient:

o NEMTAB_639ppm / NEMTABR_639ppm.

o NEMTAB_1000ppm / NEMTABR_1000ppm. 

4.3 PARCS

PARCS is the acronym for "Purdue Advance Reactor Core Simulator." This 3-D neutronic code is 

distributed by the NRC for its coupling with the latest versions of the thermal-hydraulic code RELAP 

and TRACE.

The most important feature of PARCS is that it can be used to simulate steady state operation and 

transient resulting from reactivity disturbances. The code solves the diffusion equation for two 

groups of neutrons in 3-D geometry and with temporal dependency.

The PARCS input deck represents the core configuration in one burnup point, the burning of the 

core model during the operation of the cycle is performed by other codes like CASMO / SIMULATE. 

The most relevant input data are the geometric information, the cross sections and the initial and 

boundary conditions. These three pieces of input are prepared in independent files and 

automatically generated from the SIMULATE output files.

For this application, the 36 different FAs (“radial.mapN” see next figure) obtained with SIMTAB, plus 

the radial reflector, are subdivided in 32 + 2 axial levels.
4-4

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17

1 0 0 0 0 0 37 37 37 37 37 37 37 0 0 0 0 0

2 0 0 0 37 37 37 11 9 10 8 6 37 37 37 0 0 0

3 0 0 37 37 8 1 12 13 13 13 12 5 10 37 37 0 0

4 0 37 37 11 2 14 21 34 4 33 21 14 2 10 37 37 0

5 0 37 9 2 28 36 20 17 19 18 19 35 28 2 10 37 0

6 37 37 5 14 35 16 30 17 22 17 31 16 35 14 1 37 37

7 37 7 12 21 19 31 25 26 23 26 25 29 20 21 12 11 37

8 37 8 13 33 18 17 26 15 32 15 26 17 17 34 13 8 37

9 37 10 13 4 19 22 24 32 3 32 23 22 19 4 13 11 37

10 37 8 13 34 17 17 27 15 32 15 26 17 18 33 13 8 37

11 37 11 12 21 20 30 25 26 23 26 25 31 19 21 12 6 37

12 37 37 1 14 36 16 31 17 22 17 29 16 35 14 5 37 37

13 0 37 8 2 28 36 19 18 19 17 20 36 28 2 10 37 0

14 0 37 37 11 2 14 21 33 4 34 21 14 2 11 37 37 0

15 0 0 37 37 10 5 12 13 13 13 12 1 9 37 37 0 0

16 0 0 0 37 37 37 7 9 11 8 10 37 37 37 0 0 0

17 0 0 0 0 0 37 37 37 37 37 37 37 0 0 0 0 0

Fig. 8. PARCS Neutronic Radial Map. “radial.mapN”

4.4 RELAP

The RELAP model used for the Trillo Plant Analyzer (APT) is modified in three aspects:

Hydraulic: replace the simplified APT core with two axial channels and a bypass by a

chosen number of channels in parallel.

Thermal: replace the simplified APT core with two heat structures representing the fuel 

rods cooled for each of the hydraulic channels to a number of chosen heat structures and 

coupled to the PARCS neutronic calculation.

Automatic: replace the simplified module [yq] used to approximate the deformation of the 

axial profile, by the direct reading of the linear power structures that represent the ECs with 

PDDs.

To create a core model with component numbers compatible with the numbering of the rest of APT

model, a specific program has been developed. For this purpose the program reads:

Two radial maps, one Hydraulic and one Thermal to obtain the component numbering.

The geometric information from SIMULATE output files.

The initial conditions from SIMULATE output files.

For the present application the 36 different FAs of the neutronic map (Fig. 8) has been extended to 

42 to represent specifically the 6 FAs that house the core instrumentation lances PDDs (Fig. 1). In 

this way the 3-D RELAP core is formed with 42 hydraulic and 42 thermal channels.
4-5

In the radial hydraulic map the 42+1 channels (PIPE components) are defined (see next figure), 

with the same axial division 32+2 levels used in CASMO / SIMULATE and PARCS. 

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

578 578 578 578 578 578 578

P 578 578 578 192 190 191 177 175 578 578 578

O 578 578 177 170 193 194 194 388 193 174 191 578 578

N 578 578 192 171 195 273 374 173 373 273 195 171 191 578 578

M 578 190 171 297 387 272 198 271 199 271 386 297 171 191 578

L 578 578 174 195 386 197 299 198 274 198 371 197 391 195 170 578 578

K 578 176 193 273 271 371 293 294 291 294 293 298 272 273 193 192 578

I 578 177 194 373 199 198 389 196 372 196 294 198 198 374 194 177 578

H 578 191 194 173 271 274 292 372 172 372 291 274 271 173 194 192 578

G 578 177 194 374 198 198 296 196 372 196 392 198 199 373 194 177 578

F 578 192 193 273 272 299 293 294 291 294 293 371 271 273 193 175 578

E 578 578 170 195 390 197 371 198 274 198 298 197 386 195 174 578 578

D 578 177 171 297 387 271 199 271 198 272 387 297 171 191 578

C 578 578 192 171 195 273 373 173 374 273 195 171 192 578 578

B 578 578 191 174 193 393 194 194 193 170 190 578 578

A 578 578 578 176 190 192 177 191 578 578 578

578 578 578 578 578 578 578

Fig. 9. RELAP Hydraulic Radial Map. “radial.mapH”

In the radial thermal map the 42+1 Heat Structures are defined (see next figure), with the same 

axial division 32+2 levels used in CASMO / SIMULATE and PARCS.

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

578 578 578 578 578 578 578

P 578 578 578 861 859 860 858 856 578 578 578

O 578 578 858 851 912 913 913 751 912 855 860 578 578

N 578 578 861 902 914 921 934 904 933 921 914 902 860 578 578

M 578 859 902 928 936 920 917 919 918 919 935 928 902 860 578

L 578 578 855 914 935 916 930 917 922 917 931 916 754 914 851 578 578

K 578 857 912 921 919 931 925 926 923 926 925 929 920 921 912 861 578

I 578 858 913 933 918 917 752 915 932 915 926 917 917 934 913 858 578

H 578 860 913 904 919 922 924 932 903 932 923 922 919 904 913 861 578

G 578 858 913 934 917 917 927 915 932 915 755 917 918 933 913 858 578

F 578 861 912 921 920 930 925 926 923 926 925 931 919 921 912 856 578

E 578 578 851 914 753 916 931 917 922 917 929 916 935 914 855 578 578

D 578 858 902 928 936 919 918 919 917 920 936 928 902 860 578

C 578 578 861 902 914 921 933 904 934 921 914 902 861 578 578

B 578 578 860 855 912 756 913 913 912 851 859 578 578

A 578 578 578 857 859 861 858 860 578 578 578

578 578 578 578 578 578 578

Fig. 10. RELAP Thermal Radial Map. “radial.mapT”
4-6

The grey channels (PIPE-578 and HS-578) represent the core bypass (coupled to the radial 

reflector). The yellow and pink channels correspond to the outer and inner channels in the 0-D core

model. The 6 light blue channels (PIPE and HS) are the FAs channels with PDD instrumentation.

The total number of PIPE volumes and HS in this new 3-D core is 1462 each.

Note that it is not necessary to use the same number of hydraulic than thermal channels. In fact it 

was tested a core model with 2 PIPES and 42 HS, obtaining quite similar steady-state results than

the 42/42 configuration.

4.5 MAPTAB

The parallel calculation process of RELAP / PARCS requires a "mapping" between the RELAP 

hydraulic and thermal components and the PARCS neutron nodes. In this way the transfer of 

thermal-hydraulic information (moderator density, fuel temperature, and boron concentration) in one 

direction and the power in the other, is performed.

This tabular map is defined in “MapTab” file, automatically generated with a specific program that 

reads the “radial.mapN”, “radial.mapH” and “radial.mapT”.

The modification of the control rod position transfer is also specified in the same “MapTab” file by 

defining the RELAP control variables that simulate the different control rod positions.

For this application each control rod bank D (4 CRs) has been divided in two semi-banks that 

moves simultaneously. For example the bank D10 is represented by semi-banks 1 and 2. In the 

next figure the D10 to D60 banks are represented in colour.

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15

0 0 0 0 0 0 0 D&L Control Rods Banks & Sequences

P 0 0 0 0 0 0 0 0 0 0 0 I II III IV

O 0 0 0 0 5 0 15 0 • 17 0 0 0 0 D20 D10 D30 D60

N 0 0 0 0 1 0 19 0 12 0 4 0 0 0 0 D50 D40 D40 D50

M 0 0 0 14 0 0 0 16 0 0 0 14 0 0 0 D10 D20 D20 D10

L 0 0 0 3 0 18 0 9 0 8 0 18 0 • 2 0 0 0 D40 D50 D50 D40

K 0 0 17 0 0 0 0 0 0 0 0 0 0 0 6 0 0

J 0 0 0 11 0 7 0 • 13 0 13 0 10 0 19 0 0 0 0 • PDD Location

H 0 0 15 0 16 0 0 0 0 0 0 0 16 0 15 0 0

G 0 0 0 19 0 10 0 13 0 13 0 • 7 0 11 0 0 0

F 0 0 6 0 0 0 0 0 0 0 0 0 0 0 17 0 0

E 0 0 0 2 0 • 18 0 8 0 9 0 18 0 3 0 0 0

D 0 0 0 14 0 0 0 16 0 0 0 14 0 0 0

C 0 0 0 0 4 0 12 0 19 0 1 0 0 0 0

B 0 0 0 0 17 0 • 15 0 5 0 0 0 0

A 0 0 0 0 0 0 0 0 0 0 0

0 0 0 0 0 0 0

Fig. 11. PARCS Control Rod Simulation.
4-7

There are four different sequences of control bank insertion (I to IV in Fig. 11). In the date of the 

analyzed transient the sequence IV was active in the plant, that is the bank 1-D corresponds to 

D60, 2-D to D50, 3-D to D10, and 4-D to D40.

4.6 PARCS Steady-State Alone calculation (P_SSA).

The first step after the generation of the PARCs input file is to perform single steady-state run of 

PARCS to compare with the reference case of CASMO/SIMULATE. The steady-state calculation is 

perform with the additional files and conditions:

Geometrical information “GEOM_LWR”.

Cross sections NEMTAB and NEMTABR at the SIMULATE boron concentration.

Fixed thermal-hydraulic boundary conditions (density and fuel temperatures from SIMULATE).

Static position of control rods.

The calculated solution of PARCS is compared in terms Keff and axial power profile with the one

obtained by CASMO / SIMULATE. In the following figure, the average core axial profile of SIMUATE 

is compare with two PARCS calculations, one with CRs in the same positions of SIMULATE and the 

other with the active sequence IV. 

TRILLO AN02-00 

Steady State / Axial Power Distribution

0

50

100

150

200

250

300

350

0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2

Power (-)

Distance from Bottom Active Core (cm)

SIMULATE CR D60=300 CR D10=300

Fig. 12. PARCS Steady-state Alone Comparison.

The first case reproduced is well adjusted, while the second (with the real control bank sequence) is 

a little more bottom peaked. 
4-8

4.7 RELAP Steady-State Alone calculation (R_SSA).

Once the simplified TH core simulation of APT-0D (2 PIPEs / 3 HSs) is replaced by the complex 3-D 

core (42 PIPEs / 42 HSs), a null transient is performed to reach the stable initial conditions for the 

full plant model. For this calculation the reactor power is fixed to 100% (3010 Mwth), with a 

distribution over the HSs obtained from the SIMULATE reference case. 

The RELAP steady-state alone calculation with APT-3D is compared against the same calculation 

APT-0D to check the primary variables, specially the mass flow rate distribution in the primary 

circuit, core, and bypass.

PDDs calibration

The linear power values in the FAs of the positions occupied by PDDs will be used as input 

measurement of the limitation system YQ. Each one of the 36 detectors surveys a region of the 

core. With the linear heat generation rate (LHGR) of each one of the 177 FAs in SIMULATE 3D 

model, the maximum value of each surveillance region is calculated.

The factor between the maximum linear power calculated for each monitoring area and the 

calculated value for the FA with PDD, is used as calibration factor YQ system simulation. These 

calibration factors are applied directly over the calculated RELAP linear power of the FAs with PDD.

The RELAP restart file of this calculation will be used in the next step. 

4.8 Coupled steady-state simulation. (_CSS).

This step is quite similar to the usual process of RELAP steady state calculation. The reactor 

kinetics is activated in the steady-state input as a restart calculation of the previous step. The 

differences between a typical APT-0D and this APT-3D steady state are:

Activation of the coupling with PARCS (card 1 in RELAP, and TH feedback in PARCS) 

Change from 0-D power variables to 3-D power in the automatic simulation.

Change from simplified YQ simulation to full YQ-3D simulation.

The coupled RELAP/PARCS calculation process is performed by running both codes in parallel with 

PVM and the mapping between TH cells and neutronic nodes defined in MAPTAB file. 

The change in the active sequence (IV) of the CRs with respect to SIMULATE configuration does 

not modify significantly the steady state power distribution in the core, from the previous step.

The restart output files of RELAP and PARCS will be used in the next step.

4.9 Coupled transient calculation. (_CTR).

The last step is the RELAP/PARCS coupled transient calculation. The initial event is simulated in 
4-9

the RELAP input file, and the response of the automatic systems modelled in APT conduct the 

dynamic evolution of the plant.

The change of the TH conditions in the core, the movement of the control rods or the change in 

boron concentration calculated with RELAP, modify the solution of the core neutron flux of PARCS, 

that feedback the power in RELAP HSs and the in-core and ex-core simulated measurement of 

RELAP.
5-1

5. PLANT TRANSIENT DESCRIPTION

The initial event of the transient occurred in Trillo NPP on the 18/10/2000, when the plant was 

operating at 100% power, was an isolation and bypass of a high pressure feedwater heating train

(Fig. 13). 

R K R P R P

R K R P R P

FW Heaters Train A

FW Heaters Train B

RF RF

RF RF

SGs FW Pumps

Fig. 13. High Pressure Heating Train Isolation.

The second train remained in operation heating the water that mixed in the feedwater header with 

the bypassed cold water of the affected train. The reduction of the feedwater temperature produces 

a cooling down of the primary circuit and the correspondent power increase by the negative 

Moderator Temperature Coefficient (MTC).

At the date of the transient the cycle was close to its middle point (MOC), so the MTC was big 

enough to increase the power over the limitation values of total power (L-RELEB), and axial profile 

deformation (PU-RELEB).

The insertion of control rods limited the increase of the reactor power (PR) over its allowed value 

(PERL). The negative reactivity of the control rods compensated the positive feedback from the 

coolant temperature reduction. Depending on the value of this coefficient (function of the cycle 

burnup), it will be required more or less length control rods insertion.

In the evolution of the transient, the insertion of control rods was important enough to generate 

boundaries of PU-RELEB (by axial power profile deformation in the lower part of the core) and DSTAFAB (by absolute position of banks). The first was the cause of the reduction of PERL at a rate 
5-2

of 4.5% / min, while the control position of D bank and the limit of STAFAB led to a temporal 

boration from TB / TA systems.

The reactor power was stabilized at about 84% by the actions of the limitation system, supported by 

the operator decision of reducing the power demanded in the turbine, that helped the recovery of 

the average temperature in the primary circuit. At the same time ordered injections of demineralised 

water to compensate the injected boron.

Subsequently the reactor power was increased up to 100%, reaching a power turbine ~97%.
6-1

6. TRANSIENT SIMULATION

The results of the transient calculation introduced in (4.9), are analyzed below in two sections:

First, the evolution of the relevant plant simulator parameters are compared with plant 

records (usual validation process). 

Second, the analysis of core behaviour by comparing the simulation versus available 

records, and 3-D representation of core status and axial profile evolution.

In the following comparison plots, the APT variables are identified with parenthesis in their legends.

6.1 Plant Results

The RELAP input for the transient calculation is exactly the same as the one used for APT-0D. It 

includes:

T=10 seconds. Flip the three-way valve RL11S001 to the bypass of heater train RL21

position.

Simplified reduction of turbine extractions to the isolated heaters.

Manual turbine power reduction (t > 300 sec)

Manual orders of demineralised water injection (t > 545 sec)

The feedwater temperature reduction and the main steam pressure evolution have a direct 

influence in the primary cold legs temperatures. 

Fig. 14. Main Steam Pressure (Loop-1).
6-2

As it can be seen in (Fig. 14) and (Fig. 15) APT-3D correctly reproduces the initial reduction of 

steam pressure and cold leg temperatures, and the posterior increase due to the turbine power 

reduction.

Fig. 15. Cold Leg Temperature (Loop-1).

With the correct simulation of the cold leg temperature conditions, the coupled model simulates the 

increase in power and the automatic bank insertion order by the limitation system (YT). First 1-D 

and L banks in parallel and second 1-D alone after reaching the first limit of the axial power profile 

deformation in the lower part PU-RELEB.
6-3

Fig. 16. Control rod Banks Insertion Depth (1D, 2D).

The allowed reactor power (PERL) is automatically reduced when the second limit of PU-RELEB is 

reached. The total reactor power (PKG) is reduced to follow the allowed value by continuous 1-D 

bank insertion.
6-4

Fig. 17. Reactor Power (PKG) and Allowed Power (PERL).

The additional 1-D insertion reaches one of the bank insertion limits (D-STAFAB) that automatically 

activate the temporal injection of borated water.
6-5

Fig. 18. Boron and Demineralised water injection.

All the plant parameters compared in the previous figures are in good agreement with the 

correspondent plant records. The answer of the new 3-D core to the reactivity perturbations 

(moderator temperatures, fuel temperatures, control rods insertion depth, boron concentration), in 

terms of total power is almost perfect.

To reproduce the automatic movement of control rod banks in this particular scenario, it is 

necessary to match the real core behaviour that is analyzed in the following section.
6-6

6.2 Core Results

The plant evolution during this transient was actually driven by the axial power profile deformation

that triggered the limits of PU-RELEB.

Fig. 19. In-core peaks. Bottom (PU) and top (PO).

The comparison of PO and PU signals calculated with the coupled model and the plant records 

indicate a very good approximation even after the stage at which rated power is impacted by the 

effect of boron injection and the turbine load reduction.

This is the key variable in this scenario and the driven force for develop the coupled methodology. 

The correct behavior of the calculated power density peak in the bottom section of the core, feeds 

the RELAP automatic simulation of PDDs, and finally triggers at the correct times the automatic 

countermeasures.

The PARCS calculated reactivity feedbacks due to the fuel, moderator, control rods and boron is 

included in the following figure.
6-7

Fig. 20. Core Reactivities (PARCS).

The 3-D methodology for coupled core calculations allows the visualization of the full core 

conditions at the initial steady-state and during the transient.

The following 3-D figures helps to understand the in-core evolution of the main contributors to 

reactivity feedback, moderator temperature and fuel linear power or temperature.
6-8

Fig. 21. Moderator Temperature Evolution. Core Middle Axial Plane.

In the sequence of figures to 0, 200, 500 and 1000 seconds, it is easy to follow a clear decrease in 

the moderator temperature at the beginning of the transient, by reducing the inlet temperature 

(without changing the radial core profile). In the subsequent power reduction the flattening of the 

radial core profile can also be identified.
6-9

Fig. 22. Moderator Temperature Evolution. Core 3D Visualization.

In the core 3-D visualization, the moderator temperature distribution along the core and its variation 

along the transient can be analyzed. The temperature reduction focused in the lower part of the 

core at the beginning, and to the upper part after the power reduction.
6-10

Fig. 23. Fuel Linear Power. Core Middle Axial Plane.

The linear power distribution in the fuel rods (middle axial plane of the core) reflects the initial power 

increase and the final power reduction.
6-11

Fig. 24. Fuel Linear Power. Core 3D Visualization.

In the 3-D view of the linear power of fuel rods, it can be seen that the effect of initial power 

increase is more pronounced in the bottom core for the insertion of bank D. The subsequent 

progressive reduction of power, due to the bank's total insertion and boron injection, balances both 

parts of the core.

Finally, in the next figure is included a chart with the axial power profile evolution in the FA E-04 

(HS-753), the one with PDD lance and maximum linear power.
6-12

Fig. 25. Fuel Linear Power. Core 3D Visualization.

In this representation it can be perfectly seen the profile deformation toward the bottom of the core, 

which shows a maximum around 200 seconds, and is the responsible for the activation of the limits 

of PU.

The effect of borating pulse collaboration (t ~ 400 sec) is to depress the total linear power profile, 

while as D1 bank completes the insertion (t ~ 500 sec), the profile pivots from bottom skewed to top 

skewed.

The recovery of the inlet temperature by the turbine power reduction, accompanied by the 

cancellation of the effect of boron by injection of demineralised water, makes than the total power 

and the axial power profile reaches a stable situation at partial power.
7-1

7. EXECUTION STATISTICS

In this chapter the execution times of APT-0D and APT-3D are compared. The code versions are 

RELAP5/mod3.3_Patch1 and PARCS v2.7. The computer used is a “Intel(R) Core(TM)2 Duo CPU 

P8700 @ 2.53GHz”.

The first comparison is between steady-state calculations, both of them realized only with 

RELAP5/mod3.3 but different nodalization. The core model expansion increased the size of APT 

model from:

Hydraulic Volumes 656 -> 2106 (3 times)

Heat Structures 204 -> 1654 (8 times)

Fig. 26. RELAP Steady-State Execution Statistics.

The CPU time for the initialization of the plant model (1000 sec simulation time) increased from 199 

sec in APT-0D (5 times faster than real time) to 3145 sec in APT-3D (3 times slower than real time).

The factor between both RELAP5 models APT-3D / APT-0D was around 16.

The second comparison is between transient calculations, the APT-0D was performed only with 

RELAP5/mod3.3, and the APT-3D with the coupled RELAP/PARCS. So the differences between 

both calculations include the model expansion and kinetic model:
7-2

Hydraulic Volumes 656 -> 2106 (3 times)

Heat Structures 204 -> 1654 (8 times)

Point kinetic (RELAP) -> 3D kinetic (PARCS)

Fig. 27. RELAP Transient Execution Statistics. 

The RELAP CPU time for the transient (1000 sec simulation time) increased from 424 sec in APT0D (2.3 times faster than real time) to 6940 sec in APT-3D (7 times slower than real time).

The factor between RELAP5 executions APT-3D / APT-0D was maintained around 16.

The PARCS CPU time for the transient, not included in (Fig. 27), was 12,283 sec. (12 times slower 

than real time).

Table 2. Execution Statistics

APT-0D APT-3D

RELAP RELAP PARCS Total

Steady-State 199 3,145

Transient 424 6,940 12,283 19,223

So the total CPU time was 19,223 sec. 68% employed by PARCS and 32% by RELAP5. The factor 

between the RELAP/PARCS transient APT-3D / RELAP APT-0D was around 45.

The conclusion is that the important increase in the execution time of the transient when moving 

from APT-0D to APT-3D is due mostly to PARCS execution time and partially to the increase in the 
7-3

size of RELAP5 model.
8-1

8. CONCLUSIONS

The coupled RELAP/PARCS methodology for full plant applications has been developed and used 

for the first time with Trillo Plant Analyzer APT-3D.

This first novel application has been a real challenge to the coupled simulation capacities, as 

included inherent reactivity perturbations and also rod control movements and changes in boron 

concentration.

The precise reproduction of the core axial profile evolution during the transient was a crucial input 

to the in-core and ex-core systems simulations that conduct the automatic plant dynamic behavior.

The important increase in the execution time of the transient when moving from APT-0D to APT-3D 

is due mostly to PARCS execution time and partially to the increase in the size of RELAP5 model.

The results obtained validate the coupling methodology extended to full plant models, and the 

proper settings of the codes modifications related to control rod movement and boron 

concentration.
9-1

9. REFERENCES

1. “RELAP5/MOD3.3 CODE MANUAL", NUREG/CR-5535, December 2001, Information 

Systems Laboratories, Inc. Rockville, Maryland Idaho Falls, Idaho.

2. T. Downar, Y. Xu, T. Kozlowski, D. Carlson, “PARCS v2.7 US NRC Core Neutronics 

Simulator”. School of Nuclear Engineering, Purdue University, 2006.

3. "APT - Simulación de Trillo con RELAP5 y NPA.", CNT1-TH-111, J.C. Martínez-Murillo, 

CNAT, Rev.0, Dic-2000.

4. “PVM: Parallel Virtual Machine. A Users' Guide and Tutorial for Networked Parallel 

Computing”. MIT Press Scientific and Engineering Computation, Janusz Kowalik, Editor, Massachusetts 

Institute of Technology, 1994.

5. Snider. D.M., "Nuclear Plant Analyzer. Computer Visual System (CVS)". EGG-EAST9095, Reference Manual. Version 1.4, 1/8/94.

6. Snider. D.M., "Nuclear Plant Analyzer (NPA)”. EGG-EAST-9096, Reference Manual. 

Version 1.4, 1/9/94.

7. M.F. Lozano, P. Moreno, A.L. Cedillo, M.N. Sanjurjo “Assessment of a Reactor 

Coolant Pump Trip for Trillo NPP with RELAP5/Mod3.2”, NUREG/IA-0177, March, 

2000.

8. “C.Trillo. Aislamiento de la cadena 1 de calentadores de alta presión. 

Reproducción del transitorio AN-02/2000 con el código RELAP5/MOD3.2”, CO01/005, J.C. Martínez-Murillo, CNAT. Rev.0, Nov-2001.