NUREG/IA-0217 

 

 International 

 Agreement Report

 

Investigations of the VVER-1000 

Coolant Transient Benchmark I with the 

Coupled Code System RELAP5/PARCS 

Prepared by: 

V. Sánchez-Espinoza 

Forschungszentrum Karlsruhe GmbH 

Hermann-von-Helmholtz-Platz 1 

76344 Eggenstein-Leopoldshafen 

Germany 

A. Calvo, NRC Project Manager 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

December 2009 

 

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 

This report presents the investigations performed to validate the coupled code RELAP5/PARCS 

using data of a pump restart-up test carried out at the Kozloduy nuclear power plant (KNPP). 

An integral plant model—including the relevant primary and secondary systems as well as a 

detailed model of the reactor pressure vessel and of the core—was developed for RELAP5 and 

PARCS to simulate the pump restart-up test. From the comparison of the thermal-hydraulic data 

obtained during the tests with the code’s predictions, it can be stated that the overall trends of 

most plant parameters are in a reasonable agreement with the experimental data. The 

investigations have shown some limitations of 1D thermal-hydraulic coarse models of the 

downcomer, lower, and upper plenum. Hence, multidimensional thermal-hydraulic models are 

needed for a more realistic description of the coolant mixing phenomena within the reactor 

pressure vessel. 

FOREWORD

This validation report presents the investigations performed by the Forschungszentrum 

Karlsruhe GmbH to validate the coupled code system RELAP5/PARCS. The plant data used 

for this work was distributed in the frame of the OECD/NEA VVER-1000 Coolant Tansient 

Benchmark. A pump restart-up test was performed at the Kozloduy nuclear power plant 

(KNPP) in Bulgaria. This type of data is very unique for the validation of the neutron 

physics/thermal hydraulic coupled codes. International code validation and benchmarking 

activities are very important to increase the confidence of coupled codes. The CAMPProgram of the US NRC is one of the most important and long lasting activities to foster the 

improvement of thermal hydraulics and coupled neutronic/thermal hydraulic system codes for 

light water reactor safety assessment. 

v 

vii 

TABLE OF CONTENTS

Abstract...............................................................................................................iii 

Foreword.............................................................................................................v 

Executive Summary...............................................................................................xi 

Nomenclature......................................................................................................xiii 

1 Introduction ....................................................................................................................... 1

2 Short Description of the Reference Plant.......................................................................... 3

3 Description of the Scenario............................................................................................... 7

3.1 Pre-Test Plant Conditions ............................................................................................ 7

3.2 Switch-on of the Main Coolant Pump #3 Test .............................................................. 8

4 Developed Plant Models and Nodalisation ....................................................................... 9

4.1. Integral Plant Model ..................................................................................................... 9

4.2. Multidimensional Core Modelling ............................................................................... 11

4.2.1. Neutron kinetics core model................................................................................ 12

4.2.2. Thermal-hydraulic core model............................................................................. 13

5. Performed calculations ................................................................................................... 15

6. Selected results of the calculations................................................................................. 17

6.1. Steady state results for integral plant model .............................................................. 17

6.2. RELAP5/PARCS results for the 3D core model ......................................................... 19

6.3. Transient results for the integral plant model ............................................................. 20

6.3.1. Global plant response ......................................................................................... 20

6.3.2. Code predictions versus experimental data ........................................................ 23

6.3.3. Multidimensional core behaviour......................................................................... 30

7. Results for the extreme scenario .................................................................................... 33

8. Conclusions .................................................................................................................... 37

9. References...................................................................................................................... 39
viii

LIST OF FIGURES

Figure 1. Horizontal Arrangement of the Primary Loops of the VVER-1000 Plant......................... 3 

Figure 2. Vertical Arrangement of the VVER-1000 Primary Components. .................................... 4 

Figure 3. Typical Horizontal Steam Generator of the VVER-1000 Plant. ...................................... 4 

Figure 4. Vertical Cut Through the Reactor Pressure Vessel of the VVER-1000 Plant................. 5 

Figure 5. Core Configuration with the Position of the Different Control Rod Groups..................... 5 

Figure 6 Nodalization of the Kozloduy Plant (RPV with Two Loops) ........................................... 10 

Figure 7. Core-Averaged Axial Power Profile at BOC-Conditions before the test. ...................... 11 

Figure 8. One Sixth of the Core with the Number of FA- and RA-types at BOC.......................... 11 

Figure 9. Radial Arrangement of the Different FA-Types in the Core (Type 1 to 28)................... 12 

Figure 10. Axial Discretisation of Each FA/RA-Type in the Core as Used in PARCS.................. 13 

Figure 11. Coolant Channels Numbering for the Mapping Between RELAP5 and PARCS......... 14 

Figure 12. Axial nodalization of the core (RELAP5 part). ............................................................ 14 

Figure 13. Flow conditions in Downcomer................................................................................... 18 

Figure 14. Flow conditions in Upper Plenum ............................................................................... 18 

Figure 15. Complex Coolant Mixing in the Upper Plenum Predicted by CFX-5........................... 18 

Figure 16. Comparison of the Predicted Axial Power Profile for the HFP Steady State. ............. 20 

Figure 17. Predicted Reverse Flow of the Loop-3 during the Test............................................... 21 

Figure 18. Predicted Change of the Total Primary Mass Flow Rate During the Test. ................. 21 

Figure 19. Predicted Core Averaged Coolant Temperature During the Test............................... 22 

Figure 20. Predicted Power Increase for Exercise-1 and Exercise-3. ......................................... 22 

Figure 21. Predicted Reactivity Increase for Exercise-1 and Exercise-3. .................................... 23 

Figure 22. Predicted Reactivity with Main Contributors. .............................................................. 23 

Figure 23. Comparison of Predicted and Measured Pressure in Upper Head............................. 24 

Figure 24. Comparison of the Predicted and Measured PZR Water Level.................................. 25 

Figure 25. Comparison of Predicted Cold Leg-1 Coolant Temperature with Data....................... 25 

Figure 26. Comparison of Predicted Hot Leg-1 Coolant Temperature with Data......................... 26 

Figure 27. Comparison of Predicted Pressure Drop over MCP-3 with Data................................ 26 

Figure 28. Comparison of Measured and Predicted Water Level of the SG-1............................. 27 

Figure 29. Comparison of Predicted Pressure Drop over MCP1 with Data. ................................ 28 

Figure 30. Comparison of Predicted Collapse Liquid Level with Data. ........................................ 28 

Figure 31. Comparison of Predicted Hot Leg-1Temperature with Data....................................... 29 

Figure 32. Comparison of Predicted Cold Leg-3 Temperature with Data. ................................... 29 

Figure 33. Predicted Core Averaged Axial Power Profile for Three Times.................................. 30 

Figure 34. Predicted 2D Rel. Radial Power per FA at Steady State Conditions.......................... 31 

Figure 35. Predicted Rel. Power in the Core at Steady State Conditions.................................... 31 

Figure 36. Predicted Rel. Radial Power in the Core 15 s After the Transient Initiation................ 32 

Figure 37. Predicted Rel. Radial Power in the Core at 15 s After the Transient Begins.............. 32 

Figure 38. Comparison of Predicted Total Power Change During the Transient......................... 33 

Figure 39. Comparison of the Predicted Core Reactivity for the Two Scenarios......................... 34 

Figure 40. Predicted Relative Radial Power Profile for the Extreme Scenario. ........................... 34 

Figure 41. Predicted Relative Radial Power per Fuel Assembly. ................................................ 35 

Figure 42. Predicted Relative Radial Power per Fuel Assembly at 13.3 s................................... 35 
ix

LIST OF TABLES

Table 1. Dimensions of the Fuel Rod and Fuel Assembly. ............................................................ 6 

Table 2. Measured Plant Data Before the Test with Error Band of Measurements ....................... 7 

Table 3: Comparison of Predicted and Measured Data for Steady State Conditions. ................. 17 

Table 4.HZP Results Obtained with RELAP5/PARCS Compared to Reference Values. ............ 19 

Table 5. Position of the Control Rod Groups for the HFP States (100: out, 0: in). ...................... 19 

Table 6. HFP Results Obtained with RELAP5/PARCS for the Steady State Conditions............. 20 
x
xi

EXECUTIVE SUMMARY

This report presents the validation of the coupled code RELAP5/PARCS using plant data 

obtained during the commissioning phase of the Kozloduy nuclear power plant unit 6 regarding 

the pump-restart test. 

The measured data were made available to a wider community in the frame of the Organization 

for Economic Cooperation and Development/Nuclear Energy Agency/U.S. Nuclear Regulatory 

Commission (OECD/NEA/NRC) VVER-1000 Coolant Transient Benchmark Phase 1. This 

benchmark includes the following exercises: (1) investigation of the integral plant response using 

a best-estimate thermal-hydraulic system code with a point kinetics model, (2) analysis of the core 

response for given initial and transient thermal-hydraulic boundary conditions using a coupled 

code system with 3D-neutron kinetics model, and (3) investigation of the integral plant response 

using a best-estimate coupled-code system with 3D-neutron kinetics. 

Because the reactor was operated at low power with only three main coolant pumps in operation, 

complex flow conditions existed within the reactor pressure vessel (RPV) already before the test 

began. These conditions were characterized by coolant mixing in downcomer and upper plenum 

caused by the reverse flow through the loop-3 (pump was not in operation). The test was initiated 

by switching on the main coolant pump of loop-3 leading to a reversal of the flow through the 

respective loop. After about 15 s the mass flow rate through this loop reaches values comparable 

with the one of the other loops. During this time period, the increased primary coolant flow 

causes a reduction of the core-averaged coolant temperature and thus an increase of the core 

power. Later on, the power stabilizes at a level higher than the initial power. 

 

In this analysis, special attention is paid to the prediction of the spatial asymmetrical core cooling 

during the test and its effects on the local power distribution within the core. The code’s 

predictions are strongly influenced by the way the analyst models the coolant mixing by means of 

1D thermal-hydraulic codes. 

An integral plant model—including the relevant primary and secondary systems as well as a 

detailed model of the reactor pressure vessel and of the core—was developed for RELAP5 and 

PARCS to describe the key phenomena within the reactor pressure vessel appropriately. 

 

Selected results of these investigations will be presented and discussed. A comparison of the 

thermal-hydraulic data obtained during the tests with the code’s predictions showed that despite 

the use of the 1D coarse model, the overall trends of most plant parameters are in a reasonable 

agreement with the experimental data. In addition, multidimensional thermal- hydraulic models 

are clearly needed for a more realistic description of the coolant-mixing phenomena within the 

reactor pressure vessel. The coupled code RELAP5/PARCS is able to catch the physics of the 

investigated test as well as to predict the behavior of the core in case of nonsymmetrical situations 

in sufficient detail (i.e., at the fuel assembly level that represents a step forward compared to the 

point kinetics models). But additional model development is necessary for the prediction of local 

safety parameters based on pin level. 
xii
xiii

NOMENCLATURE

RELAP5 Reactor loss-of-coolant analysis program 

PARCS Purdue advanced reactor core simulator 

VVER Water-water energy reactor 

HFP Hot full power 

HZP Hot zero power 

FA Fuel assembly 

RA Reflector assembly 

MCP Main coolant pump 

BOC Beginning of cycle 

PZR Pressurizer 

PWR Pressurized water reactor 

NPP Nuclear power plant 

EFPD Effective full power days 

NEA Nuclear Energy Agency 

NRC U.S. Nuclear Regulatory Commission 

OECD Organization for Economic Cooperation and Development 
0 
1

1 Introduction 

The Karlsruhe Research Center (FZK) is involved in the overall qualification of computational 

tools for the safety evaluation of nuclear power plants of different design to improve their 

prediction capability and acceptability. In this framework, the code RELAP5/PANBOX was 

qualified within the Organization for Economic Cooperation and Development/Nuclear Energy 

Agency (OECD/NEA) pressurized-water reactor (PWR) main steam line break (MSLB)-

benchmark [San00]. As continuation of this work, partly as a contribution to the international 

Code Assessment and Maintenance Program (CAMP) of the U.S. Nuclear Regulatory 

Commission (NRC), the coupled code system RELAP5/PARCS is being validated. The PARCS 

capabilities for quadratic fuel assembly geometry have been qualified in the frame of both the 

PWR TMI-1 main steam line break (MSLB) [Kozl00] and the boiling-water reactor (BWR) Peach 

Bottom Turbine Trip (PBTT) [Bous04] benchmarks. FZK is especially interested in the new 

PARCS capability [Joo02] for hexagonal geometries. These aspects are important not only for 

the VVER-type light-water reactor (LWR) but also for innovative reactor concepts. The 

international OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 1 is an excellent 

opportunity to validate the overall simulation capability of RELAP5/PARCS regarding both the 

thermal-hydraulic plant response (RELAP5) using measured plant data and the neutron physics 

models (PARCS). The Phase 1 of this benchmark is devoted to the analysis of the pump restart 

test while the other three pumps are in operation and covers following three exercises. Exercise-1 

is an investigation of the integral plant response using a best-estimate thermal-hydraulic system 

code with a point kinetics model. Exercise-2 is an analysis of the core response for given initial 

and transient thermal-hydraulic boundary conditions using a coupled code system with a threedimensional (3D)-neutron kinetics model. Exercise-3 is an investigation of the integral plant 

response using a best-estimate coupled code system with 3D-neutron kinetics. For the analysis of 

these exercises, the following steps are followed: 

• Development of an integral model of the Kozloduy nuclear power plant (NPP) including all 

major systems for RELAP5. 

• Development of a 3D core model for RELAP5/PARCS. 

• Integration of the above developed models in one integral coupled model to investigate the 

plant response. 

 

This report presents both the plant components and the developed models. The performed 

investigations are described and the main results are discussed in detail by comparing the 

predictions with the measured data. Final conclusions are drawn and identified points for further 

investigations are outlined. 
2 
3

2 Short Description of the Reference Plant 

The Kozloduy Plant unit 6 was selected as the reference plant for the Benchmark. It is a Russian 

design VVER-1000 reactor of type W320 with a thermal power of 3,000 MWth. The plant 

consists of four loops, each one with a horizontal steam generator (SG) and a main coolant pump 

(MCP) [Ivan02a]. The reactor is equipped with only one turbine. Figure 1 and Figure 2 present 

details of the layout and construction of the plant. As shown in Figure 3, the horizontal steam 

generator is characterized by a large water inventory on the secondary side compared to westerndesigned vertical steam generators. The reactor pressure vessel design also differs from that of the 

western pressurized-water reactor (PWR), especially because of the constructive peculiarities in 

the lower and upper plenum that strongly impact the flow patterns during normal and accidental 

situations. Figure 4 shows a vertical cut through the reactor pressure vessel (RPV). The lower 

plenum consists of an elliptical cone with many perforations that result in a narrowing gap in the 

direction of the central RPV-point. In addition, 163 support columns are present in the lower 

plenum where the lower part is a full slab and the upper part is a tube with wall perforations of 

different sizes. Hence, the flow coming from the downcomer has to pass through very complex 

flow paths to enter into the core. In the upper plenum, two concentric cylinders with perforations 

are present where the lower part of the outer cylinder is conic. 

As shown in Figure 5, the core consists of 163 fuel assemblies (FA) and 48 reflector assemblies 

(RA). Each fuel assembly has 312 fuel pins that are characterized by a central hole of 0.7 mm 

diameter. Table 1 presents the main data about the FA and fuel rod design. 

Figure 1. Horizontal Arrangement of the Primary Loops of the VVER-1000 Plant. 
4

Figure 2. Vertical Arrangement of the VVER-1000 Primary Components. 

Figure 3. Typical Horizontal Steam Generator of the VVER-1000 Plant. 

1. Vessel 

2. Drainage Nozzle 

3. Blow Down Nozzle 

4. Heat-Exchange Tubes 

5. Separation Units 

6. Main Feedwater Spray Unit

7 Gas Removal Nozzle 

8 Emergency Feedwater Spray Unit 

9 Steam Nozzle 

10 Steam Header 

11 Emergency Feedwater Nozzle 

12 Access Airlock
5

Figure 4. Vertical Cut Through the Reactor Pressure Vessel of the VVER-1000 Plant.

Upper Block

Main Seal Parts

Temperature Templates

Shielding Tubes Block 

Spacer Ring 

Barrel 

Outlet Nozzle

Reactor Vessel 

Inlet Nozzle 

Neutron Templates

Core Basket

Fuel Assembly

Console

Figure 5 shows the 10 control rod groups (I to X) are arranged in the core symmetrically. The 

main coolant pumps (MCP-1 to MCP-4) are located asymmetrically with respect to the main axis 

I-III. The fuel pins are arranged in a triangle within the FA, where the central position is occupied 

by an empty rod (instrumentation). In addition, the fuel pins have a central hole of about 1.4 mm 

diameter while the western-type pins are a full slab. 

Figure 5. Core Configuration with the Position of the Different Control Rod Groups. 

1 2 3 4 5 6 

7 8 9 

IV 

10 11 

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88

I 

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20°30’

34°30’

34°30’ 

20°30’ 

MCP 1 

MCP 2

MCP 3

MCP 4 

1 2 3 4 5 6 

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20°30’

34°30’

34°30’ 

20°30’ 

MCP 1 

MCP 2

MCP 3

MCP 4 
6

Table 1. Dimensions of the Fuel Rod and Fuel Assembly.

Parameter Value 

Pellet diameter, mm 7.56 

Central void diameter, mm 1.4 

Clad diameter (outside), mm 9.1 

Clad wall thickness, mm 0.69 

Fuel rod total length, mm 3837 

Fuel rod active length (cold state), mm 3530 

Fuel rod active length (hot state), mm 3550 

Fuel rod pitch, mm 12.75 

Fuel rod grid Triangular 

Number of guide tubes 18 

Guide tube diameter (outside), mm 12.6 

Guide tube diameter (inside), mm 11.0 

Number of fuel pins 312 

Number of water rods/assembly 1 

Water rod diameter (outside), mm 11.2 

Water rod diameter (inside), mm 9.6 

FA wrench size, mm 234 

FA pitch, mm 236 
7 

3 Description of the Scenario 

3.1 Pre-Test Plant Conditions 

The pump restart test (switch-on MCP-3) was performed during the decommissioning phase 

(beginning of cycle [BOC]) of the NPP Kozloduy, Unit 6, to investigate the behaviour of the 

plant. The pre-test phase started reducing the power from 75 percent to about 21 percent by 

consecutive switching off of MCP-2 and MCP-3. Afterwards, the MCP-2 was switched on some 

hours before the test. Then the reactor power stabilized at around 27.47 percent of the nominal 

power (i.e., 824 MWth). The average core exposure amounted to 30.7 effective full power days 

(EFPD). Before the test, the reactor is operated with only three main coolant pumps (MCPs) at a 

thermal power of 824 MWth while the MCP-3 is nonoperable. Under such conditions, part of the 

coolant injected into the downcomer by the three pumps is flowing back through the piping of the 

affected loop-3. This results in a considerable mixing of cold and hot coolant in the upper 

plenum. 

Table 2 summarizes the main plant parameters of the primary and secondary system just before 

the test. 

Table 2. Measured Plant Data Before the Test with Error Band of Measurements

 Unit Data Accuracy 

Thermal core power MW 824 ± 60 

RCS mass flow rate Kg/s 13611 ± 800 

Primary side pressure MPa 15.60 ± 0.3 

Sec. side pressure MPa 5.94 ± 0.2 

Cold leg temp. loop-1 °K 555.55 ± 2 

Cold leg temp. loop-2 °K 554.55 ± 2 

Cold leg temp. loop-3 °K 554.35 ± 2 

Cold leg temp. loop-4 °K 555.25 ± 2 

Hot leg temp. loop-1 °K 567.05 ± 2 

Hot leg temp. loop-2 °K 562.85 ± 2 

Hot leg temp. loop-3 °K 550.75 ± 2 

Hot leg temp. loop-4 °K 566.15 ± 2 

Mass flow rate loop-1 Kg/s 5031 ± 200 

Mass flow rate loop-2 Kg/s 5069 ± 200 

Mass flow rate loop-3 Kg/s -1544 ± 200 

Mass flow rate loop-4 Kg/s 5075 ± 200 

PZR water level m 7.44 ± 0.15 

Water level in SG-1 m 2.30 ± 0.075 

Water level in SG-2 m 2.41 ± 0.075 

Water level in SG-3 m 2.49 ± 0.075 

Water level in SG-4 m 2.43 ± 0.075 

DP over core MPa 0.225 ± 0.2 

DP over MCP-1 MPa 0.492 ± 0.2 

DP over MCP-2 MPa 0.469 ± 0.2 

DP over MCP-3 MPa 0.179 ± 0.2 

DP over MCP-4 MPa 0.500 ± 0.2 
8

3.2 Switch-on of the Main Coolant Pump #3 Test 

The test starts by switching on the MCP-3. Immediately, the reverse flow through the loop-3 

goes to zero within 15 s leading to an increase of the total mass flow rate through the core. The 

increased coolant inventory leads to a decrease of the coolant temperature and to an increase of 

the coolant density. Hence, the core power undergoes initially a rapid increase stabilizing later on 

at a power level higher than the initial power. Because of the coolant mixing in the downcomer, 

the mass flow rate of loop-1, -2, and -4 slightly decreases while the mass flow rate of loop-3 

greatly increases until around 15 s. Afterwards, the mass flow rates of all four loops are similar. 

These modifications of the coolant stream influence the heat transfer across the steam generators 

leading, for example, to an increase of the water level of steam generators 1, 2, and 4 and to 

nonsymmetrical core cooling. The test is characterized by interactions between the core 

neutronics and the system thermal hydraulics. The experiment lasted for 130 s., In this time, most 

of the important primary and secondary thermal hydraulic parameters of the plant were measured. 
9

4 Developed Plant Models and Nodalisation 

To study the plant response during the main coolant pump (MCP)-switching-on test with system 

codes, both the primary and secondary plant systems—including the safety and control systems—

need to be represented in the numerical model. All data needed to develop the respective models 

for PARCS and RELAP5 were taken from the benchmark specifications [Ivan02a]. Based on this 

information, three models were elaborated—an integral plant model starting with a point kinetics 

approach (Exercise-1), a three-dimensional (3D) neutron kinetics and thermal-hydraulic core 

model (Exercise-2), and an integral plant model with a 3D core model (Exercise-3). Details of 

these models will be described in the following subchapters. 

4.1. Integral Plant Model 

Figure 6 Nodalization of the Kozloduy Plant (RPV with Two Loops) shows the integral plant 

model developed for Exercise-1 and Exercise-3, where only two of the four loops are exhibited. 

This model represents the most relevant primary, secondary, and safety systems of the Kozloduy 

plant [Metz03]. For Exercise-1, a point kinetics model was implemented. The Core (volumes 

845 and 843) is represented by two parallel volumes, one representing the core average channel 

and the other one the core bypass. The downcomer (volumes 108, 208, 308, and 408) is 

represented in four equal parts, each one connected to one loop so that the complex flow 

conditions prevailing during the pre-test phase and during the first 15 s of the transient are 

simulated appropriately. The primary circuit consisting of the piping system (loop-1: volumes 

146, 140, 141, 142, 144, and 145); the pumps (volumes 144, 244, 344, and 444); and the steam 

generator tubes (SG-1 volumes: 120, 121, and 122) is fully incorporated in the model. In 

addition, the pressurizer (PZR) with the four groups of heaters is included in the model. 

Moreover, the make-up and drainage system is included in the model. The full data of the 

Russian-type MCPs are taken from the specifications. Each steam generator (SG) consists of 

11,000 tubes that are horizontally arranged between the hot and cold collector tubes. They are 

vertically grouped in three units associated to the primary and secondary volumes. 

The secondary side of the steam generators is characterized by a complex 3D flow inside this big 

volume. The back flow in the SG-downcomer (SG-1 volumes: 150, 151, 152, and 109) also is 

represented in the model (Figure 6). The feedwater system (SG-1: volume 190) is simply 

modelled by a volume providing a constant coolant mass flow with the predefined coolant 

temperature. No emergency feed water system is considered in the model because these systems 

are not expected to be activated during the transient. The steam lines (loop-1 volumes: 181, etc.) 

are in detail modelled including the valves, common header, turbine stop valves and the 

associated safety steam valves, and the steam dump valve groups. The core fuel pins, the steam 

generator tubes, the PZR-heaters as well as the walls of all relevant primary and secondary 

systems (RPV, cold and hot legs, steam generator shell) are considered in the model as heat 

structure components with its respective heat transfer area, heater diameter, material data, and 

heat source when available. They are connected to the corresponding fluid volumes via 

convective boundary conditions. In the point kinetics model, the given neutron physical data 

characterizing the VVER-1000 fuel like prompt neutron lifetime, effective fraction of delayed 

neutrons, decay constants of delayed neutrons, axial power profile, moderator and Doppler 

reactivity coefficients are implemented. The Doppler feedback is calculated using the following 

Doppler temperature (TDoppler) instead of the volume-averaged fuel temperature in all three 

exercises: 
10

0.7 0.3 surface center T T Doppler =⋅ +⋅ fuel Tfuel . 

The moderator and Doppler reactivity coefficients for this model were given in the Specification 

as follows: 

• Moderator temperature coefficient (MTC): -4.2652 e-3 $/K 

• Doppler temperature coefficient (DTC): -2.2853 e-3 $/K 

The effective fraction of delay neutrons (ßeff) amounts 0.007268. The axial power profile was 

predicted by a 3D-neutronic calculation performed by the benchmark team. Figure 7 shows the 

axial power shape at the BOC-conditions before the test. 

Figure 6 Nodalization of the Kozloduy Plant (RPV with Two Loops) 
11

Figure 7. Core-Averaged Axial Power Profile at BOC-Conditions before the test. 

0

0,2

0,4

0,6

0,8

1

1,2

1,4

0,0 0,4 0,7 1,1 1,4 1,8 2,1 2,5 2,8 3,2 3,6

Core Height (m)

FAX

4.2. Multidimensional Core Modelling 

For Exercise-2 and Exercise-3, a multidimensional core model is needed for both the thermalhydraulic as well as the neutronic representation of the core for the coupled code system 

RELAP5/PARCS. This core model is developed for Exercise-2, where only the core behavior is 

evaluated for given boundary and initial conditions at the core inlet and outlet. Later on this 

model is fully merged with the integral plant model so that the test can be analyzed with the 

coupled code system RELAP5/PARCS.

Figure 8 shows that 28 types of fuel assemblies and 2 additional reflector assemblies (radial and 

axial) are in the core. In addition, the figure shows the enrichment of the pins varies from 2 up to 

3.3 percent. 

Figure 8. One Sixth of the Core with the Number of FA- and RA-types at BOC. 

1 

1.219 

2 

1.283 

3 

1.311 

4

1.553 

5

1.360 

6

1.409

7 

1.391

29 

0.0 

8 

1.603 

9 

1.271 

10 

1.286 

11 

1.676 

12 

1.578

13 

0.909

29 

0.0 

14 

1.271 

15 

1.213 

16 

1.271 

17 

1.246

18 

1.065

29 

0.0 

19 

1.286 

20 

1.271 

21 

1.489 

22 

1.053

29

0.0 

23 

1.676 

24 

1.246 

25 

1.053

29 

0.0 

26 

1.578 

27 

1.065 

29 

0.0 

28 

0.909 

29 

0.0 

29 

0.0 

1

2

1 – Type of fuel assembly

2 – Burnup MWd/kgU

Fuel assembly with 

Enrichment 2.0% 

Fuel assembly with 

Enrichment 3.0% 

Fuel assembly with 

Enrichment 3.3% 

Fuel assembly with 

Enrichment 3.3% * 

Radial reflector 

1 

1.219 

2 

1.283 

3 

1.311 

4

1.553 

5

1.360 

6

1.409

7 

1.391

29 

0.0 

8 

1.603 

9 

1.271 

10 

1.286 

11 

1.676 

12 

1.578

13 

0.909

29 

0.0 

14 

1.271 

15 

1.213 

16 

1.271 

17 

1.246

18 

1.065

29 

0.0 

19 

1.286 

20 

1.271 

21 

1.489 

22 

1.053

29

0.0 

23 

1.676 

24 

1.246 

25 

1.053

29 

0.0 

26 

1.578 

27 

1.065 

29 

0.0 

28 

0.909 

29 

0.0 

29 

0.0 

1

2

1 – Type of fuel assembly

2 – Burnup MWd/kgU

Fuel assembly with 

Enrichment 2.0% 

Fuel assembly with 

Enrichment 3.0% 

Fuel assembly with 

Enrichment 3.3% 

Fuel assembly with 

Enrichment 3.3% * 

Radial reflector 

In the Benchmark Specifications, the 29-FA and -RA-types are subdivided into 12 axial 

elevations of equal height but of different material composition (10 nodes for the core region and 

two for the upper and lower reflector). Hence, the neutronic data needed for 3D-core calculations 

were prepared for a total of 283/260 unrodded/rodded material compositions. 
12

Each one is characterized by unique material properties like enrichment, density as well as 

burnup, absorber rod history, and spectral history. The Benchmark team delivered these crosssection data to all participants as look-up tables. All neutronic parameters for two energy groups 

such as diffusion coefficients, scattering, absorption, and fission macroscopic cross sections, 

assembly discontinuity factors, etc. are given in these libraries. The look-up tables provide 

additional information about the delayed neutron fractions, decay constants, and neutron velocity. 

PARCS has several models to read in different formats of cross-section libraries. The nuclear 

data for the fuel assemblies with absorber rods are delivered in an additional library containing 

260 material compositions to account for absorber rods movement. 

4.2.1. Neutron kinetics core model 

The 3D neutron kinetics model (PARCS) is coupled to RELAP5 via PVM [Bar98]. It solves the 

time-dependent neutron diffusion equation in two or more energy groups using the triangular 

polynomial expansion (TPEN) method [Joo02].

Figure 9 shows that, in the core model, eight hexagonal fuel assembly (FA) rings and a reflector 

assembly (RA) ring are considered to describe the whole core, where each assembly represents a 

numerical node in the radial plane. Figure 10 shows that, axially, all assemblies are subdivided 

into 12 equal nodes, one for the lower and upper reflector and 10 for the fuel zone. A total of 283 

material compositions (unrodded) are considered for this problem. The look-up tables are 

functions of fuel temperature and coolant density. A suitable parameter range of these variables 

was selected to cover the expected parameter changes for the steady state and during the transient 

progression. PARCS uses a multidimensional interpolation scheme for the online update of the 

cross sections during the transient phase in dependence of the actual parameters. 

Figure 9. Radial Arrangement of the Different FA-Types in the Core (Type 1 to 28). 
13

Figure 10. Axial Discretisation of Each FA/RA-Type in the Core as Used in PARCS 

4.2.2. Thermal-hydraulic core model 

Figure 11 shows the thermal-hydraulic core model for the coupled calculations consists of 18 

parallel channels that are associated with the FA-nodes according to the mapping scheme 

proposed in the specification. According to this, the whole core is in radial direction divided into 

19 thermal-hydraulic channels—18 for the core region and 1 for the radial reflector. All fuel 

assemblies with the same number are associated to one thermal-hydraulic channel. An additional 

channel is considered (channel 19) to represent the flow area of the 48 reflector assemblies (RA). 

Figure 12 shows that, in axial direction, the parallel channels are subdivided in 12 nodes—the 

bottom and top nodes for the axial reflector and the remaining 10 nodes for the active core. The 

additional fluid volumes at core inlet and outlet are also represented because they are needed to 

define the initial and boundary conditions for Exercise-2. At the core inlet (volume 601 up to 

619), the mass flow rate and the coolant temperature are given. The system pressure is defined at 

the core outlet (volume 800). In the RELAP5-model, 18 heat structures components representing 

the 18 groups of FA are modelled. They are linked to the 18 core channels by convective 

boundary conditions. These heat structures have the same axial nodalization like the 

corresponding fluid channels. In radial direction each heat structure is subdivided in 7 zones, 4 in 

the fuel, one gap and two in the cladding material. 
14

Figure 11. Coolant Channels Numbering for the Mapping Between RELAP5 and PARCS.

15 9 9 3 3 3 6 6 6 6 12 12 18

14 14 8 8 2 2 2 1 6 6 12 12 18 18

14 14 8 8 2 2 1 1 6 12 12 18 18

14 14 8 8 2 1 1 1 12 12 18 18

14 14 8 8 7 7 7 7 12 18 18

14 14 8 7 7 7 7 7 18 18

14 14 13 13 13 13 13 13 18

13 13 13 13 13 13

15 15 9 9 3 3 4 5 5 5 11 11 17 17

15 15 9 10 10 10 10 11 11 17 17

15 15 9 9 4 4 4 5 11 11 17 17

15 15 9 9 3 4 4 5 5 11 11 17 17

15 15 10 10 10 10 10 11 17 17

15 16 16 16 16 16 16 17 17

16 16 16 16 16 16

19

The multidimensional neutron kinetics and thermal hydraulics core model was entirely 

incorporated in the integral plant model to perform the coupled calculations for Exercise-3. 

Figure 12. Axial nodalization of the core (RELAP5 part).
15

5.Performed calculations 

According to the main goal of this benchmark, the following investigations were performed: 

• Exercise-1: Integral plant simulation with a system code using point kinetics approach 

(RELAP5/MOD3.3) to demonstrate that the developed plant model is well balanced and that 

the main steady state plant parameters are well predicted compared to the plant data. 

• Exercise-2: Multidimensional core simulation with the coupled system code 

(RELAP5/PARCS) to test the operability of the coupled code system such as the correctly 

reading the cross-sections (look-up tables), the appropriateness of the interpolation scheme for 

the cross-section update, the convergence of both the neutronics, and the thermal-hydraulic 

model in a coupled calculation, etc. 

• Exercise-3: Integral plant simulation with a coupled system code (RELAP5/PARCS) to 

simulate the overall plant response, especially the space-time core behaviour for the reference 

scenario. 

The numerical simulation of the MCP-switch-on test with the coupled system RELAP5/PARCS 

was performed on a LINUX platform with PVM-environment including the following steps: 

• Run RELAP5-stand alone with the null transient option for some 200 s until stable thermal 

hydraulic plant conditions are reached. 

• Run the RELAP5/PARCS coupled system with the steady state option until the eigenvalue 

calculation of PARCS converged. 

• Then run coupled system code with the transient option restarting from the latter steady-state 

condition for both codes until 130 s (duration of the test). 

In the subsequent chapter, selected results of the performed calculations will be presented. 
16 
17

6.Selected results of the calculations 

6.1. Steady state results for integral plant model 

As part of the qualification of the initial steady state, attention was paid also to the primary circuit 

power balance (i.e., the difference between the core power plus main coolant pump (MCP) power 

and the thermal power transferred over all four steam generators, which amounts around 12 MW. 

Table 3 shows the good agreement between data and predictions using RELAP5 and 

RELAP5/PARCS for the stationary plant conditions before the test demonstrates that the 

developed integral plant model is appropriate for the subsequent study of the plant response. 

Table 3: Comparison of Predicted and Measured Data for Steady State Conditions.

Parameter Benchmark Data RELAP5 Deviation R5/PARCS 

 Unit Data Accuracy Exercise-1 % Exercise-3 

Thermal MW 824 ± 60 824 0 824 

RCS mass Kg/s 13611 ± 800 13577 -0.25 13612 

Primary side MPa 15.60 ± 0.3 15.62 0.13 15.62 

Sec. side MPa 5.94 ± 0.2 6.105 2.83 6.106 

Cold leg °K 555.55 ± 2 555.43 -0.02 555.44 

Cold leg °K 554.55 ± 2 554.61 0.01 554.62 

Cold leg °K 554.35 ± 2 554.94 0.11 554.95 

Cold leg °K 555.25 ± 2 555.16 -0.02 555.17 

Hot leg °K 567.05 ± 2 566.18 -0.15 566.19 

Hot leg °K 562.85 ± 2 563.71 0.15 563.72 

Hot leg °K 550.75 ± 2 550.65 -0.02 550.66 

Hot leg °K 566.15 ± 2 565.43 -0.13 565.43 

Mass flow Kg/s 5031 ± 200 5021 -0.20 5029 

Mass flow Kg/s 5069 ± 200 5036 -0.65 5043 

Mass flow Kg/s -1544 ± 200 -1503 -2.66 -1491 

Mass flow Kg/s 5075 ± 200 5034 -0.81 5041 

PZR water m 7.44 ± 0.15 7.44 0.00 7.44 

Water level m 2.30 ± 0.075 2.305 0.22 2.304 

Water level m 2.41 ± 0.075 2.409 -0.04 2.409 

Water level m 2.49 ± 0.075 2.439 -2.05 2.439 

Water level m 2.43 ± 0.075 2.458 1.15 2.457 

DP over core MPa 0.225 ± 0.2 0.2570 14.22 0.255 

DP over MPa 0.492 ± 0.2 0.4845 -1.52 0.4825 

DP over MPa 0.469 ± 0.2 0.4818 2.73 0.4798 

DP over MPa 0.179 ± 0.2 0.1811 1.17 0.1787 

DP over MPa 0.500 ± 0.2 0.4824 -3.52 0.4804 

It can be seen that the deviation of most parameters is very small for both calculations. Some of 

the parameters are slightly underpredicted, and others are slightly overpredicted by both 

calculations. This good agreement between data and predictions for the stationary plant conditions 

before the test demonstrates that the developed integral plant model is appropriate for subsequent 

studies of the plant response. A prerequisite for the successful simulation of the plant conditions 

before the test was the appropriate modelling of the downcomer, lower, and upper plenum of the 

VVER-1000 reactor that was realized based on in-house CFD-simulations [Boett04]. 
18

As can be seen in Figure 13 and Figure 14, the flow conditions in the downcomer and the upper 

plenum are rather complex. To catch the physical phenomena, the downcomer was subdivided 

into four parallel volumes that are connected to each other by cross-flow junctions. To assess the 

flow conditions in the upper plenum, especially the distribution of the reverse flow of about 1,500 

kg/s entering into the outer ring between the RPV-wall and the inner perforated shell, an isolated 

CFX-model was developed for the upper plenum [Boett04]. Figure 15 shows the predicted flow 

redistribution in the upper plenum, especially in the outer ring as obtained by a CFX simulation. 

It can be seen that a considerable part of the cold backflow of loop-3 is flowing sideward in the 

outer ring to the outlet orifice of loop-2 since loop-3 and loop-2 are located next to each other. A 

minor part of the coolant of loop-3 is also sideward redirected into the outlet orifice of loop-4. It 

should be noted that the rest of the cold flow of the loop-3 is entering the inner volume of the 

upper plenum through the perforations of the inner shell. Based on these results, the fluid volumes 

representing the outer ring of the upper plenum were connected by cross-flow junction in the 

RELAP5 model so that the reverse flow of loop-3 can go through the outlet orifices of loop-2 and 

loop-4. The mixing of a part of the flow of loop-3 in the upper plenum before leaving is also 

allowed. With these model extensions, the prediction of the stationary plant conditions was 

improved. 

Figure 13. Flow conditions in Downcomer Figure 14. Flow conditions in Upper Plenum 

Figure 15. Complex Coolant Mixing in the Upper Plenum Predicted by CFX-5.
19

6.2. RELAP5/PARCS results for the 3D core model 

In the frame of the benchmark Exercise-2, the investigations are focused on the testing of the 3D 

core model regarding the mapping scheme between the thermal-hydraulic and neutronic nodes as 

well as the consistence of the nodal cross-sections. In addition, different neutronic parameters 

like the stuck-rod, shut-down, and tripped rod worth are predicted. Several calculations were 

performed with RELAP5/PARCS for both the hot zero power (HZP) and hot full power (HFP) 

conditions of the Kozloduy core using the multidimensional model described above. The initial 

and boundary conditions as well as the control rod positions used to predict the different reactivity 

worth of the HZP state are taken from the specifications. The HZP conditions are characterized by 

a nominal power of 0.1 percent of the total power and fixed feedback thermal-hydraulics 

conditions (i.e., the moderator density in the core is 767.1 kg/m3 and the fuel temperature amounts 

552.15 K). 

To assess the developed 3D core models, the effective multiplication factor (keff) was predicted 

for different core states defined by different positions of the absorber rod groups. In Table 4, 

neutron physical parameters of the HZP state predicted by RELAP5/PARCS are compared with 

some data from the specification, where predictions and reference values are in a reasonably good 

agreement. The reactivity worth for different HZP-core states calculated by the coupled code is 

compared to the values given in the specifications. Both are close to each other. 

Table 4.HZP Results Obtained with RELAP5/PARCS Compared to Reference Values.

Parameters RELAP5/PARCS Reference Data 

Keff 0.999669

Radial power peaking factor 1.4034

Axial power peaking factor 1.514

Axial offset -0.1726

Ejected rod worth % dk/k 0.078 0.09 

Control rod group 10 worth, -0.69 - 0.61 

Tripped rod worth, % dk/k -7.24 - 7.02 

The HFP core state is characterized by beginning of cycle (BOC) fuel conditions with an average 

exposure of 30.7 effective full power days (EFPD) and a thermal power of 824 MW. For this 

core state, both steady-state and transient calculations were performed with the coupled code 

system using the initial and transient boundary conditions given in the specifications. Table 5

indicates the position of the absorber rod group attained for the HFP state in comparison to the 

HZP state. 

Table 6 summarizes the main neutron physical parameters predicted for the stationary conditions 

of the HFP by RELAP5/PARCS. 

Table 5. Position of the Control Rod Groups for the HFP States (100: out, 0: in).

Core State G1-4 G5 G6-8 G9 G10 G10 

EjRod 

Hot Zero Power (HZP) 100 100 100 64 0 0 

Hot Full Power (HFP) 100 100 1 0 0 100 36 36 
20

Table 6. HFP Results Obtained with RELAP5/PARCS for the Steady State Conditions. 

Calculated Parameters RELAP5/PARCS 

Keff 1.000425

Radial Power Peaking Factor 1.3471

Axial Power Peaking Factor 1.408

Axial Offset -0.1734

The RELAP5/PARCS predicted a non-symmetrical axial power distribution for the steady state 

conditions expressed by an axial offset of 17.34 percent. Figure 16 compares the predicted core 

averaged axial power peaking to the one given in the specification that was calculated by the 

benchmark team (PSU). It can be seen that both curves show the same trends with slight 

deviations mainly around 1 m and 2 m height. 

Figure 16. Comparison of the Predicted Axial Power Profile for the HFP Steady State. 

0

0,5

1

1,5

0 0,5 1 1,5 2 2,5 3 3,5 4

Height (m)

Axial Power Peaking (-)

PSU

RELAP5/PARCS

6.3. Transient results for the integral plant model 

The following subsections present the selected results obtained with both RELAP5/Point Kinetics 

and RELA5/PARCS and compare them to the plant data. Finally, the subsections present and 

discuss the detailed results obtained with PARCS. 

6.3.1. Global plant response 

The transient is initiated by the switch-on of the MCP-3. As a consequence, the mass flow rate of 

the loop-3 starts to re-invert (Figure 17), leading to a continuous increase of the primary coolant 

mass flow (Figure 18). After about 15 s, all loops reached similar mass flow rates, which remain 

almost unchanged during the transient. As a result, the core averaged coolant temperature 

decreases some degrees for the first 15 s (Figure 19). Like the coolant temperature, the fuel 

temperature undergoes the same trend during the first 15 s because the cooling conditions of the 

core have improved. Consequently, the total reactor power increases (Figure 20) rapidly until 

around 15 s because of the moderator and Doppler reactivity feedbacks. 
21

Afterwards, this trend continues until the end of the transient but with a very moderate change 

rate. The power increase predicted with the point kinetics model (Exercose-1) is higher (5.5 

percent of nominal power) than the one predicted with the 3D-neutron kinetics model (Exercise-3, 

3.7 percent of nominal power) at the end of the transient. The reason is the use of a constant axial 

power, peaking Doppler and moderator reactivity coefficients, estimated for the core conditions at 

the beginning of the test for the whole transient. This leads to an overestimation of the reactivity 

inserted into the core by the point kinetics approach. On the contrary, PARCS solves a 3D 

problem with local estimation of the feedbacks by means of using cross-section sets depending on 

local thermal-hydraulic parameters that represent a more realistic description of the underlying 

asymmetrical core behaviour. The predictions of Exercise-1 may improve if a more detailed and 

sophisticated point kinetics model is used for this problem. 

Figure 17. Predicted Reverse Flow of the Loop-3 during the Test. 

-2000

-1000

0

1000

2000

3000

4000

5000

0 20 40 60 80 100 120

T im e (s )

G-Loop3 (K g

Ex -1

Ex -3

12000

13000

14000

15000

16000

17000

18000

19000

20000

0 20 40 60 80 100 120

Time (s)

G (kg/s)

Ex1

Ex3

Figure 18. Predicted Change of the Total Primary Mass Flow Rate During the Test. 
22

559

560

561

562

563

0 20 40 60 80 100 120

Time (s)

T (K)

Ex1

Ex3

Figure 19. Predicted Core Averaged Coolant Temperature During the Test. 

820

830

840

850

860

870

880

0 20 40 60 80 100 120

Time (s)

P (MW)

Ex1

Ex3

Figure 20. Predicted Power Increase for Exercise-1 and Exercise-3. 
23

Figure 21. Predicted Reactivity Increase for Exercise-1 and Exercise-3. 

0

0,004

0,008

0,012

0 20 40 60 80 100 120

Time (s)

Reactivity ($)

Ex1

EX3-BS

Figure 22. Predicted Reactivity with Main Contributors. 

-0,0090

-0,0065

-0,0040

-0,0015

0,0010

0,0035

0,0060

0,0085

0,0110

0,0135

0,0160

0 20 40 60 80 100 120

Time (s)

Reac ($)

ReacMT ReacDT ReacTOT

6.3.2. Code predictions versus experimental data 

During the MCP switch-on test, several parameters of the plant were measured and its error bands 

were estimated (see Table 2). To show the quality of the RELAP5/PARCS predictions, selected 

parameters are chosen and compared to available data. Figure 23 compares the pressure of the 

upper plenum predicted by the codes to the measured values. 
24

The couple code is able to predict the initial time and decrease rate of the pressure during the first 

10 s. Figure 24 shows that, later on, the contraction of the primary system coolant is 

overpredicted by the simulations, which results in a faster decrease of the PZR level (around 55-

80 s in the transient), and both predictions and measured PZR level are very close to each other. 

The primary to secondary heat transfer is almost constant in the calculation, while a slow but 

steady cooldown of the primary system is observed in the measured data. 

Figure 25 and Figure 26 show the changes of coolant temperature of the loop-1 for both cold and 

hot legs predicted by the codes in comparison with the measurements data. 

It must be noted that the changes of the coolant temperature are moderate and smaller than the 

error band of the temperature measurement devices. The overall trend of the measured data can 

be reproduced by the calculation. The predictions tend to estimate a larger variation of the 

coolant temperature than one shown by the data. In Figure 27, the measured pressure drop over 

the MCP of loop 3 is compared with the values predicted by the code. The agreement is quite 

good for the whole transient. On the secondary side, the measured variation of the water level in 

the steam generator 1 is also compared with the predicted one in Figure 28. Although prediction 

and data start from around the same level, the code overestimates the heat transferred to the 

secondary side for the first 15 s. Then, the primary-to-secondary heat transfer is underestimated 

until the end of the transient. 

15,3

15,35

15,4

15,45

15,5

15,55

15,6

15,65

0 20 40 60 80 100 120

Time (s)

P (MPa)

Data-Upln

Ex3

Figure 23. Comparison of Predicted and Measured Pressure in Upper Head. 
25

7,2

7,25

7,3

7,35

7,4

7,45

7,5

0 20 40 60 80 100 120

Time (s)

Lvl (m)

Data 

Ex3

Figure 24. Comparison of the Predicted and Measured PZR Water Level. 

552

552,5

553

553,5

554

554,5

555

555,5

556

0 20 40 60 80 100 120

Time (s)

T(K)

Data-TcldL1 

Ex3

Figure 25. Comparison of Predicted Cold Leg-1 Coolant Temperature with Data 
26

561

562

563

564

565

566

567

568

0 20 40 60 80 100 120

Time (s)

T(K)

Data-ThotL1

Ex3

Figure 26. Comparison of Predicted Hot Leg-1 Coolant Temperature with Data. 

0,15

0,25

0,35

0,45

0,55

0,65

0,75

0 20 40 60 80 100 120

Time (s)

DP (MPa)

Data-MCP3

Ex3

Figure 27. Comparison of Predicted Pressure Drop over MCP-3 with Data. 
27

2,22

2,26

2,3

2,34

2,38

2,42

0 20 40 60 80 100 120

Time (s)

Lvl (m)

Data-SG1

Ex3

Figure 28. Comparison of Measured and Predicted Water Level of the SG-1. 

Figure 29 to Figure 32 compare the measured data of selected parameters to the predictions of 

both the point kinetics model (Exercise-1) and the 3D-kinetics models. The figures show that 

both simulations calculated very similar qualitative global trends. Because the predicted power 

by the point kinetics and 3D kinetics is different, other parameters like the collapsed liquid level 

in the pressurizer are quantitatively different for both simulations. 
28

Figure 29. Comparison of Predicted Pressure Drop over MCP1 with Data. 

Figure 30. Comparison of Predicted Collapse Liquid Level with Data.

0,45

0,49

0,53

0,57

0,61

0,65

0 20 40 60 80 100 120

Time (s)

DP-MCP1 (MPa)

Data

Ex-1

Ex-3

7,20

7,25

7,30

7,35

7,40

7,45

7,50

7,55

7,60

0 20 40 60 80 100 120

Time (s)

PZR-Lvl (m)

Data

Ex-1

Ex-3
29

Figure 31. Comparison of Predicted Hot Leg-1Temperature with Data. 

561

562

563

564

565

566

567

568

0 20 40 60 80 100 120

Time (s)

Thot-loop1 (K)

Data

Ex-1

Ex-3

Figure 32. Comparison of Predicted Cold Leg-3 Temperature with Data. 

550

551

552

553

554

555

556

557

0 20 40 60 80 100 120

Time (s)

Tcld-loop3 (K)

Data

Ex-1

Ex-3
30

6.3.3. Multidimensional core behaviour 

The use of coupled codes with 3D neutron kinetics models allows a more detailed analysis of the 

core response compared to the point kinetics. Figure 33 shows the core averaged axial power 

peaking predicted by PARCS for three time windows during the transient. The figure shows that 

for the basic scenario only a very moderate variation of the power peaking occurs. A similar 

trend was observed when the core averaged radial power profile was analyzed. Figure 34 to 

Figure 37 show that minor changes of the local radial power profile occur at different time 

windows. The maximal relative radial peak power changed slightly from 1.3450 at 0.0 s to 

1.3473 at 15 s. 

Figure 33. Predicted Core Averaged Axial Power Profile for Three Times 

0

0,2

0,4

0,6

0,8

1

1,2

1,4

1,6

0 0,5 1 1,5 2 2,5 3 3,5

Height (m)

Rel. FAX (-)

0 s

15 s

800 s
31

Figure 34. Predicted 2D Rel. Radial Power per FA at Steady State Conditions. 

Figure 35. Predicted Rel. Power in the Core at Steady State Conditions. 
32

Figure 36. Predicted Rel. Radial Power in the Core 15 s After the Transient Initiation. 

Figure 37. Predicted Rel. Radial Power in the Core at 15 s After the Transient Begins. 
33

7. Results for the extreme scenario 

Since the local power distribution during this transient was very moderate, an additional scenario 

(called extreme scenario) was defined in the Specifications to better check the capabilities of 

coupled codes. For this extreme scenario the ejection of a control rod located in the affected core 

sector corresponding to the loop-3 (FA#123) was assumed to happen at 13 s. In Figure 38 a 

comparison of the power evolution during the transient predicted by RELAP5/PARCS is given. It 

can be seen that due to the assumptions the power increase of the extreme scenario is very 

pronounced compared to the one of the basic scenario mainly due to the ejection of the control 

rod group # 10 at 13 s which has a reactivity of 0.95 $. This power increase was stopped mainly 

by the Doppler and also by the moderator reactivity coefficient. Both the fuel temperature and the 

moderator density started to increase after the control rod #10 was ejected out of the core. At the 

time of the highest power core reactivity amounted 0,157 $ and the maximal relative power 

radial/axial power was around 1.605/1.372, see Figure 39. These values are much higher than for 

the basic scenario. A higher reactivity insertion due to assumption of control rod ejection at 13 s 

was predicted by RELAP5/PARCS than for the basic scenario. The detailed 3D radial power 

distribution for the extreme scenario is shown in Figure 40 to Figure 42. Here the fuel 

assemblies with the highest power can be identified very clearly. 

Figure 38. Comparison of Predicted Total Power Change During the Transient. 

820,0

870,0

920,0

970,0

1020,0

1070,0

0 20 40 60 80 100 120

Time (s)

P (MW)

Basic SC

Extreme SC
34

Figure 39. Comparison of the Predicted Core Reactivity for the Two Scenarios. 

0,000000

0,050000

0,100000

0,150000

0 20 40 60 80 100 120

Time (s)

Reac ($)

Basic SC

Extreme SC

Figure 40. Predicted Relative Radial Power Profile for the Extreme Scenario. 
35

Figure 41. Predicted Relative Radial Power per Fuel Assembly. 

Figure 42. Predicted Relative Radial Power per Fuel Assembly at 13.3 s.
36 
37

8. Conclusions 

The Phase 1 of the V1000 Coolant Transient was analyzed with different computational tools 

using both point kinetics and three-dimensional (3D) kinetics models coupled to system codes. It 

was demonstrated that the developed integral plant model as well as the multidimensional core 

model are appropriate to describe the main plant and core response in case of an MCP-restart 

transient. The comparison of the predicted plant conditions with the plant data for the stationary 

reactor conditions showed a good agreement. 

For the transient phase, most of the predicted parameters show trends that are qualitatively in 

good agreement with the experimental data measured during the test. The predicted reactivity 

worth for the different states of the cold zero power is close to the ones given in the 

specifications. Also, the axial power-peaking factor for the steady state hot full power estimated 

by RELAP5/PARCS agreed well with the reference solutions. 

From the multidimensional detailed results for the core, it is apparent that the nonsymmetrical 

spatial power perturbation for the investigated test is rather moderate. But the analysis of the 

extreme scenario showed that the coupled system RELAP5/PARCS is able to predict core 

conditions with pronounced power distortions. 

In general, the performed investigations clearly illustrate the capability of coupled codes systems 

with 3D neutron kinetics models as promising simulation tools to predict local hot spots within 

the core at a fuel assembly level. It can be stated that the time-dependent neutron diffusion 

solution for hexagonal geometries of PARCS works quite well in connection with the thermal 

hydraulic part (RELAP5) and that the RELAP5/PARCS works stable and fast enough under both 

Linux and Windows platforms. 

Finally, the analysis of the MCP switch-on test showed the limitations of 1D thermal-hydraulic 

models of RELAP5 to describe the coolant mixing process. Hence, a more realistic description of 

such transients may only be possible using 3D thermal-hydraulics (3D coarse mesh, subchannel, 

or CFD codes) models coupled with the multidimensional neutron kinetics models. This kind of 

investigation is envisaged for the Phase 2 of this benchmark (V1000CT-2). 

Although PARCS is able to predict the detailed axial and radial peaking factor within the core 

where each fuel assembly is considered as a computational node, the current coupled codes are 

not appropriate to predict the local fuel pin power within the fuel assembly with the maximal 

radial power. Note that exactly these local parameters like the pin power, maximal cladding, and 

fuel temperature are very important to assess the safety margins of any reactor design. Here 

additional model developments are necessary to extend the prediction capability of safety analysis 

codes aiming to implement transport and subchannel codes in the transient solution scheme. 

Summarizing the following statements can be made based on this analysis: 

• The investigations demonstrated the capability of coupled code systems to simulate complex 

transients with tight feedbacks between the system thermal-hydraulics and the core 

neutronics. 

• Both RELAP5/Point Kinetics and RELAP5/PARCS are able to predict the overall plant 
38

response in case of the MCP-3 restart physical sound. 

• RELAP5/Point Kinetics tends to overpredict the reactivity insertion (power increase) partly 

due to the use of fixed axial power distribution for the whole transient time and by using fixed 

reactivity feedbacks coefficients (DTC, MTC). 

• Although the basic scenario is mild in terms of reactivity perturbation, coupled codes like 

RELAP5/PARCS permit the gain of detailed spatial information about hot spots in the core 

(here at fuel assembly level). 

The following areas for further improvements were identified: 

• Implementation of an automatic mapping scheme between the thermal-hydraulics and 

neutronics nodes is needed for RELAP5/PARCS similar to the one of TRACE/PARCS. This 

capability would minimize errors and considerably reduce the tedious work of generating the 

mapping manually. 

• Development of 3D plotting and movies capabilities to represent the PARCS and also the 

RELAP5 results obtained for detailed core models. 

• Correction of the PARCS output regarding important fuel-assembly-related parameters like 

axial power profile, reactivity contribution by moderator, and fuel temperature changes, 

specifically for hexagonal fuel assembly geometries. 

• Implementation of an option in PARCS that allows the user to select the plot frequency. This 

is very important because the output and restart file of PARCS can become very large 

analysing real nuclear power plant problems (in the order of five or more gigabytes). 

• Currently, PARCS auxiliary output files for hexagonal geometries are fixed to MSLB and, in 

addition, the data of many of them are useless. Modifications of these files in a more general 

sense (i.e., case-independent) is desirable: 

o MSLB_TMMAP.DAT (Assembly averaged coolant temperature map at each time step) 

o MSLB_TFMAP.DAT (Assembly averaged fuel temperature map at each time step) 

o MSLB_RPDMAP.DAT (Assembly averaged relative radial power map at each time step) 

o MSLB_KINFMAP.DAT (Assembly averaged Kinf map at each time step) 

o MSLB_DMMAP.DAT (Assembly averaged coolant density map at each time step) 

• Improvement of the convergence between the thermal-hydraulic (RELAP) and the neutron 

kinetics solution (PARCS). 
39

9. References 

[Bar98] Barber, D., Downar, T., and Wang, W. Final Completion Report for the Coupled 

RELAP5/PARCS Code. Report PU/NE-98-31. Purdue University. November 1998 

[Boett04] M. Böttcher. Detailed study of coolant mixing within the reactor pressure vessel of a 

VVER-1000 reactor during a non-symmetrical heat-up test. NED Vol38 Issue 3, March 2008 

p445-452. 

[Bous04] Bousbia-Salah, A., Vedovi, J., D’Auria, F., Ivanov, K., Galassi, G. Analysis of the 

Peach Bottom Turbine Trip 2 Experiment by Coupled RELAP5-PARCS Three-Dimensional 

Codes. Nuclear Science and Engineering, Vol.148, No.2, October 2004. 

[Ivan02a] Ivanov, B., Ivanov, K., Groudev, P., Pavlova, M., Hadjiev, V. V1000-Coolant 

Transient Benchmark PHASE 1(V1000CT-1) Vol.1: Main Coolant Pump (MCP) Switching OnFinal Specifications. NEA/NSC/DOC(2002)6. 

[Joo02] Joo, H. G., Barber, D., Jiang, G., and Downar, T. PARCS: A multidimensional twogroup reactor kinetics code based on the nonlinear analytical nodal method. PARCS Manual 

Version 2.20. Purdue University, School of Nuclear Engineering. July 2002. 

[Kozl00] Kozlowski, T., Miller, R.M., Downar, T.J., Ebert, D. Analysis of the OECD MSLB 

Benchmark with RELAP5/PARCS. PHYSOR-2000. Pittsburgh, Pennsylvania. May 2000. 

[Metz03] Metz, Olivier. Investigations of the VVER-1000 plant behavior during a coolant 

transient by RELAP5. Diplomarbeit FZK/Universität Karlsruhe Fakultät Maschinenbau, Institut 

für Kerntechnik und Reaktorsicherheit. May 2003. 

[San00] Sánchez, V., Hering, W., Knoll, A., Böer, R. Main Steam Line Break Analysis for the 

TMI-1 NPP with the Best-Estimate Code System RELAP5/PANBOX. Annual Meeting on 

Nuclear Technology, 23-25 May 2000. Bonn, Germany.