NUREG/IA-0449 

International 

 Agreement Report

Post-Test Analysis of Upper Plenum 11% 

Break at PSB-VVER Facility using TRACE 

V5.0 and RELAP5/MOD3.3 Code

 

Prepared by:

Petr Heralecky

TES Ltd.

Prazska 597

674 01 Trebic, Czech Republic

 

K. Tien, NRC Project Manager 

 

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: May 2014

Date Published: July 2014

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0449 

International 

 Agreement Report

Post-Test Analysis of Upper Plenum 11% 

Break at PSB-VVER Facility using TRACE 

V5.0 and RELAP5/MOD3.3 Code

 

Prepared by:

Petr Heralecky

TES Ltd.

Prazska 597

674 01 Trebic, Czech Republic

 

K. Tien, NRC Project Manager 

 

 

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Manuscript Completed: May 2014

Date Published: July 2014

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

Published by

U.S. Nuclear Regulatory Commission

iii 

ABSTRACT 

The best estimate thermo-hydraulic computer code TRACE V5.0 and RELAP5 MOD3.3 

hadbeen assessed using Upper Plenum 11% break experiment at the large-scale test facility 

PSB-VVER. The PSB-VVER facility is a 1:300 volume scaled model of VVER 1000 NPP located 

in Electrogorsk, Russia. An extensive TRACE and RELAP5 input decks of PSB-VVER facility 

were developed including all important components of the PSB-VVER facility: reactor, 4 

separated loops, pressurizer, break units, main circulation pumps, steam generators, and 

important parts of secondary circuit. The TRACE (TRAC/RELAP Advanced Computational 

Engine) is the latest in a series of advanced, best-estimate reactor systems codes developed by 

the U.S. Nuclear Regulatory Commission in frame of CAMP (Code Application and Maintenance 

Program) and the RELAP5 code is its predecessor. The TRACE and RELAP5 codes are a 

component-oriented reactor systems analysis codes designed to analyze light water reactor 

transients up to the point of significant fuel damage. The original validation of both codes was 

mainly based on experiments performed on experimental facilities of typical PWR design. There 

are some different features of VVER design comparing to PWR. Therefore the validation of the 

thermo-hydraulic codes for VVER types of reactors is often required by national regulators. The 

purpose of performed analysis is to extend the validation of the TRACE and RELAP5 code 

focused on VVER type of NPPs. The TRACE calculation was performed in the frame of R&D 

project co-funded by The Ministry of Industry and Trade of Czech Republic. The RELAP5 

calculation was performed to support standardization of the RELAP5 code in TES Company. 

v 

CONTENTS

Page 

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

FIGURES ...................................................................................................................... vii 

TABLES ......................................................................................................................... ix 

ACKNOWLEDGMENTS ................................................................................................ xi

ABBREVIATIONS ..................................................................................................... ...xiii

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

2. FACILITY AND TEST DESCRIPTIONS ................................................................. 2-1

2.1 PSB-VVER Facility ......................................................................................................... 2-1

2.2 Experiment UP-11-07 ..................................................................................................... 2-5

2.2.1 Facility configuration ........................................................................................... 2-5

2.2.2 Initial Conditions ................................................................................................. 2-5

2.2.3 Boundary Conditions (test scenario) .................................................................. 2-6

3. THE TRACE AND RELAP5/MOD3.3 CODES ........................................................ 3-1

3.1 The TRACE and RELAP5/MOD3.3 code assessment ................................................... 3-2

3.2 VVER typical features related to TRACE and RELAP5/MOD3.3 code assessment ...... 3-2

4. INPUT DECK DESCRIPTION ................................................................................. 4-1

4.1 The TRACE input deck ................................................................................................... 4-1

4.2 The RELAP5 input deck ................................................................................................. 4-4

5. RESULTS ............................................................................................................... 5-1

5.1 TRACE Steady-state calculation .................................................................................... 5-1

5.2 RELAP5 Steady-state calculation .................................................................................. 5-1

5.3 TRACE transient calculation .......................................................................................... 5-2

5.4 RELAP5 transient calculation ......................................................................................... 5-5

5.5 Quantitative Assessment of the Calculations ................................................................. 5-7

6. RUN STATISTICS .................................................................................................. 6-1

7. CONCLUSIONS ..................................................................................................... 7-1

8. REFERENCES ....................................................................................................... 8-1
vi 

Appendix A: INPUT DECK NODALISATION SCHEMES...............................................A-1

Appendix B: MEASUREMENT LOCALISATION AT PSB-VVER FACILITY.......................B-1 

Appendix C: COMPLETE SET OF COMPARISON PLOTS FOR RELAP5 CALCULATION..C-1 

Appendix D: COMPLETE SET OF COMPARISON PLOTS FOR TRACE CALCULATION...D-1 
vii 

FIGURES

Page 

Figure 1: General View of PSB-VVER Facility ....................................................................... 2-2

Figure 2: Upper Plenum Break Unit ....................................................................................... 2-3

Figure 3: Discharge line ......................................................................................................... 2-4

Figure A-1: TRACE Nodalization Scheme of Primary Circuit of PSB-VVER Facility ............... A-2

Figure A-2: TRACE Nodalization Scheme of Secondary Circuit of PSB-VVER Facility .......... A-3

Figure A-3: RELAP5 Nodalization Scheme of Primary Circuit of PSB-VVER Facility .............. A-4

Figure A-4: RELAP5 Nodalization Scheme of Secondary Circuit of PSB-VVER Facility ......... A-5

Figure B-1: PSB-VVER Reactor Model Measurements ........................................................... B-2

Figure B-2: PSB-VVER Loop 1 and SG-1 Model Measurement .............................................. B-3

Figure C-1: Primary Pressure .................................................................................................. C-2

Figure C-2: Primary Pressure (detail) ...................................................................................... C-2

Figure C-3: Fuel Cladding Temperature (Top of the Core) ...................................................... C-3

Figure C-4: Presurizer Level .................................................................................................... C-3

Figure C-5: HPIS + LPIS Flow (Boundary Condition) .............................................................. C-4

Figure C-6: MCP Rotor Speed (Boundary Condition) .............................................................. C-4

Figure C-7: Accumulators Levels ............................................................................................. C-5

Figure C-8: Accumulators Pressure ......................................................................................... C-6

Figure C-9: Fuel Rod Simulator Power (Boundary Condition) ................................................. C-7

Figure C-10: Core By-pass Power (Boundary Condition) ........................................................ C-7

Figure C-11: Secondary Side Pressures ................................................................................. C-8

Figure C-12: Steam Generators Levels ................................................................................... C-9

Figure C-13: Loop 1 Temperatures ........................................................................................ C-10

Figure C-14: Loop 2 Temperatures ........................................................................................ C-11

Figure C-15: Loop 3 Temperatures ........................................................................................ C-12

Figure C-16: Loop 4 Temperatures ........................................................................................ C-13

Figure C-17: Pressure Differences DP01-DP04 (DC) ............................................................ C-14

Figure C-18: Pressure Differences DP05-DP08 (Lower Plenum + Lower Part of FRS) ........ C-15

Figure C-19: Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum) ........ C-16

Figure C-20: Pressure Differences DP13-DP16 ( Upper Part of Upper Plenum) ................... C-17

Figure C-21: Break Flow ........................................................................................................ C-18

Figure C-22: Reactor Collapsed Level ................................................................................... C-18

Figure D-1: Primary Pressure ................................................................................................. D-2

Figure D-2: Primary Pressure (detail) ..................................................................................... D-2

Figure D-3: Fuel Cladding Temperature (Top of the Core) ..................................................... D-3

Figure D-4: Presurizer Level ................................................................................................... D-3

Figure D-5: HPIS + LPIS Flow (Boundary Condition) ............................................................. D-4

Figure D-6: MCP Rotor Speed (Boundary Condition) ............................................................. D-4

Figure D-7: Accumulators Levels ............................................................................................ D-5

Figure D-8: Accumulators Pressure ........................................................................................ D-6

Figure D-9: Fuel Rod Simulator Power (Boundary Condition) ................................................ D-7

Figure D-10: Core By-pass Power (Boundary Condition) ........................................................ D-7

Figure D-11: Secondary Side Pressures ................................................................................. D-8

Figure D-12: Steam Generators Levels ................................................................................... D-9

Figure D-13: Loop 1 Temperatures ........................................................................................ D-10
viii 

Figure D-14: Loop 2 Temperatures ........................................................................................ D-11

Figure D-15: Loop 3 Temperatures ........................................................................................ D-12

Figure D-16: Loop 4 Temperatures ........................................................................................ D-13

Figure D-17: Pressure Differences DP01-DP04 (DC) ............................................................ D-14

Figure D-18: Pressure Differences DP05-DP08 (Lower Plenum + Lower Part of FRS) ........ D-15

Figure D-19: Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum) ........ D-16

Figure D-20: Pressure Differences DP13-DP16 ( Upper Part of Upper Plenum) ................... D-17

Figure D-21: Break Flow ........................................................................................................ D-18

Figure D-22: Reactor Collapsed Level ................................................................................... D-18
ix 

TABLES

Page 

Table 1: Main Operational Characteristics of PSB-VVER Comparing to VVER-1000 ............. 2-3

Table 2: Test Facility Configuration in UP-11-07 Test ............................................................. 2-5

Table 3: Measured initial condition for UP-11-07 test .............................................................. 2-5

Table 4: Main Events During UP-11-07 Test ........................................................................... 2-6

Table 5: TRACE Components Statistic .................................................................................... 4-2

Table 6: List of the main systems and components of PSB-VVER TRACE input deck ........... 4-3

Table 7: RELAP5 Components Statistic .................................................................................. 4-5

Table 8: List of the main systems and components of PSB-VVER RELAP input deck............ 4-6

Table 9: Initial Conditions (TRACE and RELAP5 calculation vs. experiment comparison) ..... 5-1

Table 10: Chronology of main events (TRACE and RELAP5 calc. vs. comparison) ................ 5-6

Table 11: ACAP metrics settings .............................................................................................. 5-7

Table 12: Acceptability criterions .............................................................................................. 5-7

Table 13: Evaluation of FOM for UP-11-07 test ........................................................................ 5-9

Table 14: Run statistics ............................................................................................................. 6-1

xi 

ACKNOWLEDGMENTS 

The project FI-IM5/150 was funded by The Czech Ministry of Industry and Trade in frame of 

RD activities. 

xiii 

ABBREVIATIONS 

ACC Hydro-accumulator 

CL Cold Leg 

CS Core Simulator 

DC Downcomer 

ECCS Emergency Core Cooling System 

EREC Electrogorsk Research and Engineering Institute 

FRS Fuel Rod Simulator 

HL Hot Leg 

HPIS High Pressure Injection System 

IE Initiating Event 

ITF Integral Effect Test Facility 

LOCA Loss of Coolant Accident 

LP Lower Plenum 

LPIS Low Pressure Injection System 

MCP Main Coolant Pump 

MCP Main Coolant Pump 

NEA Nuclear Energy Agency 

NPP Nuclear Power Plant 

NRC Nuclear Regulatory Commission 

OECD Organization for Economic Co-operation and Development 

PKL Experimental Facility in Finland 

PCS Plant Control System 

PRZ Pressurizer 

PWR Pressurized Water Reactor 

RD Research and Development 

RPV Reactor Pressure Vessel 

RPV Reactor Pressure Vessel 

SCRAM Emergency Shutdown of a Nuclear Reactor 

SETF Separate Effect Test Facility 

SG Steam Generator 

TC Thermo Couple 

TH Thermal Hydraulic 

UP Upper Plenum 

VVER Russian Pressurized Water Type Reactor 

1-1 

1. INTRODUCTION

The assessment of PWR safety codes is mainly performed on the basis of experimental data 

coming from scaled-down integral or separate test facilities. The TRACE and RELAP5 code 

validation process was mainly based on the data from experimental facilities or real NPPs of 

western PWR type as well. There is a significant number of VVER type of reactors operating all 

over the world and many other are under construction or under preparation as well. VVER 

reactors are in many aspects similar to western PWRs. Therefore a lot of experimental data 

measured on PWRs or PWR based test facilities are valuable also for VVER research. On the 

other hand, the VVER design has several specific features such as larger volumes of primary 

coolant, horizontal steam generators, different ECCS injection points and so on. Therefore the 

validation of the thermo-hydraulic codes for VVER types of reactors is often required by national 

regulators. The purpose of performed analyses is to extend the validation of the TRACE and 

RELAP5 code focused on VVER type of NPPs. The best estimate thermo-hydraulic computer 

code TRACE V5.0 and RELAP5/MOD3.3. were assessed using Upper Plenum 11% break 

experiment at the large-scale test facility PSB-VVER. The PSB-VVER facility is a 1:300 volume 

scaled model of VVER 1000 NPP located in Electrogorsk, Russia. In order to perform code 

validation an extensive TRACE and RELAP5 input decks of PSB-VVER facility were developed. 

Both models include all important components of the PSB-VVER facility: reactor, four separated 

loops, pressurizer, break unit, main circulation pumps, steam generators, and important parts of 

secondary circuit. 

2-1 

2. FACILITY AND TEST DESCRIPTIONS 

Detail information about the PSB-VVER test facility systems and elements is given in Ref 4. 

Only a brief description of the PSB-VVER facility is given here. The hardware configuration for 

UP-11-07 test is reported below. 

2.1 PSB-VVER Facility 

PSB-VVER is a large-scale integral test facility which structurally corresponds to primary circuit 

of NPP with VVER-1000 (V-320 design). The volumetric and power scale is 1:300, and the main 

equipment elevations correspond to those of the prototype reactor. 

The facility consists of four loops linked up to the reactor model. Each loop has a circulation 

pump, a steam generator, hot and cold legs. One of the loops (loop No.4, "broken") has special 

branch pipes for connection to primary leakage simulation system. The test facility also includes 

a pressurizer (PRZ) and ECCS, which has, as in actual VVER-1000, three subsystems: a 

passive system and two active ones. 

The reactor model comprises four elements: an external downcomer, core model, core bypass 

and an upper plenum. The PSB-VVER core model consists of 168 full-height indirectly 

electrically heated fuel rod simulators with uniform power distribution. The rod simulator pitch 

(12.75 mm) and diameter (9.1 mm) are identical to those of the reference reactor. The fuel rod 

simulators are arranged on a triangular grid. The rod bundle cross section has the shape of 

regular hexagon with "wrench" size of 168 mm. The core model represents the central part of 

the reference fuel rod assembly. The PSB rod simulator bundle has 15 spacer grids with 

prototypic geometry. 

PSB-VVER pressurization system includes a pressurizer, surge lines, spray lines, and a relief 

valve. By means of surge and spray lines the pressurizer can be connected to the “broken” loop 

(loop #4) or to one of the intact loops (loop #2) of the facility. The PRZ vessel height, the bottom 

elevation and location of nominal level correspond to the reference ones. An electric heater with 

a power of up to 80 kW is built in the lower part of the pressurizer vessel. 

PSB main circulation pumps are used to provide forced circulation in primary circuit. The 

circulation pumps are variable-speed ones of vertical centrifugal single-stage type and can 

operate under two-phase fluid conditions. 

The passive ECCS system consists of four accumulators connected in pairs to an inlet and 

outlet chamber of the reactor pressure vessel. The active ECCS system consists of high 

pressure injection system (HPIS) and low pressure injection system (LPIS). Cooling water of 

active ECCS can be supplied to three loops, both to cold and hot legs as original facility design. 

The PSB-VVER SG is a vertical vessel with two vertical headers inside. A bundle of horizontal 

spiral heat-exchanging tubes of full size is mounted between the two headers. The PSB-VVER 

SG is designed in such way that the reference tube bundle elevations and tube lengths to be 

conserved, as well as the flow area. Heat transfer surface and secondary fluid volume to be 

matched the scale factor. On the secondary side, the feed water system and the main steam 
2-2 

lines are simulated. The turbine and the condenser are not modeled. 

Primary and secondary circuits of the PSB-VVER facility are operated at nominal pressure of a 

reactor prototype. 

Figure 1 depicts an isometric projection of the test facility. Main operational characteristics of 

the test facility are given in Table 1. 

Figure 1: General View of PSB-VVER Facility 
2-3 

Table 1: Main Operational Characteristics of PSB-VVER Comparing to VVER-1000 

Parameter Units VVER-1000 PSB-VVER 

Coolant - water water 

Number of circulation loops - 4 4 

Primary circuit 

Pressure MPa 15.7 15.7 

Coolant temperature (hot/cold leg) deg 290/320 290/320 

Coolant flowrate m3/h 82485 < 280 

Core power MW 3000 15 

Secondary circuit 

Steam generator pressure MPa 6.3 6.3 

Feed water temperature deg 220 < 270 

Thermal power of one SG MW 750 2.5 

The PSB test facility is equipped with special break systems to facilitate research of thermal 

hydraulics during break accidents. There is a special system to simulate accumulator water 

supplying pipe rupture which is utilized for 11% upper plenum break experiments (UP-11-07 

and PSBV1). The break system consists of a break unit, a discharge pipeline with isolating 

valves and catch tank-condenser. Principal scheme of the break unit is given in Figure 2 and 

geometric characteristics of the discharge line are represented in Figure 3. 

Figure 2: Upper Plenum Break Unit 
2-4 

Figure 3: Discharge line 
2-5 

2.2 Experiment UP-11-07 

The test UP-11-07 „Upper Plenum Break 11%“ was performed in the PSB-VVER test facility at 

Electrogorsk Research and Engineering Center (EREC) in Russia. The thermal-hydraulic 

processes related to upper plenum break 11% were researched. 

2.2.1 Facility configuration 

The information on the test facility hardware and configuration of the system specific for UP-11-

07 test are given in the Table 2. 

Table 2: Test Facility Configuration in UP-11-07 Test 

Equipment Status 

Pressurizer Connected to the loop #2 

Core by-pass 2 throttles with 2 orifices of diameter 7 mm were installed at inlet 

and outlet of core by-pass 

HPIS One channel was connected to the cold leg of the loop #1 

LPIS Two channels connected to the cold and hot legs of the loop #1 

ACCs ACCs #1 and 3 were connected to UP. ACC #4 was connected 

to downcomer. ACC #2 was switched off (isolated) 

SGs All SGs were connected through steam lines 

Feed water heater In use. SG levels under steady-state were maintained by supply 

of feed water 

Large break unit Break is located in upper plenum. 

Leak channel is a throttle of 16 mm, L = 160 mm. 

Horizontal blowdown line of 45 mm is located below hot legs 

inlet by 200 mm. Coolant discharging is realized from annular 

chamber between annular screen and UP wall 

2.2.2 Initial Conditions 

The main initial conditions of UP-11-07 test are given in the Table 3. UP-11-07 test has been 

performed under reduced initial core power corresponding to approximately 15% of nominal 

power. 

Table 3: Measured initial condition for UP-11-07 test 

Parameter Units Value 

Primary circuit 

Pressure in upper plenum (gauge YC01P16) MPa 15.744 

Coolant temperature 

(DC inlet/UP outlet - gauges C01T02/YC01T04) deg 274.6 / 306.6 

Primary loops flow rates (gauges YA01÷04F01) kg/s 2.321 / 2.289 / 2.343 / 2.357 

Core power (gauge YC01N01) kW 1496.5 

Core by-pass power (gauge YC01N02) kW 17.1 

Coolant level in PRZ (gauge YP01L02) m 6.472 

Secondary circuit 

Pressure in SGs (gauges RA01÷04P01) MPa 6.269 / 6.300 / 6.192 / 6.285 

Level in SGs (gauges YB0104L01) m 1.694 / 1.692 / 1.835 / 1.664 
2-6 

ECCS 

Pressure in ACCs (gauges TH0104P01) MPa 6.015 / 6.019 / 5.890 / 5.898 

Level in ACCs (gauges TH0104L01) m 4.800 / 5.172 / 4.807 / 4.806 

2.2.3 Boundary Conditions (test scenario) 

Detail information about the UP-11-07 test boundary conditions is given in Ref 5. The main 

events of UP-11-07 test are described in the Table 4. 

Table 4: Main Events During UP-11-07 Test 

Event Time [s] 

IE – Break Opening 0 

Pressure in UP (YC01P16) < 13.73 MPa  conditions for SCRAM 2 

- Start of MCP coastdown 2 

- Start of core power and core bypass power reduction 3 

- Stop of feed water flow supply 9 

- Stop of steam removal from steam generators 11 

HPIS activation 12.0 

ACCs activation 110 / / 112 / 109 

LPIS activation 383 

ACCs empty 504 / - / 499 / 489 

LPIS termination 3645 

HPIS termination 3696 

End of test (FRS power switched off) 5593 

The experiment is started with opening of an isolation valve XL10S01 in the leak line. 

When the UP pressure PUP = 13.73 МPа PCS automatically cuts the electric power from PRZ 

heaters and starts to simulate operation of NPP automatics in accordance with the SCRAM 

signal (station blackout takes place simultaneously), which provides for the following actions 

to be performed at the test facility: 

Power reduction on core simulator and core by-pass is started 

 the procedure of MCPs coastdown is started 

 stop steam removal from steam generators 

 close feed water supply 

There was rather nonstandard coastdown of MCPs realized. MCPs rotational speed was 

temporarily stabilized at revolution corresponding to 29.5% of nominal value for 200 second, 

and then the standard coastdown continued. 

After achievement of value of 150 kW, the core model power is fixed and is the same up to 

the end of the experiment. At this moment the core by-pass power is also fixed and is the 

same up to the end of the experiment. 

After achievement of two conditions 1) UP pressure decreases to the value PUP = 10.88 

МPа and 2) 10 s later the moment when PUP = 13.73 МPа (interval of time to start operation 

of diesel-generator), start of high pressure injection system operation is simulated. Cooling 

water from HPIS is supplied to the cold leg of loop #1. HPIS mass flow rate 0.10 kg/s is 

provided. Temperature of injected cooling water is 28°C in the beginning of test. 
2-7 

The accumulators start operate when pressure in upper plenum decreases below 5.89 MPa 

(accumulator #2 does not operate in this test). Operation of accumulators is stopped when 

level in accumulators falls down to 0.100 m. 

 

When UP pressure decreases to the value PUP = 1.76 МPа, regular starting of low pressure 

ECCS pump is simulated. Cooling water of LPIS is supplied to cold and hot legs of loop #1, 

mass flow rate is 0.205 kg/s per each line. HPIS and LPIS keep water delivery to primary 

circuit till total water supply of HPIS and LPIS achieves a value 1.72 m3

 (simulation of 

emptying of LPIS water tank), when operation of HPIS and LPIS is terminated. 

The experiment is stopped when cladding temperature is reaching a value of 1000°C.. 

. 

. 

3-1 

3. THE TRACE AND RELAP5/MOD3.3 CODES 

The TRACE (TRAC/RELAP Advanced Computational Engine) is the latest in a series of 

advanced, best-estimate reactor systems codes developed by the U.S. Nuclear Regulatory 

Commission for analyzing transient and steady-state neutronic-thermal-hydraulic behavior in 

light water reactor. The RELAP5 computer code is one of four TRACE’s predecessor. 

Both codes have been widely used by U.S. Nuclear Regulatory Commission (NRC) and other 

organizations for rulemaking, licensing audit calculations, evaluation of operator guidelines, and 

as a basis for a nuclear plant analyzer. Specific applications of their capability have included 

simulations of transients in LWR systems, such as loss of coolant, anticipated transients without 

scram (ATWS), and operational transients such as loss of feedwater, loss of offsite power, 

station blackout, and turbine trip. The TRACE and RELAP5 are a highly generic code that, in 

addition to calculating the behavior of a reactor coolant system during a transient, can be used 

for simulation of a wide variety of hydraulic and thermal transients in both nuclear and 

nonnuclear systems involving mixtures of steam, water, noncondensable gases, and solute. 

 

The TRACE code is a component-oriented reactor systems analysis code designed to analyze 

light water reactor transients up to the point of significant fuel damage. The TRACE code solves 

a finite-volume two-phase multidimensional compressible flow with one, two and three 

dimensional flow geometry. The TRACE code can model heat structures and control systems 

that interact with component models and the fluid solution. The TRACE code has capability to 

use build-in point reactor kinetics or 3D reactor kinetics through coupling with Purdue Advanced 

Reactor Core Simulator (PARCS). In addition the TRACE code can be coupled with another 

TRACE jobs or other codes (CFD, CONTAIN ...) through its exterior communications interface 

(ECI). TRACE uses what is commonly known as a 6-equation model for two-phase flow (mass 

equation, equation of motion and energy equation for each phase). Additional equations can be 

solved for noncondensable gas, dissolved boron, control systems and reactor power. There are 

five additional closure relationships for field equations: equations of state, wall drag, interfacial 

drag, wall heat transfer and interfacial heat transfer. These constitutive models are semi 

empirical equations. There are two numerical methods available in TRACE: semi-implicit 

method and the stability enhancing two-step (SETS) method. 

RELAP5/MOD3.3 uses a one-dimensional, two fluids, nonequilibrium, six equation 

hydrodynamic model with a simplified capability to treat multi-dimensional flows. This model 

provides continuity, momentum, and energy equations for both the liquid and the vapor phases 

within a control volume. The energy equation contains source terms which couple the 

hydrodynamic model to the heat structure conduction model by a convective heat transfer 

formulation. The code contains special process models for critical flow, abrupt area changes, 

branching, crossflow junctions, pumps, accumulators, valves, core neutronics, and control 

systems. A countercurrent flow limitation model can also be applied at vertical junctions 
3-2 

3.1 The TRACE and RELAP5/MOD3.3 code assessment 

Confidence in the computational tools (codes) and establishment of their validity for a given 

application depends on proper assessment. TRACE and RELAP5, like other two-fluid codes, is 

composed of numerous models and correlations. When applied to full scale nuclear power plant 

conditions, many of these models and correlations can be applied outside of their original 

scope. By assessing the code against thermal-hydraulic tests, it is possible to show that the 

code and its constituent model packages can be extended to conditions beyond those for which 

many of the individual correlations were originally intended (Ref 16). The assessment process 

however, can also indicate potential deficiencies in the code. There are following four sources of 

data for code assessment (Ref 17): 

 “Fundamental” experiments 

 Separate effect test facilities (SETF) 

 Integral test facilities (ITF) 

 Real plant data 

3.2 VVER typical features related to TRACE and RELAP5/MOD3.3 code assessment 

The TRACE and RELAP5 code validation process is mainly based on the data from 

experimental facilities or real NPPs of Western PWR type. VVER reactors are in many aspects 

similar to Western PWRs. Therefore a lot of experimental data measured on PWRs or PWR test 

facilities is valuable also for VVER research. On the other hand, the VVER design has several 

specific features. From the hardware point of view the main differences between VVER-1000 

and PWR are the following (Ref 19): 

 Horizontal steam generators with 2 headers 

 Lower plenum internal structures 

 Fuel assemblies with hexagonal fuel rod arrangements 

 ECCS injection points 

 Secondary side water volume of the steam generators is larger 

 Operational conditions and set points of actuation of ECCS 

 Working conditions of secondary side of steam generators and set points for the 

operation of feedwater and steam line 

There are approximately 50 operating units of VVER type (Ref 18). It is a meaningful number in 

comparison to approximately 216 operating units of PWR reactors (Ref 18). Therefore 

corresponding attention should be given to code validation for VVER type of reactors. 
4-1 

4. INPUT DECK DESCRIPTION 

An extensive TRACE and RELAP5 input deck of PSB-VVER facility was developed including all 

important components of the PSB-VVER facility: reactor, 4 separated loops, pressurizer, break 

units, main circulation pumps, steam generators, break sections and important parts of 

secondary circuit. Both input decks were designed on the basis of PSB-VVER facility 

documentation (Ref 4, 20). 

4.1 The TRACE input deck 

Nodalization diagrams of the TRACE Input Deck are presented in Figure A-1 (reactor + primary 

circuit) and Figure A-2 (secondary circuit) in the Appendix A. The TRACE model of the reactor 

pressure vessel (RPV) with internal structures is divided into 3 parts. The first part represents 

Downcomer (DC) + Lower Plenum (LP), the second part represent Fuel Rod Simulator (FRS) + 

Upper Plenum (UP) and the third part represents core by-pass from DC to UP. The RPV model 

employs VESSEL component for DC + LP (includes 26 axial layers, 1 azimuthal theta sector 

and 2 radial rings), 3 PIPE components for core by-pass and the next VESSEL component for 

FRS + UP (includes 32 axial layers, 4 azimuthal theta sector and 6 radial rings). The model also 

includes by-pass piping between DC to UP and UP heating pipe between cold leg of loop #1. 

Each of the four coolant loops comprises of: a hot leg, steam generator, pump suction loop seal 

piping, main coolant pump, and a cold leg including control valve between MCP discharge and 

DC. 

The pump performance is based on single-phase head and torque characteristic of the pump 

TsNIS 1620 from Ref. 20. No two-phase degradation was modeled because of no appropriate 

data. 

The pressurizer is modeled using component PIPE equipped with heaters and with surge lines 

connected to the hot legs of loop #2 and loop #4. Pressurizer can be optionally connected to 

loop #2 or loop #4 (as original facility design). 

Active ECCS are modeled using simple boundary condition - FILL component, accumulators are 

modeled using PIPE and VALVE components. High pressure injection system (HPIS) optionally 

provides flow to the hot leg of loop #1 and #4. Low pressure injection system (LPIS) optionally 

provides flow to the hot and cold leg of loop #1 and #3 (test UP-11-07) and loop#3. Four 

accumulators provide flow to the downcomer and upper plenum (two to each location), and any 

of them can be switch off. Cooling water delivery from ECCS depends on hardware 

configuration of a particular test. 

Steam generator is modeled using multi-tube approach. The primary side of SG Input Deck 

consists of 5 axial layers of heat exchanging tubes and two headers (the original facility SGs 

consists of 34 tubes). Each axial layer is divided into 15 segments. The SG secondary part is 

modeled as original three-channel complex with 10 axial layers (5 of them in the area of 

exchanging tubes). The feedwater system, the steam lines connected to all SGs and the 
4-2 

common steam header are also modeled. The BREAK component simulates the release of 

secondary steam from steam header. 

The heat loss from the primary and secondary is represented in the TRACE model by entering 

the thickness of the insulation on the outside of all the pipes and other system components. 

Appropriate material properties are input for the insulation. A constant boundary temperature 

and heat transfer coefficient of outer air is applied. 

The model contains 1756 volumes, 3433 junctions, and 1285 heat structures with 4814 mesh 

points. Standard modeling guidelines were followed in developing the nodalization of the 

system. 

Components Statistic for TRACE model – see the next Table 5. 

Table 5: TRACE Components Statistic 

TRACE Component Notes 

VESSEL 2 DC+LP; FRS+UP 

PIPE 67+66*1 - 

HSTR 157+1*1 - 

POWER 1+1*1 FRS simulator + By-pass heating 

VALVE 25 - 

PUMP 4 MCPs 

BREAK 4 Upper plenum and Large Break unit + release 

of secondary steam 

FILL 8 HPIS, LPIS, Feedwater 

Whole No of Components 336 - 

*1 the second number is the number of spawned component 
4-3 

Table 6: List of the main systems and components of PSB-VVER TRACE input deck 

PSB system Input deck Used TRACE components 

Rector (YC) 

Downcomer + VESSEL + HTSTR 

Lower plenum + VESSEL + PIPE + HTSTR 

Core + VESSEL + HTSTR + POWER + CONTROL BLOCK 

Upper plenum + VESSEL + HTSTR 

Core bypass + PIPE + HTSTR + POWER + CONTROL BLOCK 

DC to UP bypass + VALVE + HTSTR 

UP heating + VALVE + HTSTR 

LOOP (YA) 

Hot leg + PIPE + HTSTRT 

Loop seal + PIPE + HTSTRT 

Cold leg + VALVE + HTSTRT 

Main cooling pump (YD) PUMP + HTSTR + CONTROL BLOCK 

Pressurizer (YP) 

Vessel + PIPE + HTSTR 

heaters + HTSTR + CONTROL BLOCK 

Surge line + VALVE + HTSTR 

Relief valve - 

ECCS (TJ, TH) 

HPIS (+) FILL + CONTROL BLOCK 

LPIS (+) FILL + CONTROL BLOCK 

Accumulators + PIPE + VALVES + HTSTRT 

Steam generators (YB) 

Vessel + PIPE + HTSTR 

Heat exchange tubes + PIPE + HTSTR 

Primary headers + PIPE + HTSTR 

Steam lines + VALVE + HTSTR 

Feedwater (+) FILL + CONTROL BLOCK 

Relief valve - 

Steam headers (RA) + VALVE + HTSTR + BREAK 

Key: + a fine model 

 (+) a simplified model 

 - not modeled 
4-4 

4.2 The RELAP5 input deck 

Nodalization diagrams of the RELAP5 Input Deck are presented in Figure A-3 (reactor + primary 

circuit) and Figure A-4 (secondary circuit) in the Appendix A. The RELAP5 model of the reactor 

pressure vessel (RPV) with internal structures is divided into 3 parts. The first part represents 

Downcomer (DC) + Lower Plenum (LP), the second part represent Fuel Rod Simulator (FRS) + 

Upper Plenum (UP) and the third part represents core by-pass from DC to UP. All parts of RPV 

employ PIPE, BRANCH and SNGLJUN components. The model also includes by-pass piping 

between DC to UP. 

Each of the four coolant loops comprises of: a hot leg, steam generator, pump suction loop seal 

piping, main coolant pump, and a cold leg including control valve between MCP discharge and 

DC. 

The pump performance is based on single-phase head and torque characteristic of the pump 

TsNIS 1620 from Ref. 20. No two-phase degradation was modeled because no appropriate data 

were available. 

The pressurizer is modeled using component PIPE equipped with heaters and with surge lines 

connected to the hot legs of loop #2 and loop #4. Pressurizer can be optionally connected to 

loop #2 or loop #4 (as original facility design). Relief valves are included on the pressurizer. 

Active ECCS are modeled using simple boundary condition - TMDPJUN + TMDPVOL 

components, accumulators are modeled using ACCUM, PIPE and VALVE components. High 

pressure injection system (HPIS) optionally provides flow to the hot leg of loop #1 and #4. Low 

pressure injection system (LPIS) optionally provides flow to the hot and cold leg of loop #1 and 

#3 (test UP-11-07) and loop #3. Four accumulators provide flow to the downcomer and upper 

plenum (two to each location), and any of them can be switch off. Cooling water delivery from 

ECCS depends on hardware configuration of a particular test 

The steam generator secondary side is modeled with a single stack of volumes in the tube 

bundle region. There is no physical barrier between the tube bundle and the outer shell of the 

steam generator. The secondary side of the steam generators included main and auxiliary 

feedwater, individual steam lines and the common steam header, and all of the steam lines is 

equipped with relief valve. 

The heat loss from the primary and secondary is represented in the RELAP5/MOD3.3 model by 

entering the thickness of the insulation on the outside of all the pipes and other system 

components. Appropriate material properties were input for the insulation. A constant boundary 

temperature and heat transfer coefficient of outer air is applied. 

Choking was turned off at most of the junctions, as recommended in the code user guidelines. 

Exceptions were the break, at valves, at the pressurizer connection to the surge line and the 

surge line connections to the hot leg, and at the core outlet. The break piping was attached to 

the upper plenum one volume below the hot leg connections. The break was modeled as an 

abrupt area change with user-input loss coefficients. 

The RELAP5/MOD3.3 countercurrent flow limitation (CCFL) model was applied at five locations 

in the model: in the downcomer below the accumulator injection nozzles, at the core outlet, in 
4-5 

the upper plenum below the accumulator injection nozzles, at the plate between the upper 

plenum and the upper head, and at the outlet of the riser section on the secondary side of the 

steam generators. These junctions were selected because they were vertically oriented, 

represented places where the flow area changed, and modeled regions of the facility where 

CCFL might be expected to occur. 

The model contains 536 volumes, 559 junctions, and 521 heat structures with 5549 mesh 

points. Standard modeling guidelines were followed in developing the nodalization of the 

system. 

Table 7: RELAP5 Components Statistic 

TRACE Component Notes 

PIPE 61 - 

BRANCH 37 - 

SNGLVOL 4 - 

ANNULUS 2 - 

PUMP 4 MCPs 

ACCUM 4 Accumulators 

VALVE 23 - 

SNGLJUN 6 - 

MTPLJUN 8 - 

TMDPVOL 

19 

Upper plenum and Large Break unit, release 

of secondary steam, HPIS, LPIS, relief valves 

BC 

TMDPJUN 10 HPIS, LPIS, Feedwater 

HSTR 132 - 

Whole No of Components 310 - 
4-6 

Table 8: List of the main systems and components of PSB-VVER RELAP input deck 

PSB system Input deck Used RELAP components 

Rector (YC) 

Downcomer + PIPE + BRANCH + HTSTR 

Lower plenum + PIPE + BRANCH + SNGLJUN + HTSTR 

Core + PIPE + HTSTR + CONTRL BLOCK 

Upper plenum + PIPE + BRANCH + SNGLJUN + HTSTR 

Core bypass + PIPE + SNGLJUN 

DC to UP bypass + PIPE + VALVE + BRANCH + HTSTR 

UP heating - 

LOOP (YA) 

Hot leg + PIPE + HTSTR 

Loop seal + PIPE + HTSTR 

Cold leg + PIPE + SNGLVOL + HTSTR 

Main cooling pump (YD) PUMP + HTSTR + CONTRL BLOCK 

Presurizer (YP) 

Vessel + PIPE + HTSTR 

heaters + HTSTR + CONTROL BLOCK 

Surge line + PIPE + BRANCH + VALVE + HTSTR 

Relief valve (+) VALVE + TMDPVOL 

ECCS (TJ, TH) 

HPIS (+) TMDPJUN + TMDPVOL + CONTROL BLOCK 

LPIS (+) TMDPJUN + TMDPVOL + CONTROL BLOCK 

Accumulators + ACCUMULATOR + PIPE + HTSTR 

Steam generators (YB) 

Vessel + PIPE/ANNULUS + HTSTRT 

Heat exchange tubes + PIPE + HTSTR 

Primary headers + PIPE + BRANCH + MTPLJUN + HTSTRT 

Steam lines + BRANCH + VALVE 

Feedwater (+) TMDPJUN + TMDPVOL + CONTROL BLOCK 

Relief valve TMDPJUN + TMDPVOL 

Steam headers (RA) + PIPE + SNGLJUN + TMDPVOL 

Key: + a fine model 

 (+) a simplified model 

 - not modeled 
5-1 

5. RESULTS 

5.1 TRACE Steady-state calculation 

In order to achieve stable initial conditions of the UP-11-07 test, the steady state was calculated 

for 300 s. The following controllers were used for the first 300 s: 

 Pressurizer pressure controller 

 Steam generators level controllers 

The other controlled parameters (fuel rod simulator power, core bypass power, feedwater 

temperature, main steam header pressure) were entered as boundary conditions. The steam 

generator pressures is lower than the measured values, because the steam header pressure 

was adjusted to get the desired reactor vessel inlet temperature (average cold legs 

temperature) in steady state calculation. A real PSB controller that kept the liquid level in all 

SGs within a desired band was replaced by a PI-controller for 200 s of steady-state calculation, 

for the rest of steady state calculation and for transient calculation was feed water flow to SGs 

entered as a boundary condition. Steady state calculation took approximately 30 min. Main 

calculated and measured parameters are compared in the Table 9. 

5.2 RELAP5 Steady-state calculation 

In order to achieve stable initial conditions of the UP-11-07 test, the steady state was calculated 

for 1600 s. The following controllers were used for the first 1600 s: 

• Pressurizer pressure controller 

• Pressurizer level controller 

• Main steam header pressure controller 

• Steam generators level controllers 

• Main circulation pumps velocity controllers 

The same boundary conditions like in the TRACE calculation were used in the RELAP5 

calculation. Because of longer steady state calculation, feedwater steady state PI-controller was 

in operation for 1500 s and for the rest of steady state calculation and for transient calculation 

SG feedwater flow was entered as a boundary condition in the same way like in TRACE 

calculation. Steady state calculation took approximately 6 min. Main calculated and measured 

parameters are compared in the Table 9 . 

Table 9: Initial Conditions (TRACE and RELAP5 calculation vs. experiment comparison) 

Parameters Units accuracy UP-11-07*2 RELAP

5 TRACE

Upper plenum pressure (YC01P16) MPa ± 0.16 15.774 ± 0.021 15.765 15.718 

Pressure drop at FRS (YC01DP07-DP10) kPa ± 4.60*1 -28.63 -28.94 -28.32 

Hot Leg outlet Coolant Temp. (TA01T03) °C ± 3 305.4 ± 0.3 305.7 305.8 

Hot Leg outlet Coolant Temp. (TA02T03) °C ± 3 305.2 ± 0.3 306.7 305.8 

Hot Leg outlet Coolant Temp. (TA03T03) °C ± 3 304.3 ± 0.4 306.6 305.8 
5-2 

Parameters Units accuracy UP-11-07*2 RELAP

5 TRACE

Hot Leg outlet Coolant Temp. (TA04T03) °C ± 3 304.7 ± 0.3 306.7 305.8 

Cold Leg outlet Coolant Temp. (YA01T02) °C ± 3 276.0 ± 0.3 276.3 275.2 

Cold Leg outlet Coolant Temp. (YA02T02) °C ± 3 275.7 ± 0.3 276.2 275.1 

Cold Leg outlet Coolant Temp. (YA03T02) °C ± 3 274.7 ± 0.3 276.1 275.3 

Cold Leg outlet Coolant Temp. (YA04T02) °C ± 3 274.5 ± 0.3 276.2 275.2 

Loop-1 flow rate (YA01F01) kg/s ± 0,05 2.321 ± 0.017 2.299 2.318 

Loop-2 flow rate (YA02F01) kg/s ± 0,05 2.289 ± 0.010 2.316 2.286 

Loop-3 flow rate (YA03F01) kg/s ± 0,05 2.343 ± 0.020 2.322 2.340 

Loop-4 flow rate (YA04F01) kg/s ± 0,05 2.357 ± 0.018 2.323 2.354 

FRS power (YC01N01) kW ± 15 1496.5 ± 11.3 1500.0 1500.0 

Core by-pass power (YC01N02) kW ± 0,4 17.1 ± 0.2 17.0 17.0 

Collapsed level in PRZ (YP01L02) m ± 0.3 6.472 ± 0.051 6.431 6.473 

 

Pressure in SG-1 (YB01P01) MPa ± 0.05 6.269 ± 0.04 6.136 6.057 

Pressure in SG-2 (YB02P01) MPa ± 0.05 6.300 ± 0.035 6.149 6.073 

Pressure in SG-3 (YB03P01) MPa ± 0.05 6.192 ± 0.066 6.135 6.071 

Pressure in SG-4 (YB04P01) MPa ± 0.05 6.285 ± 0.031 6.134 6.054 

Collapsed level in SG-1 (YB01L01) m ± 0.07 1.694 ± 0.047 1.624 1.689 

Collapsed level in SG-2 (YB02L01) m ± 0.07 1.692 ± 0.037 1.709 1.714 

Collapsed level in SG-3 (YB03L01) m ± 0.07 1.835 ± 0.124 1.864 1.871 

Collapsed level in SG-4 (YB04L01) m ± 0.07 1.664 ± 0.023 1.620 1.683 

 

ACCU-1 pressure (TH01P01) MPa ± 0.03 4.800 ± 0.002 4.800 4.800 

ACCU-2 pressure (TH02P01) switched off MPa ± 0.03 5.172 ± 0.002 5.172 5.172 

ACCU-3 pressure (TH03P01) MPa ± 0.03 4.807 ± 0.001 4.807 4.807 

ACCU-4 pressure (TH04P01) MPa ± 0.03 4.806 ± 0.001 4.806 4.806 

ACCU-1 collapsed level (TH01P01) m ± 0.07 6.015 ± 0.004 6.016 6.015 

ACCU-2 collapsed level (TH02P01) m ± 0.07 6.019 ± 0.005 6.020 6.019 

ACCU-3 collapsed level (TH03P01) m ± 0.07 5.890 ± 0.005 5.889 5.890 

ACCU-4 collapsed level (TH04P01) m ± 0.07 5.898 ± 0.004 5.898 5.898 

*1 sum of accuracy of pressure drop YC01DP07-DP10 (accuracy of YC01DP07,08,10 = ± 1.2 

kPa, YC01DP09 = ± 1.0 kPa) 

*2 - Average value ±standard deviation of measured parameters at initial steady state condition 

of the test facility 

5.3 TRACE transient calculation 

The post-test calculation of UP-11-07 experimental test at PSB-VVER facility started at the time 
5-3 

0 s with an initiating event – upper plenum break 11 % (to simulate accumulator water supplying 

pipe rupture). A comparison of calculated and experimental times of the occurrence of main 

events is presented in the Table 10. Time courses of all important parameters and their 

comparison with experimental data are presented in Appendix C. 

A brief overview of the behavior observed in the experiment is provided here, then the 

comparisons between the measured data and calculations will be presented and discussed. 

Following the break opening, the system pressure decreased rapidly. Measured and calculated 

pressures in the upper plenum are presented in Figure C-1. At the time 2 s upper plenum 

pressure decreased bellow 13.73 MPa and SCRAM signal was triggered simulated. The 

SCRAM initiated core and by-pass power reduction along to the specified function, that 

represents simulation of core decay heat after the SCRAM and MCP coastdown, with temporary 

stabilization at 24 % of nominal speed. The depressurization slowed near 40 s, as liquid began 

to boil in the entire core. As the pressure continued to decrease, ECCS injection begun: HPIS 

injection at 18 s was followed by accumulators injection approximately at 108 s and finally by 

LPIS injection at 383 s. As the system pressure decreased the total ECC injection could 

compensate decreasing break flow. Collapsed level in vessel achieved minimal value 

approximately at +600 s, see pressure differences in Figure C-18 and C-19, collapsed level in 

vessel slowly increased after this time until ECC injection was terminated at +3697 s. After ECC 

injection was terminated drainage of the reactor vessel begun again and top portion of the core 

started heat up at 5323 s, see Figure C-3. When cladding temperature in the top portion of the 

core exceeded 1000°C, electric power supply to core was switched off and experiment was 

terminated. 

During first seconds of calculation calculated primary system depressurization rate was higher than 

measured one despite the fact that calculated break flow was slightly higher than measured one, 

see Figure C-2 and C-21. But overall agreement between both parameters (primary system 

pressure and break flow) during the whole transient course was very good (according to Ref. 4 - 

error of measured break flow can be up to 20 – 30 %). 

After break was opened pressurizer heaters automatically tried to prevent the system pressure from 

decreasing and increased heating to maximum value. But at +1.0 s the system pressure decreased 

below heaters setpoint 13.73 MPa and heaters were switched off. 

At 1.0 s the system pressure decreased below 13.73 MPa, thus SCRAM signal was simulated and 

MCP coast down begun. SCRAM signal was followed by the core and by-pass power reduction at 

+2.5 s. Time courses of power reduction of core and by-pass and MCP coast down were specified 

as boundary conditions based on experiment data see Figure C-9, C-10, and C-6. 

In the beginning of the experiment the high break flow was not compensated and the primary 

system was emptying. Calculated emptying rate of pressurizer was nearly the same like measured. 

The part of pressurizer above heaters (gauges YP01L02) was emptied at +15 s in calculation 

whereas in experiment at 16 s (see Figure C-4). 

The pressure in steam generators is presented in Figure C-11. The measured and calculated 

pressure increased rapidly to a peak 6.86 MPa in experiment and 6.81 in calculation when isolation 

valve at common part steam header RA06S01 was closed. The secondary system pressure did not 

achieve a value of BRU-A activation neither in experiment nor in calculation. Calculated absolute 

value of the secondary pressure peak agree with experiment very well. Calculated pressure change 
5-4 

from steady state level pressure to pressure peak was 20% higher, because of lower calculated 

steady state level pressure before pressure peak (approximately 0.15 MPa lower than measured 

one). It was probably due to the used heat structure correlation that is not capable to perfectly 

model the coil tube bundle in the steam generator. During the rest of calculation secondary 

pressure decreased more quickly than in experiment. It was probably caused by overshooting of 

heat losses of SG vessels and steam lines (primary loops were emptied and there were no heat 

exchange between primary and secondary side of SG). 

When the primary system pressure dropped below the secondary system pressure (between 80 

and 90 s in the experiment and approximately at 60 s in the calculation), steam generator did not 

remove primary heat any more. As the primary pressure continued to decrease below 10.88 MPa 

and HPI injection begun at 18 s in the experiment and in 16 s in calculation, mass flow 0.105 kg/s 

was injected to the hot leg loop #1, see Figure C-5. 

Primary system pressure was continuously decreasing, so the primary system pressure dropped 

below accumulators pressure at +108 s in the experiment and at +121 s in the calculation.. 

Condensation of some of the steam in the system by the cold ECC liquid caused the 

depressurization rate to increase. When levels in accumulators fell down to 0.1 m they were cut off 

at +489-500 s in experiment and at +579-632 s in calculation. The accumulators injection was 

followed by LPI injection at + 383 s in experiment and at + 347 s in calculation, so the mass flow 0.2 

kg/s was provided to the hot and cold leg of loop #1. As the system pressure decreased, the total 

ECC injection could compensate decreasing break flow. Collapsed level in the reactor vessel 

achieved minimal value approximately 600 s (see Figure C-22) after this time slowly increased until 

ECC injection was terminated at +3697 s in experiment and at +3730 s in calculation. HPI and LPI 

injection were terminated when total water delivery achieved 1.72 m3

 (simulation of emptying ECC 

tanks). After ECC injection was terminated drainage of the reactor vessel begun again, see 

pressure differences in core portion of vessel (YC01DP07 – DP10) in Figure C-18 and C-19. 

Deviation of calculated cladding temperature from experiment data was up to 5 °C approximately 

until +2600 s – see Figure C-3. During following time course deviation increased up to 18 °C 

(accuracy of measured cladding temperature channel YC01T10 was ±10.8°C ). Increasing 

deviation of calculated cladding temperature from the experiment was caused by lower calculated 

average void fraction in the core region of vessel – it means that TRACE calculated less liquid in 

core than was measured in experiment, see pressure differences in core portion of vessel 

(YC01DP07 – DP10) Figure C-18 and C-19. Possible reason of undershooting of core liquid 

inventory are that CCFL at the core exit is preventing liquid from falling back into the core region in 

the calculation 

In the calculation, the heat-up of top portion of the core started at +4950 s, it was about 

approximately 350 s earlier than in test, see Figure C-3. Calculated earlier start of heat-up was 

caused by less calculated water inventory in the core than in the test - possible reasons are 

mentioned in a previous paragraph. Calculated heat-up rate was higher than those in the test, see 

Figure C-3. Similar behavior was observed in other post-test calculations of an upper plenum small 

break performed in the code RELAP5/MOD3.2. (Ref. 7, 8) and also in the code CATHARE (Ref. 

11). 

At +5593 s power supply to core and by-pass was switched off and experiment was terminated. 
5-5 

5.4 RELAP5 transient calculation 

The post-test calculation of UP-11-07 experimental test at PSB-VVER facility started at the time 

0 s with initiating event – upper plenum break 11 % (to simulate accumulator water supplying 

pipe rupture). A comparison of calculated and experimental times of the occurrence of main 

events is presented in the Table 10. Time courses of all important parameters and their 

comparison with experimental data are presented in Appendix D. 

The comparisons between the measured data and calculations will be presented and discussed 

here. 

Calculation started at 0s by opening the valve on leakage line from upper plenum. 

The calculated break flow was little more higher than was measured immediately after the break 

was opened so the system pressure decreased a little more rapidly, see Figure D-2 and D-21. 

But overall agreement both parameters during the whole transient course with measured data 

were very good (Ref. 5 - error of measured break flow can be up to 20 – 30 %). 

After the break was opened pressurizer heaters automatically tried to prevent the system 

pressure from decreasing and increased heating to maximum value. But at +3.5 s the system 

pressure decreased below heaters setpoint 13.73 MPa so heaters were switched off. 

At 3.5 s the system pressure decreased bellow 13.73 MPa and SCRAM signal was simulated. 

SCRAM signal was followed by the core and by-pass power reduction and MCP coast down. 

Time course of power reduction of the core and by-pass and MCP coast down were specified as 

boundary conditions based on experimental data, see Figure D-9, D-10, and D-6. 

In the beginning of the experiment the high break flow was not compensated so the primary 

system was emptying. Calculated higher primary system depressurization rate caused quicker 

emptying of pressurizer. The pressurizer portion above heaters (gauges YP01L02) was emptied 

at +13 s in the calculation whereas in experiment at 16 s. 

The pressure in steam generators is presented in Figure D-11. The measured and calculated 

pressure increased rapidly to a peak 6.86 MPa in experiment whereas to 7.10 MPa in the 

calculation after isolation valve at common part steam header RA06S01 was closed. The 

secondary system pressure did not achieve a value of BRU-A activation neither in experiment 

nor in calculation. Calculated secondary pressure decrease was slightly higher than in the 

experiment. It was probably due to the used heat structure correlation that is not capable to 

perfectly model the coil tube bundle in the steam generator. 

When the primary system pressure dropped below the secondary system pressure (between 80 

and 90 s in the experiment and approximately at 55 s in the calculation), steam generator did 

not remove primary heat any more. The end of removing primary heat to secondary system was 

followed by temporary decreasing primary system depressurization rate and decreasing primary 

system cooldown rate see Figure D-1. 

As the primary pressure continued to decrease below 10.88 MPa and HPI injection begun (at 18 

s in the experiment and in 13 s in calculation), mass flow 0.105 kg/s was injected to hot leg loop 

#1, see Figure D-5. 
5-6 

Primary system pressure was continuously decreasing so the primary system pressure dropped 

below accumulators pressure at +108 s in the experiment and at +100 s in the calculation.. 

Condensation of some of the steam in the system by the cold ECC liquid caused the 

depressurization rate to increase. When level in accumulators fell down to 0.1 m accumulators 

were cut off at +489-500 s in experiment and at +528-535 s in calculation. The accumulators 

injection was followed by LPI injection at + 383 s in experiment and at + 405 s in calculation, 

mass flow 0.2 kg/s was provided to hot and cold leg loop #1. As the system pressure 

decreased, the total ECC injection could compensate decreasing break flow. Collapsed level in 

vessel achieved minimal value approximately 600 s (see Figure D-22) after this time slowly 

increased until ECC injection was terminated at +3697 s in experiment and at +3770 s in 

calculation. HPI and LPI injections were terminated when total water delivery achieved 1.72 m3

– simulation of emptying ECC tanks. After ECC injection was terminated drainage of vessel 

begun again, see pressure differences in core portion of vessel (YC01DP07 – DP10) in Figure 

D-18 and D-19. 

Deviation of calculated cladding temperature from experiment data was up to 5 °C 

approximately until +2600 s. During following time course deviation increased up to 20 °C. 

Increasing deviation of calculated cladding temperature from the experiment was caused by 

lower calculated average void fraction in the core region of vessel – it means that RELAP5 

calculated less liquid in the core than was measured in experiment, see pressure differences in 

the core portion of the reactor vessel (YC01DP07 – DP10) Figure D-18 and D-19. Possible 

reason of under prediction of core liquid inventory were that CCFL at the core exit was 

preventing liquid from falling back into the core region in the calculation, that inter-phase drag 

was over predicted at the core exit, again preventing liquid from separating from vapor flow and 

draining the lower portion of the reactor vessel. 

In the calculation, the heat-up of top portion of core started at +4640 s, it was approximately 700 

s earlier than in the UP-11-07 test, see Figure D-3. Calculated earlier start of heat-up was 

caused by less calculated water inventory in the core than in the test - possible reasons are 

mentioned in a previous paragraph. Calculated heat-up rate was higher than those in the test, 

see Figure D-3. Similar behavior was observed in other post-test calculations of an upper 

plenum small break performed in the code version RELAP5/MOD3.2. (Ref. 7, 8) and also in the 

code CATHARE (Ref. 11). 

At +5593 s power supply to core and by-pass was switched off and experiment was terminated. 

Table 10: Chronology of main events (TRACE and RELAP5 calc. vs. comparison) 

Event Time [s] 

UP-11-07 RELAP5 TRACE 

Initiating Event – break opens 0 0 0 

Pressure in UP (YC01P16) < 13.73 MPa – SCRAM signal 2 3.4 1 

Start of MCP coastdown 2 3.5 1 

Start of core and core by-pass power reduction 3 3.9 2.5 

Stop of feedwater flow supply 9 11 10 

Stop of steam removal from steam generators 11 11 10 

Pressure in UP (YC01P16) < 10.88 MPa 11 10 8 
5-7 

Event Time [s] 

UP-11-07 RELAP5 TRACE 

Start of HPIS injection into Loop #1 18 13 17 

PRZ empty (according to YP01L02) 16 13 15 

UP pressure < SG pressure 84 55 60 

Pressure in UP/DC < pressure in ACCUs 108 ÷ 110 100 ÷ 110 121 ÷ 129

Start of ACCU-1 injection 110 100 121 

Start of ACCU-2 injection (switched off) - - - 

Start of ACCU-3 injection 108 109 129 

Start of ACCU-4 injection 108 110 127 

Start of LPIS injection into Loop #1 383 405 347 

End of accumulators injection (TH01-04L01 < 0.1 m) 489 ÷ 500 528 ÷ 535 579 ÷ 632

End of injection of LPIS injection into hot leg of loop #1 3648 3770 3730 

End of injection of LPIS and HPIS injection into hot leg of 

loop #1 

3697 3770 3730 

Heat up of top portion of core (YC01T10) 5323 4640 4950 

End of the experiment 5593 5593 5593 

5.5 Quantitative Assessment of the Calculations 

To quantify agreement of presented TRACE and RELAP5 calculations the figure of merit (FOM) 

was evaluated using software ACAP (Automated Code Assessment Program), which is a part of 

the software package SNAP. Settings of ACAP was based on Ref 25 including choice of 

particular metrics and their weighting factors - and see the Table 11. 

Table 11: ACAP metrics settings 

Metric name Abbreviation Weighting factor 

D'Auria Fast Fourier Transformation FFT 0.35 

Mean Error Magnitude MEM 0.35 

Size-Independent (Pred - Perf) Norm SI-PMPN 0.15 

Degree of Randomness DOR 0.15 

To assess the value of FOM, acceptability criterions were established on the basis of Ref 26, 

where the FFTB method (Fast Fourier Transform Based Method) is described. FOM 

acceptability criterions were based on AAtot (total average amplitude). Value of AAtot is 

transformed to FOM using the equation of D'Auria FFT metric: 

Where k is weighted frequency importance factor and value k = 0 was applied, which means 

that pure magnitude error is evaluated using D'Auria FFT metric. The next Table 12 contains 

values of acceptability criterions range and their meaning. 

Table 12: Acceptability criterions 
5-8 

AAtot range FOM range Abbreviation Color indication 

AAtot ≤ 0.30 FOM ≥ 0.77 Very good code predictions green 

0.30 < AAtot ≤ 0.50 0.67 ≤ FOM < 0.77 Good code predictions blue 

0.50 < AAtot ≤ 0.70 0.59 ≤ FOM < 0.67 Poor code predictions orange 

AAtot > 0.70 FOM < 0.59 Very poor code predictions red 

To assess TRACE and RELAP5 calculation, the representative set of 65 parameters were 

chosen including: 

 Primary pressure 

 Fuel cladding temperature 

 Pressurizer water level 

 Break flow 

 RPV Pressure drops 

 LOOPs pressure drops and mass flow rates 

 Accumulator water levels and pressures 

To evaluate overall FOM uniform weighting factors were used for each of parameters. 

The UP-11-07 experiment was a long lasting transient where many different TH phenomena 

were expected. In order to carefully assess both TRACE and RELAP5 calculations the whole 

time course isdivided into three time windows of interest as follows: 

 W1: 0 ÷ 100 s – an early stage of the test when only HPIS flow was provided 

 W2: 100 ÷ 3800 s – a middle stage of the test when HPIS, LPIS and accumulators flow 

was provided 

 W3: 3800 ÷ 5600 s – the final stage of the test when no ECCS flow was provided 

To assess the whole time course of the test FOM calculations with no time segmentation was 

performed as well. The following table contains all evaluated FOMs for all time windows of 

interest. To make results more readable color indication mentioned in the Table 12 was applied. 

FOM = 1 means the best agreement and FOM = 0 the worst agreement. Location of PSB-VVER 

measurements is depicted in Appendix B for evaluated parameters. 
Table 13: Evaluation of FOM for UP-11-07 test 

RELAP TRACE 

W0 W1 W2 W3 W0 W1 W2 W3 

Parametr TAG 0 – 5596 0 – 100 100 – 3800 3800-5596 0 – 5596 0 – 100 100 – 3800 3800-5596 Water level in accumulator of ECCS TH01B01 TH01L01 0.967 0.820 0.965 N/A 0.939 0.827 0.935 N/A Pressure of water in accumulator of ECCS TH01B01 TH01P01 0.955 0.751 0.951 N/A 0.920 0.747 0.912 N/A Water level in accumulator of ECCS TH02B01 TH02L01 0.945 0.921 0.947 N/A 0.945 0.922 0.946 N/A Pressure of water in accumulator of ECCS TH02B01 TH02P01 0.935 0.811 0.935 N/A 0.941 0.808 0.944 N/A Water level in accumulator of ECCS TH03B01 TH03L01 0.946 0.897 0.943 N/A 0.936 0.905 0.932 N/A Pressure of water in accumulator of ECCS TH03B01 TH03P01 0.974 0.880 0.971 N/A 0.942 0.890 0.936 N/A Water level in accumulator of ECCS TH04B01 TH04L01 0.952 0.929 0.950 N/A 0.937 0.929 0.932 N/A Pressure of water in accumulator of ECCS TH04B01 TH04P01 0.977 0.888 0.975 N/A 0.947 0.892 0.942 N/A Break line flow rate XL01F01 0.757 0.821 0.678 0.530 0.762 0.813 0.684 0.539 Press drop in HL of loop 1 (elevation part) YA01DP02 0.597 0.841 0.593 0.653 0.770 0.852 0.751 0.652 Press drop in CL of loop 1 (from SG outlet to the flow meter) YA01DP04 0.729 0.894 0.662 0.640 0.755 0.913 0.688 0.532 Press drop in CL of loop 1 (from flow meter to mechanical filter) YA01DP05 0.732 0.887 0.724 0.791 0.704 0.808 0.681 0.803 Press drop in CL of loop 1 (section with a mechanical filter, pump and regulating valve) YA01DP06 0.638 0.829 0.611 0.635 0.692 0.845 0.660 0.824 Press drop on the filter in CL of loop 1 YA01DP08 0.747 0.763 0.742 0.677 0.755 0.739 0.754 0.651 Press drop on MCP of loop 1 YA01DP09 0.646 0.886 0.609 0.805 0.678 0.916 0.641 0.598 Press drop in "cold" header of SG 1 YB01W01 YA01DP13 0.871 0.878 0.713 0.433 0.853 0.816 0.704 0.431 Press drop in "hot" header of SG 1 YB01W01 YA01DP14 0.694 0.913 0.590 0.613 0.893 0.916 0.774 0.618 Flow rate in the pump circulation loop YD01D01 YA01F01 0.698 0.784 N/A N/A 0.705 0.805 N/A N/A Press drop in HL of loop 2 (elevation part) YA02DP02 0.627 0.709 0.359 0.752 0.843 0.848 0.748 0.751 Press drop in CL of loop 2 (from SG outlet to the flow meter) YA02DP04 0.884 0.847 0.761 0.748 0.861 0.818 0.764 0.731 Press drop in CL of loop 2 (from the flow meter to mechanical filter) YA02DP05 0.751 0.905 0.710 0.726 0.740 0.841 0.704 0.694 
RELAP TRACE 

W0 W1 W2 W3 W0 W1 W2 W3 

Parametr TAG 0 – 5596 0 – 100 100 – 3800 3800-5596 0 – 5596 0 – 100 100 – 3800 3800-5596 Press drop in CL of loop 2 (section with mechanical filter, pump and regulating valve) YA02DP06 0.744 0.919 0.734 0.648 0.734 0.922 0.719 0.648 Press drop on filter in CL of loop 2 YA02DP08 0.700 0.718 0.706 0.597 0.716 0.701 0.731 0.594 Press drop on MCP of loop 2 YA02DP09 0.758 0.860 0.750 0.572 0.721 0.884 0.703 0.571 Press drop in "cold" header of SG 2 YB02W01 YA02DP13 0.905 0.891 0.790 0.733 0.853 0.781 0.715 0.732 Press drop in "hot" header of SG 2 YB02W01 YA02DP14 0.773 0.879 0.464 0.530 0.898 0.869 0.779 0.534 Flow rate in the pump circulation loop YD02D01 YA02F01 0.668 0.819 N/A N/A 0.653 0.826 N/A N/A Press drop in HL of loop 3 (elevation part) YA03DP02 0.805 0.815 0.707 0.678 0.807 0.796 0.737 0.674 Press drop in CL of loop 3 (from SG outlet to the flow meter) YA03DP04 0.839 0.852 0.757 0.722 0.826 0.806 0.756 0.720 Press drop in CL of loop 3 (from the flow meter to mechanical filter) YA03DP05 0.712 0.901 0.687 0.390 0.740 0.848 0.671 0.723 Press drop in CL of loop 3 (section with mechanical filter, pump and regulating valve) YA03DP06 0.687 0.909 0.689 0.478 0.727 0.915 0.712 0.654 Press drop on filter in CL of loop 3 YA03DP08 0.667 0.687 0.613 0.574 0.712 0.677 0.661 0.638 Press drop on MCP of loop 3 YA03DP09 0.710 0.898 0.717 0.410 0.732 0.923 0.717 0.751 Press drop in "cold" header of SG 3 YB03W01 YA03DP13 0.863 0.877 0.773 0.743 0.867 0.898 0.762 0.739 Press drop in "hot" header of SG 3 YB03W01 YA03DP14 0.913 0.895 0.790 0.759 0.910 0.877 0.789 0.755 Flow rate in the pump circulation loop YD03D01 YA03F01 0.616 0.790 N/A N/A 0.616 0.837 N/A N/A Press drop in HL of loop 4 (elevation part) YA04DP02 0.652 0.822 0.531 0.563 0.671 0.829 0.551 0.562 Press drop in CL of loop 4 (from SG outlet to the flow meter) YA04DP04 0.917 0.896 0.743 0.547 0.853 0.798 0.738 0.539 Press drop in CL of loop 4 (from the flow meter to mechanical filter) YA04DP05 0.694 0.916 0.669 0.532 0.687 0.841 0.661 0.525 Press drop in CL of loop 4 (section with mechanical filter, pump and regulating valve) YA04DP06 0.679 0.883 0.689 0.518 0.677 0.891 0.687 0.513 Press drop on filter in CL of loop 4 YA04DP08 0.708 0.534 0.715 0.612 0.707 0.533 0.716 0.602 Press drop on MCP of loop 4 YA04DP09 0.727 0.879 0.731 0.586 0.724 0.900 0.727 0.584 Press drop in "cold" header of SG 4 YB04W01 YA04DP13 0.889 0.883 0.720 0.457 0.857 0.825 0.685 0.455 
RELAP TRACE 

W0 W1 W2 W3 W0 W1 W2 W3 

Parametr TAG 0 – 5596 0 – 100 100 – 3800 3800-5596 0 – 5596 0 – 100 100 – 3800 3800-5596 Press drop in "hot" header of SG 4 YB04W01 YA04DP14 0.818 0.928 0.465 0.568 0.825 0.816 0.446 0.543 Flow rate in the pump circulation loop YD04D01 YA04F01 0.622 0.816 N/A N/A 0.615 0.774 N/A N/A Press drop on DC (upper part with inlet chamber) YC01DP01 0.816 0.413 0.780 0.594 0.853 0.587 0.809 0.595 Press drop on DC (inlet in the middle part) YC01DP02 0.753 0.875 0.741 0.757 0.713 0.837 0.703 0.755 Press drop on DC (middle part) YC01DP03 0.643 0.938 0.563 0.666 0.684 0.915 0.614 0.682 Press drop on DC (outlet from the middle part) YC01DP04 0.667 0.899 0.694 0.438 0.710 0.869 0.747 0.447 Press drop in lower DC chamber YC01DP05 0.772 0.890 0.773 0.592 0.804 0.929 0.803 0.628 Press drop between DC LC and CS YC01DP06 0.734 0.888 0.721 0.609 0.723 0.899 0.703 0.674 Press drop in the channel with fuel rod simulators (lower part) YC01DP07 0.619 0.842 0.619 0.596 0.615 0.857 0.619 0.571 Press drop in the channel with fuel rod simulators (inlet in the middle part) YC01DP08 0.542 0.837 0.549 0.523 0.607 0.855 0.610 0.571 Press drop in the channel with fuel rod simulators (outlet from the middle part) YC01DP09 0.618 0.872 0.523 0.687 0.694 0.867 0.607 0.727 Press drop in the channel with fuel rod simulators (upper part) YC01DP10 0.587 0.871 0.554 0.610 0.691 0.938 0.662 0.683 Press drop between CS and UP YC01DP11 0.651 0.886 0.555 0.692 0.731 0.780 0.670 0.722 Press drop in UP (on the insert) YC01DP12 0.724 0.922 0.675 0.754 0.749 0.819 0.715 0.770 Press drop in UP (lower part) YC01DP13 0.763 0.889 0.751 0.745 0.752 0.744 0.742 0.740 Press drop in UP (middle part) YC01DP14 0.791 0.857 0.673 0.545 0.777 0.828 0.658 0.566 Press drop in UP (top piece) YC01DP15 0.899 0.870 0.745 0.516 0.882 0.784 0.731 0.516 Press drop between DC and UP YC01DP16 0.698 0.899 0.661 0.698 0.701 0.823 0.666 0.698 Press drop on CS BP YC01DP17 0.581 0.938 0.563 0.609 0.637 0.910 0.621 0.638 Pressure of coolant in UP YC01P16 0.966 0.948 0.967 0.533 0.945 0.935 0.950 0.570 Fuel rod simulator wall temperature (protection) YC01T10 0.820 0.828 0.946 0.772 0.777 0.949 0.949 0.677 Water level in pressurizer YP01B01 YP01L02 0.948 0.948 N/A N/A 0.944 0.938 N/A N/A FOM avg - 0.764 0.853 0.715 0.618 0.779 0.841 0.738 0.637

6-1 

6. RUN STATISTICS 

The transients were calculated on calculation server with Intel Xeon 5440 processor 2.83 GHz 

under GNU/Linux Debian 5.0 Lenny x64. The run statistics is shown in the following Table 14. 

The TRACE calculation run substantially slower than real time due to the application of 3-D 

vessel components. 

Table 14: Run statistics 

 RELAP5 TRACE 

Number of components 310 336 

Number of time steps 280 092 381 186 

Transient time 5 600 s 5600 s 

CPU time 2 818 s 114 806s 

CPU time / Transient time 0.503 20.5 

7-1 

7. CONCLUSIONS

The main goal of these analyses was to assess the TRACE TH code and its predecessor the 

RELAP5/MOD3.3 using the 11% upper plenum break UP-11-07 in the large scale test facility 

PSB-VVER. The second reason of using two TH codes, RELAP5 and TRACE, is to compare 

overall capability of the new code to its predecessor to catch all important phenomena that take 

place/occur during investigated transients. A part of these analyses is quantitative assessment 

of agreement of the calculations against the experiment data that can help identify pros and 

cons of an applied way of modeling integral test facility in an environment of assessed codes. 

Comparisons of both post-test TRACE and RELAP5 calculations with experiment data proved 

that both TRACE and RELAP5 codes are capable to model PSB-VVER integral system effects 

reasonably. The calculated time courses of the main facility parameters was similar to that of 

the test, indicating that all of the significant events that occurred in the test were present in the 

calculation. 

To quantify errors/deviation of presented TRACE and RELAP5 calculations the figure of merit 

(FOM) was evaluated using software ACAP. FOMs of 65 main measured and calculated 

parameters were evaluated analogously for TRACE and RELAP5 calculations. The following 

table shows the final average FOM evaluated for both calculations at pre-defined time windows 

of interest. 

FOM avg 

W0 W1 W2 W3 

Time 0 – 5596 0 – 100 100 – 3800 3800-5596 

RELAP5 0.76

good prediction

0.85

very good prediction

0.72

good prediction

0.62

poor prediction

TRACE 0.78

very good prediction

0.84

very good prediction

0.74

good prediction

0.64

poor prediction

Presented overall final FOMs (time window W0) prove that the both codes predicted behavior of 

test facility during the whole transient acceptable, although TRACE prediction seems “slightly” 

better. It is clearly visible that better agreements were reached during the early stage of the 

transient whereas the worst agreements were identified in the end transient. These results 

corresponds to the duration of the UP-11-07 test (5600 s), that was longer than “common” 

LOCA tests with duration of tenths of seconds. The accumulation of minor deviations (e.g. void 

distribution in the primary circuit in the UP-11-07 test) might lead to gradual increasing of 

deviations of main calculated parameters. 

However the quantitative assessment gave mainly good or very good predictions of the selected 

main parameters (both for TRACE and RELAP5), following particular discrepancies were 

identified. Both codes predicted different liquid distribution over reactor vessel during the ECC 

injection part of experiments comparing to the measured data. TRACE and RELAP5 calculated 

less liquid in the core region of the vessel and more uniform core axial void profile than was 

measured. Possible reason of under prediction of core liquid inventory might be following: CCFL 

at the core exit was preventing liquid from falling back into the core region in the calculation, 
7-2 

that inter-phase drag was over predicted at the core exit, again preventing liquid from separating 

from vapor flow and draining the lower portion of the reactor vessel. The different liquid 

inventory in the core caused faster heat-up of top portion of the core comparing to the measured 

data. 

The calculation cost of the TRACE calculation was much higher than RELAP5 calculation. 

Worse time efficiency of the TRACE calculation was caused by using a very fine discretized 

VESSEL component representing FRS and UP. VESSEL component consists of 768 cells 

whereas the full RELAP input deck consists of 536 cells. If number of cells of VESSEL 

component is reduced to 40% it will increase calculation speed approximately 4 times, based on 

our experience. 
8-1 

8. REFERENCES

1. TRACE V5.0 THEORY MANUAL, Field Equations, Solution Methods, and Physical Models; 

Division of Systems Analysis, Office of Nuclear Regulatory Research, U. S. Nuclear 

Regulatory Commission, Washington, DC 20555-0001; May 2012 

2. TRACE V5.0 USER’S MANUAL, Volume 1: Input Specification; Division of Systems 

Analysis; Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, 

Washington, DC 20555-0001; December 2011 

3. TRACE V5.0 USER’S MANUAL, Volume 2: Modeling Guidelines; Office of Nuclear 

Regulatory Research, U. S. Nuclear Regulatory Commission, Washington, DC 20555-0001; 

July 2011 

4. O. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of VVER 1000. 

Description of PSB-VVER Test Facility; Moscow Power Engineering Institute (Technical 

University), NPP Department; Moscow; 2008 

5. O. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of VVER 1000. Upper 

Plenum Break 11%; Moscow Power Engineering Institute (Technical University), NPP 

Department; Moscow; 2008 

6. P. Heralecký; Engineering Handbook modelu PSB-1.0; TES-Z-09-175; TES s.r.o. Třebíč; 

prosinec 2009 

7. P.D. Bayless, O. Melikhov, V. Melikhov, Y. Parfenov, O. Gavritenkova, I. Elkin, I. Lipatov; 

RELAP5/MOD3.2 Assessment Using INSC SP-PSBV1; Seventh International Information 

Exchange Forum On SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS OF VVER 

AND RBMK TYPES; Slovakia, Piešťany; 28-30 October 2003 

8. B. Neykov, M. Cherubini, F.D. Auria; Relap 5 Thermal Hydraulic analysis 11% upper plenum 

leakage for PSB-WWER test facility and WWER NPP 

9. A. Del Nevo, F.D. Auria, M. Mazini, M. Vylov, I.V. Elkin, A. Suslov; The Design of PSBVVER Experiments Relevant to Accident Management; Journal of Power and Energy 

Systems; Vol2, No. 1, 2008 

10. V. Blinkov, O.I. Melikhov; Effect of Steam Generátor Nodalization on RELAP5 Simulation of 

Loss-of-Feed-Water Transient in PSB-VVER Facility; Seventh International Information 

Exchange Forum On SAFETY ANALYSIS FOR NUCLEAR POWER PLANTS OF VVER 

AND RBMK TYPES; Slovakia, Piešťany; 27-31 October 2003 

11. L. Sabotinov, P. Chevrier; Post-Test Analysis of 11% Break at PSB-VVER Experimental 

Facility using Cathare 2 Code; Journal of Power and Energy Systems; Vol2, No. 2, 2008 

12. Thermal-Hydraulic Codes News, Volume 09, No.3&4; U.S. Nuclear Regulatory Commission; 

December 2009 

13. P. Heralecký; Srovnávací výpočet testu PSB-VVER UP-11-07 v prostředí kódu TRACE 5.0; 

TES-Z-09-179; TES s.r.o.; Třebíč; prosinec 2009 

14. Preliminary Safety Assessment Report, Rev.2 - Input Data for Safety Analysis of NPP 

Dukovany, Revision V5. 01-March-2007 
8-2 

15. Staudenmeier J. TRACE Code Modelling Capabilities. Proceedings of TRACE/SNAP User 

Workshop. Potomac, Maryland, 27 – 29 March 2006 

16. Bajorek S. at al. TRACE V5.0 Assessment Manual – Main Report. Division of Risk 

Assessment and Special Projects, Office of Nuclear Regulatory Research, U. S. Nuclear 

Regulatory Commission 

17. Petruzzi A. and Auria F. Thermal-Hydraulic System Codes in Nuclear Reactor Safety and 

Qualification Procedures. Science and Technology of Nuclear Installations Volume 2008, 

Article ID 460795 

18. WNA Reactor Database. March 3, 2008 

19. OECD Support Group on the VVER Thermal-Hydraulic Code Validation Matrix. Validation 

Matrix for the Assessment of Thermal-Hydraulic Codes For VVER LOCA and Transients. 

NEA/CSNI/R(2001)4. 01 June 2001, Paris, France 

20. O. I. Melikhov, V. I. Melikhov; Analysis of LOCA thermal-hydraulics of VVER 1000. Singlephase characteristics of the PSB-VVER main circulation pumps; Moscow Power 

Engineering Institute (Technical University), NPP Department; Moscow; 2008 

21. RELAP5/MOD3.3 CODE MANUAL VOLUME I: Code Structure, System Models, and 

Solution Methods; Information System Laboratories, Inc., Rockville Maryland, Idaho Falls, 

Idaho; October 2010 

22. RELAP5/MOD3.3 CODE MANUAL VOLUME II: User’s Guide and Input Requirements; 

Information System Laboratories, Inc., Rockville Maryland, Idaho Falls, Idaho; October 2010 

23. RELAP5/MOD3.3 CODE MANUAL VOLUME V: User’s Guidelines; Information System 

Laboratories, Inc., Rockville Maryland, Idaho Falls, Idaho; October 2010 

24. P. Heralecký; RELAP5/MOD3.3 Assessment Using Experiments in the PSB-VVER Facility; 

TES-Z-12-049; TES s.r.o. Třebíč; září 2012 

25. Matthew D. Lazor; Evaluation of Assessment Techniques for Verification and Validation of 

the TRACE Nuclear Systems Code; The Pennsylvania State University, College of 

Engineering; December 2004 

26. A. Prošek, B. Mavko; Quantitative Code Assessment with Fast Fourier Transformation 

Based Method Improved by Signal Mirroring; NUREG/IA-0220; Jožev Stefan Institute; 

Slovenia; Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission; 

Washington DC, 20555-0001; December 2009 
A-1 

APPENDIX A INPUT DECK NODALISATION SCHEMES 
A-2 

Figure A-1: TRACE Nodalization Scheme of Primary Circuit of PSB-VVER Facility 
A-3 

Figure A-2: TRACE Nodalization Scheme of Secondary Circuit of PSB-VVER Facility 
A-4 

Figure A-3: RELAP5 Nodalization Scheme of Primary Circuit of PSB-VVER Facility 
A-5 

Figure A-4: RELAP5 Nodalization Scheme of Secondary Circuit of PSB-VVER Facility 

B-1 

APPENDIX B MEASUREMENT LOCALISATION AT PSB-VVER 

FACILITY 
B-2 

Figure B-1: PSB-VVER Reactor Model Measurements 
B-3 

Figure B-2: PSB-VVER Loop 1 and SG-1 Model Measurement 

C-1 

APPENDIX C COMPLETE SET OF COMPARISON PLOTS FOR 

TRACE CALCULATION 
C-2 

Figure C-1: Primary Pressure 

Figure C-2: Primary Pressure (detail) 
C-3 

Figure C-3: Fuel Cladding Temperature (Top of the Core) 

Figure C-4: Presurizer Level 
C-4 

Figure C-5: HPIS + LPIS Flow (Boundary Condition) 

Figure C-6: MCP Rotor Speed (Boundary Condition) 
C-5 

Figure C-7: Accumulators Levels 
C-6 

Figure C-8: Accumulators Pressure 
C-7 

Figure C-9: Fuel Rod Simulator Power (Boundary Condition) 

Figure C-10: Core By-pass Power (Boundary Condition) 
C-8 

Figure C-11: Secondary Side Pressures 
C-9 

Figure C-12: Steam Generators Levels 
C-10 

Figure C-13: Loop 1 Temperatures 
C-11 

Figure C-14: Loop 2 Temperatures 
C-12 

Figure C-15: Loop 3 Temperatures 
C-13 

Figure C-16: Loop 4 Temperatures 
C-14 

Figure C-17: Pressure Differences DP01-DP04 (DC) 
C-15 

Figure C-18: Pressure Differences DP05-DP08 (Lower Plenum + Lower Part of FRS) 
C-16 

Figure C-19: Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum) 
C-17 

Figure C-20: Pressure Differences DP13-DP16 ( Upper Part of Upper Plenum) 
C-18 

Figure C-21: Break Flow 

Figure C-22: Reactor Collapsed Level 
D-1 

APPENDIX D COMPLETE SET OF COMPARISON PLOTS FOR 

RELAP5 CALCULATION 
D-2 

Figure D-1: Primary Pressure 

Figure D-2: Primary Pressure (detail) 
D-3 

Figure D-3: Fuel Cladding Temperature (Top of the Core) 

Figure D-4: Presurizer Level 
D-4 

Figure D-5: HPIS + LPIS Flow (Boundary Condition) 

Figure D-6: MCP Rotor Speed (Boundary Condition) 
D-5 

Figure D-7: Accumulators Levels 
D-6 

Figure D-8: Accumulators Pressure 
D-7 

Figure D-9: Fuel Rod Simulator Power (Boundary Condition) 

Figure D-10: Core By-pass Power (Boundary Condition) 
D-8 

Figure D-11: Secondary Side Pressures 
D-9 

Figure D-12: Steam Generators Levels 
D-10 

Figure D-13: Loop 1 Temperatures 
D-11 

Figure D-14: Loop 2 Temperatures 
D-12 

Figure D-15: Loop 3 Temperatures 
D-13 

Figure D-16: Loop 4 Temperatures 
D-14 

Figure D-17: Pressure Differences DP01-DP04 (DC) 
D-15 

Figure D-18: Pressure Differences DP05-DP08 (Lower Plenum + Lower Part of FRS) 
D-16 

Figure D-19: Pressure Differences DP09-DP12 (FRS + Lower Part of Upper Plenum) 
D-17 

Figure D-20: Pressure Differences DP13-DP16 ( Upper Part of Upper Plenum) 
D-18 

Figure D-21: Break Flow 

Figure D-22: Reactor Collapsed Level 



UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA-0449 Post-Test Analysis of Upper Plenum 11% Break at PSB-VVER Facility using 

TRACE V5.0 and RELAP5/MOD3.3 Code