NUREG/IA-0435 

International 

 Agreement Report

Assessment of RELAP5/MOD3.3 and TRACE V5.0

Computer Codes against LOCA Test Data from 

PSB-VVER Test Facility 

Prepared by:

M. Kynčl, R. Pernica

Research Centre Řež,TSO 

Hlavní 130

250 68 Husinec-Řež, 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: July 2013

Date Published: December 2013

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0435 

International 

 Agreement Report

Assessment of RELAP5/MOD3.3 and TRACE V5.0

Computer Codes against LOCA Test Data from 

PSB-VVER Test Facility 

Prepared by:

M. Kynčl, R. Pernica

Research Centre Řež,TSO 

Hlavní 130

250 68 Husinec-Řež, 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: July 2013

Date Published: December 2013

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission

 iii

ABSTRACT 

This report documents the USNRC system thermal-hydraulic codes assessment performed at 

Research Centre Rez. RELAP5/MOD3.3 and TRACE V5.0 computer codes were assessed 

against the experimental data from two LOCA tests carried out at PSB-VVER test facility built 

and operated at Electrogorsk Research and Engineering Center on NPPs Safety in Russia. 

Experimental data and all necessary information for developing the facility database and facility 

model notebook were available through the OECD/NEA/CSNI/WGAMA PSB Project. 

In the RELAP5/MOD3.3 case, in addition to the best estimate post test analysis of LB LOCA 

experiment denoted as CL-2x100-01 by EREC and as Test-5a as part of the OECD PSB 

project, the uncertainty and sensitivity analysis was performed, and the results are presented 

here. Double ended cold leg guillotine break with initial power scaled further down to 15% was 

simulated in Test-5a. 

In the TRACE V5.0 case, the best estimate post test analysis of SB LOCA experiment denoted 

as CL-4.1-03 by EREC and as Test-3 as part of the OECD PSB project was carried out, and the 

results are presented here. A 4.1% cold leg break was simulated in Test-3. The test scenario 

applied was very close to the scenario realized in LOBI test BL-34. 

The RELAP5/MOD3.3 thermal-hydraulic model of PSB-VVER test facility was developed 

without using the SNAP Model Editor. The development of the TRACE V5.0 thermal-hydraulic 

model of the PSB-VVER test facility started from RELAP5/MOD3.3 model in SNAP 

environment. Firstly, the reference case input deck from RELAP5/MOD3.3 study was imported 

to SNAP and secondly, the thermal-hydraulic model was converted to TRACE V5.0. The 

resulting TRACE V5.0 thermal-hydraulic model was systematically verified using the Test-5a 

experimental data. After that, only few computer runs were needed to properly define the initial 

and boundary conditions of Test-3 in order to validate the TRACE V5.0 thermal-hydraulic model 

of the PSB-VVER test facility for analysis of SB LOCA. 

 v

CONTENTS 

Page 

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

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

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

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

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

2. PSB-VVER FACILITY DESCRIPTION ...................................................................... 3

3. RELAP5/MOD3.3 CODE ASSESSMENT ............................................................... 19

3.1 PSB-VVER Test-5a Description ................................................................................... 19

3.2 RELAP5/MOD3.3 Thermal-Hydraulic Model of PSB-VVER Test Facility .................. 23

3.3 RELAP5/MOD3.3 PSB-VVER Test Facility Model Validation ..................................... 40

3.3.1 Reference Case ........................................................................................... 40

3.3.2 Uncertainty Study........................................................................................ 44

3.3.3 Sensitivity Study ......................................................................................... 52

3.4 Conclusions on RELAP5/MOD3.3 Code Assessment ............................................... 56

4. TRACE V5.0 CODE ASSESSMENT ....................................................................... 59

4.1 PSB-VVER Test-3 Description ..................................................................................... 59

4.2 TRACE V5.0 Thermal-Hydraulic Model of PSB-VVER Test Facility .......................... 63

4.3 TRACE V5.0 PSB-VVER Test Facility Model Validation ............................................. 72

4.3.1 Reference Case ........................................................................................... 72

4.3.2 Preliminary sensitivity study...................................................................... 80

4.4 Conclusions on TRACE V5.0 Code Assessment ....................................................... 81

5. REFERENCES ........................................................................................................ 83

APPENDIX A ........................................................................................................................... 85

APPENDIX B ......................................................................................................................... 113

APPENDIX C ......................................................................................................................... 149

 vii

FIGURES 

Figure 1 PSB-VVER test facility .............................................................................................. 4

Figure 2 Thermal-hydraulic diagram of PSB-VVER test facility .......................................... 5

Figure 3 Reactor model measurements ................................................................................. 7

Figure 4 Fuel rod simulator bundle cross section ................................................................ 8

Figure 5 Axial positions of thermocouples (Groups No. 1 and No. 2) ................................ 9

Figure 6 Axial positions of thermocouples (Groups No. 3 and No. 4) .............................. 10

Figure 7 Axial positions of thermocouples (Groups No. 5 and No. 6) .............................. 11

Figure 8 Primary coolant loop model measurements (Loop No. 1) .................................. 12

Figure 9 Primary coolant loop model measurements (Loop No. 2) .................................. 13

Figure 10 Primary coolant loop model measurements (Loop No. 3) ................................ 14

Figure 11 Primary coolant loop model measurements (Loop No. 4) ................................ 15

Figure 12 Pressurizer model measurements ...................................................................... 16

Figure 13 Steam generator models measurements ............................................................ 17

Figure 14 Emergency core cooling system model measurements ................................... 18

Figure 15 R5/M33 reactor pressure vessel model .............................................................. 25

Figure 16 R5/M33 primary coolant system model – Loop No. 1 ........................................ 27

Figure 17 R5/M33 primary coolant system model – Loop No. 2 ........................................ 28

Figure 18 R5/M33 primary coolant system model – Loop No. 3 ........................................ 29

Figure 19 R5/M33 primary coolant system model – Loop No. 4 ........................................ 30

Figure 20 R5/M33 pressurizer model ................................................................................... 32

Figure 21 R5/M33 emergency core cooling system model ................................................ 33
 viii

Figure 22 R5/M33 secondary coolant system model – Loop No. 1 ................................... 35

Figure 23 R5/M33 secondary coolant system model – Loop No. 2 ................................... 36

Figure 24 R5/M33 secondary coolant system model – Loop No. 3 ................................... 37

Figure 25 R5/M33 secondary coolant system model – Loop No. 4 ................................... 38

Figure 26 R5/M33 main steam line model ............................................................................ 39

Figure 27 TRACE V5.0 reactor pressure vessel model ...................................................... 65

Figure 28 TRACE V5.0 primary coolant system model – Loop No. 4 ................................ 66

Figure 29 TRACE V5.0 pressurizer model ........................................................................... 68

Figure 30 TRACE V5.0 emergency core cooling system model ........................................ 69

Figure 31 TRACE V5.0 secondary coolant system model – Loop No. 4 ........................... 70

Figure 32 TRACE V5.0 main steam line model .................................................................... 71
 ix

TABLES 

Table 1 PSB-VVER test facility subsystems identification .................................................. 6

Table 2 PSB-VVER test facility measurement identification ................................................ 6

Table 3 Test-5a specific PSB-VVER test facility configuration ......................................... 20

Table 4 Test-5a specific PSB-VVER test facility initial conditions .................................... 21

Table 5 Test-5a specific PSB-VVER test facility boundary conditions ............................. 22

Table 6 R5/M33 Test-5a reference case steady state data evaluation – PCS #1 .............. 41

Table 7 Test-5a R5/M33 reference case FFTBM results ..................................................... 42

Table 8 List and description of uncertain input parameters .............................................. 46

Table 9 One-sided (0.95; 0.95) tolerance bounds for the single-valued parameters ....... 51

Table 10 Sensitivity analysis of the single valued output parameters – Spearman’s 

correlation coefficients...........................................................................53

Table 11 List of significant input uncertain parameters for the time trends output 

parameters maxT, UPP, PCM and IBF ................................................................. 55

Table 12 Test-3 specific PSB-VVER test facility configuration ......................................... 60

Table 13 Test-3 specific PSB-VVER test facility initial conditions .................................... 61

Table 14 Test-3 specific PSB-VVER test facility boundary conditions ............................. 62

Table 15 TRACE V5.0 Test-3 reference case steady state data evaluation – PRZ ........... 72

Table 16 TRACE V5.0 Test-3 reference case steady state data evaluation – RPV ........... 74

Table 17 TRACE V5.0 Test-3 reference case steady state data evaluation – PCS ........... 76

Table 18 TRACE V5.0 Test-3 reference case steady state data evaluation – SG ............. 77

Table 19 Test-3 TRACE V5.0 reference case FFTBM results ............................................. 78

 xi

EXECUTIVE SUMMARY 

The purpose of this work is to contribute to the USNRC thermal-hydraulic codes assessment as 

agreed in the CAMP agreement. RELAP5/MOD3.3/Release and TRACE V5.0 Patch 3 computer 

codes are assessed against LOCA test data from the PSB-VVER test facility. 

 xiii

ACKNOWLEDGMENTS 

This project is financially supported by Technology Agency of the Czech Republic: Project No. 

2012TA02021135. 

This report also summarizes the results of the project JC_2/2009, a project on research and 

development and innovation in the Czech Republic with public support. 

 xv

ABBREVIATIONS 

AAtot average amplitude (total) 

ACC accumulator 

ACL accumulator level 

ADS atmospheric dump system; steam dump to atmosphere 

BE best estimate 

BP (core) bypass 

CAMP Code Application and Maintenance Program 

CCFL counter current flow limitation 

CCVM CSNI code validation matrix 

CHF critical heat flux 

CL cold leg 

D discrete statistical distribution 

DAS data acquisition system 

DP differential pressure 

ECCS emergency core cooling system 

EREC Electrogorsk Research and Engineering Center on NPPs Safety 

FFTBM fast Fourier transform based method 

FRSB fuel rod simulator bundle 

gspf grid spacer factor 

HPIS high pressure safety injection system 

IBF integrated break flow 

LN lognormal statistical distribution 

LOCA loss of coolant accident

LPIS low pressure safety injection system 

maxT maximum cladding temperature 

N power of FRSB or 

normal statistical distribution 

n number of calculations 

NPP nuclear power plant 

P pressure 

PCM primary coolant mass 

PCS primary coolant system 

pdf probability distribution function 
 xvi

PRZ pressurizer 

PSB integral experimental equipment 

PWR pressurized water reactor 

R5/M3.3 RELAP/MOD3.3/Release 

SG steam generator 

T temperature 

t1 (s) time of beginning of first FRSB cladding heatup 

t2 (s) time of core quenching 

t3 (s) time of beginning of last FRSB cladding heatup 

tacc1 time of start of accumulator injection 

thpis time of start of high pressure injection 

tmax time of maximum cladding temperature maxT reached during first FRSB 

cladding heatup 

U uniform statistical distribution 

UP upper plenum 

UPP upper plenum pressure 

USNRC United States Nuclear Regulatory Commission 

VVER water cooled water moderated energetical reactor (WWER) 

 mean 

 standard deviation 
1 

1. INTRODUCTION 

The report documents an assessment performed for RELAP5/MOD3.3/Release and TRACE 

V5.0 Patch 3 thermal hydraulic system codes at Research Centre Rez, Czech Republic. The 

computer codes were assessed against the experimental data from two LOCA tests carried 

out at PSB-VVER test facility built and operated at Electrogorsk Research and Engineering 

Center on NPPs Safety in Russia [1]. The system code assessment process is structured in 

[2] as follows: 

1. Formulate assessment matrices for each class of transient studied, e.g., large-break 

LOCA, small-break LOCA. 

2. Select key parameters for these classes. 

3. Perform assessment calculations. Compare the test data with the results of calculations 

for the key parameters. 

4. Perform uncertainty analyses for the capability of the codes to predict test data and plant 

transients. 

Another step can be identified based on [3]: 

5. Perform sensitivity analyses to select important variables, for which the uncertainty 

distribution has to be determined with good accuracy. 

In case of RELAP5/MOD3.3 code assessment, all five steps are addressed below. The 

transient studied is the large-break LOCA experiment, double ended cold leg guillotine break 

with initial power scaled further down to 15% of prototypical [4]. 

In the case of TRACE V5.0 code assessment, the first three steps are addressed below. The 

transient studied is the small-break LOCA experiment, 4.1% cold leg break with initial power 

scaled further down to 15% of prototypical [5]. 

The methodology applied in step 3 is outlined in [6]. Specifically, the test data are compared 

with the results of calculations on the steady state level and on the transient level. In the first 

substep, the code results are checked for whether they are within the measurement 

uncertainty intervals; in the second substep, the methodology based on Fast Fourier 

Transform [7] is applied. 

In the case of RELAP5/MOD3.3 code assessment, the PSB-VVER test facility model had 

been developed, verified, and validated against LOCA test data within the domestic project 

JC_2/2006. 

In the case of TRACE V5.0 code assessment, the PSB-VVER test facility model 

development started with conversion of final reference case RELAP5/MOD3.3 PSB-VVER 

test facility model in SNAP software [8] environment. Further steps in TRACE V5.0 PSBVVER test facility model development, verification, and validation are outlined below. 

The report is organized so that the PSB-VVER test facility description with details addressing 

the measurement system is presented in Chapter 2. RELAP5/MOD3.3 code assessment is 

presented in details in Chapter 3, while TRACE V5.0 code assessment case is presented in 

Chapter 4. 

3

2. PSB-VVER FACILITY DESCRIPTION 

The PSB-VVER integral test facility [1] was built in the Elektrogorsk Research and 

Engineering Center on NPPs Safety (EREC) to model the primary system of a PWR NPP 

unit with VVER-1000 reactor. The facility was designed to preserve a facility to plant 

volumetric ratio 1:300 and elevation ratio 1:1. In the series of tests conducted within the 

OECD/NEA/CSNI/WGAMA PSB-VVER project, the maximum power of fuel rod simulator 

bundle (FRSB) was 1.5MWt, i.e. the power was further scaled down to 15% of prototypical. 

The PSB-VVER test facility is a model of a primary coolant system (PCS) and emergency 

core cooling system (ECCS) of a VVER-1000 NPP with all main components. In addition, the 

secondary system components are included as practical. The facility is also equipped with 

systems for LOCA simulations. 

Figure 1 shows that 4 components are used to simulate RPV in the facility, i.e. the 

downcomer model, the core model, the core model bypass and the upper plenum model. In 

the core model, a bundle of 168 fuel rod simulators is located in a hexagonal channel. Full 

height fuel rod simulators are arranged in a triangular lattice. In addition to lower and upper 

spacer grids, there are 15 spacer grids over the heated length. In the radial direction there 

are three material zones, central nichrome heater rod, periclase insulator, and stainless steel 

cladding. In the case of instrumented fuel rod simulators, two cladding layers are used and 

the thermocouples are located between the external and internal claddings. The core model 

is separated from the upper plenum model by the core upper plate simulator. A VVER-1000 

NPP unit is a four loop plant. In the facility, each loop is simulated separately with hot leg 

pipe model, SG model, loop seal pipe model, RCP model, loop isolation valve and cold leg 

pipe model. The pressurizer surgeline is divided in two by a T-joint so that the pressurizer 

model can be connected either to the hot leg in loop No. 2 or to the hot leg in loop No. 4. 

Although the SG model is a vertical vessel, the primary side arrangement of the vertical hot 

header interconnected by nearly horizontal coiled tubes to the vertical cold header is 

representative of a horizontal steam generator. There are 34 coiled tubes along the height 

per one SG. The length and the diameter of one tube is the same as of the reference plant. 

The low inertia RCP model is used to provide forced coolant circulation in the primary circuit. 

The programmable electromotor frequency is used to achieve the desired initial steady state 

coolant flow and to model a RCP coastdown if requested in case of a transient/accident 

being simulated. The pressurizer model is a full height vessel with corresponding subsystem 

models of pressurizer heaters, pressurizer spray, pressurizer relieve valve, and coolant fill 

and drain. 

In addition to primary system components, Figure 2 shows the ECCS components diagram, 

and secondary system components diagram. Two accumulators (ACC) are connected to the 

upper plenum and other two ACCs are connected to the downcomer. The active safety 

injection system is designed so that various actions of HPIS and LPIS can be simulated 

including the hot leg injection, cold leg injection, upper plenum and downcomer injection. The 

actual configuration of HPIS/LPIS is test dependent. 
4

Figure 1 PSB-VVER test facility 
5

Figure 2 Thermal-hydraulic diagram of PSB-VVER test facility 
6

The PSB-VVER measurement system designated as Data Acquisition System (DAS) 

includes more than 1000 measuring channels. The number of active measurement channels 

and the data record frequency is test specific. The measurement channels, e.g. YC01DP01 

or YP01P01, are identified according to the following rules. The first two capital letters 

designate the test facility functional subsystem. The two digit numbers that follow identify the 

subsystem number, e.g. the primary coolant loops. Next, one or two capital letters identify the 

measured parameter, and finally, two or three digit numbers specify the actual measurement 

channel. The test facility functional subsystem identifiers are presented in Table 1. The 

measured parameters identifiers are presented in Table 2. In both cases, only the 

subsystems and their corresponding measurements which are later used for the code 

assessments are listed. 

Table 1 PSB-VVER test facility subsystems identification 

Test facility technological subsystem Identifier 

Primary coolant system loops YA 

Steam generator YB 

Reactor pressure vessel simulator YC 

Reactor coolant pumps YD 

Pressurizer YP 

Accumulators TH 

Table 2 PSB-VVER test facility measurement identification 

Measured parameter Identifier 

Pressure P 

Differential pressure DP 

Fluid temperature T or TF 

Wall temperature T or TH 

Mass flow rate F 

Liquid level L 

Power N 

Actual measurement locations are identified on following Figures 3 to 14. All figures 

presented in this section are adopted from [1]. 
7

Figure 3 Reactor model measurements 
8

Figure 4 Fuel rod simulator bundle cross section 
9

Figure 5 Axial positions of thermocouples (Groups No. 1 and No. 2) 
10

Figure 6 Axial positions of thermocouples (Groups No. 3 and No. 4) 
11

Figure 7 Axial positions of thermocouples (Groups No. 5 and No. 6) 
12

Figure 8 Primary coolant loop model measurements (Loop No. 1) 
13

Figure 9 Primary coolant loop model measurements (Loop No. 2) 
14

Figure 10 Primary coolant loop model measurements (Loop No. 3) 
15

Figure 11 Primary coolant loop model measurements (Loop No. 4) 
16

Figure 12 Pressurizer model measurements 
17

Figure 13 Steam generator models measurements 
18

Figure 14 Emergency core cooling system model measurements 
19

3. RELAP5/MOD3.3 CODE ASSESSMENT 

As for as large-break LOCA analyses, the RELAP5/MOD3.3 code developmental 

assessment [9] is based on LOFT large-break LOCA test L2-5 [10-12]. Recently, the L2-5 

test was reanalyzed [13] within the frame of larger international program BEMUSE [3]. Four 

out of total 14 participants used RELAP5/MOD3.3 code in [13] and contributed to the 

independent RELAP5/MOD3.3 code assessment for large-break LOCA analyses. The 

authors of this report had participated in the BEMUSE project also with RELAP5/MOD3.3 

code as NRI-1 group. Since the results of the BEMUSE project were already published [3], 

the contribution of NRI-1 group is not repeated here. However, the methodology applied for 

BEMUSE Phases II and III is used here to present another RELAP5/MOD3.3 code 

independent assessment for large-break LOCA analyses. The results of the independent 

RELAP5/MOD3.3 code assessment against large-break LOCA test performed at PSB-VVER 

test facility [4] are reported below. 

The first two steps in the thermal-hydraulic system code assessment process are defined in 

[2] as: 

 Formulate assessment matrices for each class of transient studied, e.g., large-break 

LOCA, small-break LOCA. 

 Select key parameters for these classes. 

When addressing these steps the authors ask the reader to refer to [14]: “The construction of 

VVER Thermal Hydraulic Code Validation Matrix follows the logic of the CSNI Code 

Validation Matrices (CCVM) [12]. Similar to the CCVM it is an attempt to collect together in a 

systematic way the best sets of available test data for VVER specific code validation, 

assessment and improvement, including quantitative assessment of uncertainties in the 

modelling of phenomena by the codes.” Since the PSB-VVER test facility was not in 

operation at the time the report [14] was compiled, the phenomena vs. system test largebreak LOCA matrix refers PSB-VVER as “expected to be suitable”. Test facility vs. 

phenomena, specifically the suitability of test data for the code assessment, is further 

discussed in [4]. The selection of key parameters to be compared is based on [13] and [15]. 

3.1 PSB-VVER Test-5a Description 

The general objectives of PSB-VVER large-break LOCA experiment are listed in [4]: 

 Investigate the thermal-hydraulic response of the VVER-1000 primary system to the cold 

leg double ended guillotine break. 

 Evaluate the capability of the PSB-VVER test facility to simulate the LB LOCA in VVER1000. 

 Obtain experimental data for validation of thermal-hydraulic codes applied to LB LOCA 

analysis of VVER-1000 NPP. 

Internally at EREC, the test is denoted as CL-2x100-01. In the context of PSB-VVER OECD 

Project, the test is denoted as Test-5a. 
20

Test-5a specific PSB-VVER facility configuration is discussed in Table 3. 

Table 3 Test-5a specific PSB-VVER test facility configuration 

Subsystem Description 

Pressurizer Surgeline was connected to the hot leg loop No. 4. 

Core bypass Two orifices with a diameter of 7 mm were installed at the 

inlet/outlet of the core bypass. 

Steam generators 

Under initial steady state operation all SGs were 

connected to each other by the steam header. The 

steady state pressure was adjusted by steam dump valve 

RA06S01 located at common steam line. 

Feedwater 

Under initial steady state operation the SG secondary 

side coolant levels were adjusted by RL01S06, RL02S06, 

RL03S06, and RL04S06 valves. The feedwater 

temperature was close to 214 °C. 

Accumulators 

ACC No. 1 and ACC No. 3 surgelines were connected to 

the upper plenum simulator. ACC No. 4 surgeline was 

connected to the downcomer simulator. ACC No. 2 was 

isolated throughout the test. 

HPIS Two lines of HPIS were connected to cold legs of PCS 

loops No. 1 and No. 3. 

LPIS 

One line of LPIS was connected to PCS loop No. 3 with 

injection points at the hot leg and at the cold leg. One line 

of LPIS was connected to the RPV simulator with 

injection points at the upper plenum simulator and at the 

downcomer simulator. 

Break unit 

The break unit was located in the horizontal section of 

PCS loop No. 3 cold leg between the downcomer 

simulator inlet nozzle and RCP. Break nozzles were 

smooth edged channels with a diameter of 50 mm and a 

length of 350 mm. 

Upper plenum 

Under steady state operation the upper plenum simulator 

warm-up line was opened. The line was closed 2 minutes 

prior to the initiation of the transient. 

The PSB-VVER DAS recorded the measured parameters with a frequency of 20 Hz 

throughout the test. Data from 14 pressure gauges were recorded with a frequency of 

200 Hz. 
21

Test-5a specific PSB-VVER facility initial conditions are presented in Table 4. 

Table 4 Test-5a specific PSB-VVER test facility initial conditions 

Parameter Measurement Value Measurement 

accuracy 

Primary coolant system 

Upper plenum pressure, MPa YC01P16 15.71 ± 0.06 

Hot leg inlet coolant temperature, °C 

YA01T03 321 ± 3 

YA02T03 321 ± 3 

YA03T03 320 ± 3 

YA04T03 323 ± 3 

Cold leg outlet coolant temperature, 

°C 

YA01T02 292 ± 3 

YA02T02 292 ± 3 

YA03T02 290 ± 3 

YA04T02 290 ± 3 

FRSP power, kW YC01N01 1512 ± 15 

BP power, kW YC01N02 16.1 ± 0.7 

Pressurizer level, m YP01L02 6.5 ± 0.3 

Secondary coolant system 

SG pressure, MPa 

YB01P01 7.78 ± 0.05 

YB02P01 7.81 ± 0.05 

YB03P01 7.84 ± 0.05 

YB04P01 7.77 ± 0.05 

SG level, m 

YB01L01 1.68 ± 0.08 

YB02L01 1.70 ± 0.08 

YB03L01 1.67 ± 0.08 

YB04L01 1.71 ± 0.08 

Accumulators 

ACC pressure, MPa 

TH01P01 5.93 ± 0.03 

TH03P01 5.91 ± 0.03 

TH04P01 6.06 ± 0.03 

ACC level, m 

TH01L01 5.65 ± 0.07 

TH03L01 5.65 ± 0.07 

TH04L01 5.69 ± 0.07 
22

Test-5a specific PSB-VVER facility boundary conditions are presented in Table 5. 

Table 5 Test-5a specific PSB-VVER test facility boundary conditions 

Time, s Event 

0 Break simulation 

0 Start of RCPs coastdown simulation 

0 Steam line isolation signal, start of RA06S01 valve closing 

0.4 SCRAM, FRSB and BP power controlled to simulate decay heat 

4.1 SG feedwater isolation, start of RL02S06 valve closing 

4.7 SG feedwater isolation, start of RL01S06 valve closing 

4.7 SG feedwater isolation, start of RL04S06 valve closing 

5.9 SG feedwater isolation, RL01S06 valve closed 

6.5 SG feedwater isolation, RL04S06 valve closed 

8.7 SG feedwater isolation, start of RL03S06 valve closing 

9.9 SG feedwater isolation, RL02S06 valve closed 

10.2 Start of ACC-4 injection 

10.7 Start of ACC-1 and ACC-3 injection 

10.8 SG feedwater isolation, RL03S06 valve closed 

11.0 Pressurizer emptied 

15.5 Steam line isolation signal, valve RA06S01 closed 

40.3 Start of HPIS injection 

40.3 Start of LPIS injection 

89.0 Stop of ACC-1 injection 

92.0 Stop of ACC-4 injection 

107.0 Stop of ACC-3 injection 

159.0 Start of 1st FRSB cladding heatup 

231.3 Stop of RCPs coastdown simulation 

556.0 Stop of 1st FRSB cladding heatup (cladding quenched) 

921.8 Stop of HPIS injection 

921.8 Stop of LPIS injection 

1187.0 Start of 2nd FRSB cladding heatup 

1477 FRSB power switched off – end of the experiment 
23

In PSB-VVER Test-5a the large-break LOCA accident, double ended guillotine cold leg No. 3 

break was simulated with boundary conditions simulating coincident loss of offsite power. 

The FRSB and core bypass power levels were controlled to simulate decay heat power. The 

frequencies of RCP electro motors were programmed to simulate the RCPs coastdown. The 

ECCS actions were simulated. Two periods of FRSB cladding heatup were observed based 

on thermocouple measurements. The first one occurred just after the end of accumulator 

injection and was terminated by the continuing HPIS/LPIS injection. The second FRSB 

cladding heatup occurred after the stop of HPIS/LPIS injection. This heatup was terminated 

by operators after the cladding temperature reached the prescribed level. 

3.2 RELAP5/MOD3.3 Thermal-Hydraulic Model of PSB-VVER Test Facility 

The facility database was compiled in an MS-ACCESS environment. Facility design data and 

thermal-hydraulic characteristics as reported as part of OECD/NEA PSB-VVER Project were 

organized in two tables with a total of 600 records. The facility model notebook was compiled 

in an MS-EXCEL environment developing RELAP5/MOD3.3 T-H model of the PSB-VVER 

facility in several steps up to the code input deck. The input deck for steady state runs 

consists of 31170 lines. 

In the verification step, the initial and boundary conditions for the RELAP5/MOD3.3 PSBVVER facility model were set up to simulate Test-3 of OECD/NEA PSB-VVER project, i.e. the 

small-break LOCA experiment, 4.1% cold leg break with initial power scaled further down to 

15% of prototypical [5]. In the verification step, the “null transient” method was applied to 

reproduce the test specific initial conditions. As the acceptance criterion, the code calculated 

data were requested to be within the corresponding channel measurement accuracy interval. 

The small-break LOCA transient was analyzed in code restart mode with specification of test 

specific trips in order to simulate the boundary conditions of the experiment. As the 

acceptance criterion, the calculated data were requested to reproduce the measured data 

qualitatively (phenomena and trends), with only key parameters (e.g. primary system 

pressure) being compared quantitatively. 

In the validation step, the initial and boundary conditions of the reference case calculations 

(steady state run and transient run) were set up for the facility model in question to simulate 

Test-5a of OECD/NEA PSB-VVER project, i.e. the large-break LOCA experiment, double 

ended guillotine cold leg break with initial power scaled further down to 15% of prototypical 

[4]. This case is described in detail in Chapter 3.3 below. 

In order to facilitate the description of RELAP5/MOD3.3 T-H model of PSB-VVER test facility, 

the reference case steady state R5/M3.3 code input deck was imported to SNAP [8] and 

thermal-hydraulic views were reorganized as shown in Figures 15 to 26. 

The hydraulic part of PSB-VVER facility T-H model consists of 1580 control volumes and 

1841 junctions. Its thermal part consists of 849 heat structures with a total of 6296 mesh 

points. Resulting R5/M3.3 PSB-VVER test facility T-H model is rather detailed. The basic 

idea was to have control volume centers at the actual locations of measurement gauges. 

This approach had to be compromised only if higher order nodalization criteria had been 

applied, e.g. at core model, the node boundaries were placed at grid spacer locations, control 

volume centers at ascending legs were placed at the same elevation as control volumes in 

descending legs, etc. 
24

The nodalization of reactor pressure vessel model is illustrated in Figure 15. Downcomer inlet 

annulus is represented by a total of 12 annulus components in three axial levels and 4 

azimuthal sectors. Single junction components (cn49; n=1, 4) are used to model primary 

coolant cold leg outlets. Multiple junction components (not shown in this view) are used to 

simulate axial and azimuthal flows between the downcomer inlet annular sectors. Multiple 

junction components are also used to connect the downcomer inlet annulus with main 

downcomer that is modelled with pipe component (c030) having 36 axial control volumes. In 

order to represent the actual design of the downcomer lower part, branch component (c031) 

and annulus component (c035) are used. A combination of single junction components 

(c036, c040, c045), pipe components (c037, c051), and annulus component (c041) is used to 

model the lower plenum simulator. The hydraulic part of the core simulator is modelled with 5 

pipe components each having 16 control volumes. Four pipes (c061, c062, c063, and c064) 

represent the azimuthal sector of the corresponding PCS loop. Pipe (065) is used to model 

the central part of the FRSB. Multiple junction components (not shown in this view) are used 

to simulate radial and azimuthal cross flows in FRSB. Two fuel rod heat structures per each 

sector (central and azimuthal) are used to distinguish between uninstrumented and 

instrumented fuel rods in the bundle. The design of instrumented fuel rods differs slightly in 

that internal and external claddings are used along the heated section (thermocouples are 

located in between). The surface factors are defined to account for the actual number of fuel 

rod simulators in each sector. Other heat structures are used to model unheated parts of 

FRSB. Multiple junction component (c076), two pipe components (c077, c078), and single 

junction component (c079) are used to model the core bypass simulator. A part of pipe 

(c078) vertical section is heated. Pipe component (c070) with 33 control volumes models the 

combined simulator of the upper plenum and upper head. Upper plenum outlet annulus is 

represented by a total of 12 annulus components in three axial levels and 4 azimuthal 

sectors. Multiple junction components (not shown in this view) are used to simulate axial and 

azimuthal flows between the upper plenum outlet annular sectors. Single junction 

components (cn04; n=1, 4) are used to model the primary coolant hot leg inlets. 
25

Figure 15 R5/M33 reactor pressure vessel model 
26

The nodalization of PCS loops is illustrated in Figures 16 to 19. Pipe components (cn05; 

n=1, 4) each with 22 control volumes are used to model the hot legs. Two single junction 

components (cn09, cn14; n=1, 4) and the single volume component (cn10; n=1, 4) per loop 

are used to model the steam generator inlets. Steam generator primary side models consist 

of a hot header model, pipe components (cn15; n=1, 4), horizontal tube bundle simulator, 

pipe components (cn2r; n=1, 4; r=1, 9), and cold header model, pipe components (cn31; 

n=1, 4). Multiple junction components are used to model flow paths between the tube bundle 

and the headers. In the PSB-VVER test facility, the steam generator simulator consists of 34 

nearly horizontal coiled tubes. In R5/M3.3 facility T-H model, the individual tubes are grouped 

to have nine tube rows along the height of steam generator vessel. Two single junction 

components (cn32, cn34; n=1, 4) and the single volume component (cn33; n=1, 4) per loop 

are used to model steam generator outlets. 
27

Figure 16 R5/M33 primary coolant system model – Loop No. 1 
28

Figure 17 R5/M33 primary coolant system model – Loop No. 2 
29

Figure 18 R5/M33 primary coolant system model – Loop No. 3 
30

Figure 19 R5/M33 primary coolant system model – Loop No. 4 
31

Pipe components (cn35, n=1, 4) each with 64 control volumes are used to represent the loop 

seals. Reactor coolant pumps are modelled with pump components (cn36, n=1, 4). The intact 

loop cold legs are modelled using pipe components (cn40, n=1, 2, 4) with 3 control volumes, 

valve components (cn41, n=1, 2, 4), and pipe components (cn42, n=1, 2, 4) with 14 control 

volumes. In Test-5a, the break was simulated to occur at cold leg No. 3, so the nodalization 

of the broken loop cold leg is different and follows standard guidelines to model the double 

ended cold leg break. The actions of HPIS/LPIS are modelled using time dependent volume 

and time dependent junction components. The injection points are Test-5a specific. 

The nodalization of the pressurizer and the pressurizer surgeline models is illustrated in 

Figure 20. Only the loop No. 4 hot leg connected surgeline is modelled using single junction 

component (c506), pipe component (c507) with 5 control volumes, valve component (c508), 

and pipe component (c509) with 23 control volumes. An attempt was made to model the loop 

No. 2 hot leg connected surgeline as well. This concept had to be abandoned due to 

computational difficulties (long dead end pipe). The model of the common part of the 

pressurizer surgeline begins with the branch component (c510) and continues using the pipe 

component (c511) with 5 control volumes. The actual connection to the pressurizer is 

modelled by the single junction component (c514). The pipe component (c515) with 55 

control volumes is used to model the pressurizer vessel. The single junction component 

(c519), connected to the top of the pressurizer, and the time dependent volume component 

(c520) are used to adjust the desired steady state pressurizer pressure. Similarly, the time 

dependent junction (c529) and the time dependent volume (c530) components are used to 

adjust the desired steady state pressurizer coolant level. Time dependent junction 

components (c541 and c549), and time dependent volume components (c540 and c550) are 

used to achieve the desired steady state surgeline coolant temperature. In transient runs, the 

steady state control components are removed from the model. 
32

Figure 20 R5/M33 pressurizer model 
33

The nodalization of the accumulators and accumulator surgelines is illustrated in Figure 21. 

Hydro-accumulator (TH20B01) was isolated during the Test-5a and therefore is not modelled 

here. Two hydro accumulators (c610 and c630) are connected to the upper plenum; one 

hydro-accumulator (c640) is connected to the downcomer. Only the check valves (c612, 

c632, and c642) are modelled explicitly. The respective surgelines isolation valves are not 

part of the model. The model also accounts for the actual locations of Test-5a specific 

HPIS/LPIS injection. 

Figure 21 R5/M33 emergency core cooling system model 

The nodalization of SG secondary side and associated steam and feedwater lines is 

illustrated in Figures 22 to 25. The combination of branch components is used to model the 

steam generator secondary side. Considering the top view, the central part of the vessel 

where the coiled tube bundle is located is treated as the “boiler” section, whereas the 

peripheral annular part is treated as the “downcomer” section. The lateral cross flow paths 

are modelled to simulate the internal recirculation of the secondary coolant. Nine rows of 

branch components correspond to nine rows of tube bundle simulated on the steam 

generator primary side. The branch components (cn71; n=1, 4) are used to account for the 

steam-water separation. The upper parts of steam generator vessels are modelled using the 

pipe components (cn75; n=1, 4), each with 3 control volumes. In PSB-VVER facility, there 

are two main steam lines per one steam generator. The 1.5 MWt steam line is designed to 

remove the heat generated by the 1.5 MWt FRSB power. If 10 MWt FRSB power (to follow 

1:300 scale) were used for the particular test, the 10 MWt steam line would have been used. 

During the PSB-VVER experiments carried out within the frame of OECD/NEA PSB-VVER 

project only the 1.5 MWt FRSB power systems had been available. However, in order to 

account for unisolatable parts of 10 MWt steam lines and also for future use of the model, the 

10 MWt steam lines are modelled and discussed here. 
34

The 1.5 MWt steam line models consist of single junction components (cn83; n=1, 4), pipe 

components (cn84; n=1, 4), valve components (cn85; n=1, 4), pipe components (cn86; 

n=1, 4), single junction components (cn87; n=1, 4), and pipe components (cn88; n=1, 4). 

Actually, the later components are parts of 10 MWt steam lines. 

The 10 MWt steam line models consist of single junction components (cn93; n=1, 4), pipe 

components (cn92; n=1, 4), valve components (cn91; n=1, 4), pipe components (cn90; 

n=1, 4), and valve components (cn89; n=1, 4). Valves in the 10 MWt steam lines were closed 

throughout the Test-5a. 

The lengths and therefore the nodalization of individual steam lines vary accordingly to the 

loop specific design of the PSB-VVER test facility. 

The combinations of time dependent volume components (cn95; n=1, 4), time dependent 

junction components (cn96; n=1, 4), pipe components (cn97; n=1, 4), and single junction 

components (cn98; n=1, 4) are used to model individual steam generator feedwater lines. In 

this, the facility steam generator feedwater system model is highly simplified. 

The single junction components (cn94; n=1, 4) simulate the connections of individual steam 

lines to the main steam line. 

Valve components (cn81; n=1, 4) are used to simulate actions of power operated relieve 

valves or safety valves. The test specific trip logic has to be set up to model steam dump to 

atmosphere. Time dependent volume components (cn82; n=1, 4) are used to model the 

boundary conditions. 
35

Figure 22 R5/M33 secondary coolant system model – Loop No. 1 
36

Figure 23 R5/M33 secondary coolant system model – Loop No. 2 
37

Figure 24 R5/M33 secondary coolant system model – Loop No. 3 
38

Figure 25 R5/M33 secondary coolant system model – Loop No. 4 
39

The nodalization of the main steam line is illustrated in Figure 26. The main steam line is 

simulated using two pipe components connected with the single junction component. Valve 

component (c957) models the steam dump valve RA06S01. 

Figure 26 R5/M33 main steam line model 

A number of heat structure components are used to simulate heat transfer in steam 

generators and heat losses from the pipe walls to the environment. 

There are more than 2900 control systems used in the model. They are set up to simulate 

I&C actions, to drive the model to the desired steady state, to facilitate the comparison of 

calculated data with measured data (e.g. differential pressures), etc. 

In this project, the developmental model control option 19 was activated, i.e. Bestion 

correlation was used for interphase friction in rod bundle geometry (pipe components c06n; 

n=1, 5). 

As for as the flow process models, the counter current flow limitation models (Wallis and 

Kutateladze) are applied at selected junctions. Multiple junction component (c068) is used to 

model flow paths at the upper core plate simulator. Wallis CCFL model is selected. In order 

to account for CCFL phenomena at the upper plenum simulator, Kutateladze CCFL model is 

activated at junction No. 14 of the pipe component c070. Wallis CCFL model is activated at 

junction No. 9 of the pipe component c030 that simulates the downcomer. Kutateladze CCFL 

model is also invoked for SG inlet junctions (cn09; n=1, 4). 
40

3.3 RELAP5/MOD3.3 PSB-VVER Test Facility Model Validation 

The third step in the thermal-hydraulic system code assessment process is defined in [2] as: 

 Perform assessment calculations. Compare the test data with the results of calculations 

for the key parameters. 

This study follows the concept of the OECD/NEA BEMUSE Project, specifically Phase II [13]. 

In that, the reference case was run in two steps, i.e. the steady state run and the transient 

run to enable code data evaluation on the steady state level and on the transient level. 

3.3.1 Reference Case 

Similarly as in Phase II of the BEMUSE project, reference case in PSB-VVER Test 5a post test 

analysis was run and the test data were compared with the results of the calculation in two 

steps, on the steady state level, and on the transient level. A 500-second “null transient” code 

calculation was used to simulate the test specific initial conditions. Code control blocks were set 

up to force the code to reach predefined primary system coolant mass flows and secondary 

system steam line pressures. A series of preliminary code runs was needed to adjust energy 

loss coefficients at selected junctions. A total of 119 measured and calculated parameters 

characterizing facility initial conditions were compared. In close to 80% of cases, the calculated 

value at 500 seconds was within the measurement accuracy interval <-2; +2>, where  is 

the measured value of the parameter and 2 is the measurement accuracy of the parameter as 

reported in [4]. Table 6 illustrates this steady state data evaluation process in the case of 

primary system loop No. 1. Please refer to Table 2 and Figure 8 for the measurement channel 

identification. Please note that pressures (P) are in MPa, coolant temperatures (T) in °C, and 

differential pressures (DP) in kPa. The “FALSE” cases were further evaluated as follows: 

 In the model, hydraulic characteristics of PCS loops are identical. For example, the 

differential pressure over steam generator (YA01DP03) is false in Table 6. However, the 

corresponding parameter evaluation for other loops is true. In the case of differential 

pressure over loop seal descending leg (YA0nDP05), loop # 3 and loop #4 parameter 

evaluations are true and loop #1 and loop #2 are false. 

 A check is made for two other PSB-VVER tests modelled with RELAP5/MOD3.3. The 

“FALSE” cases are ignored in this study if two out of three cases available are true. 

Normally, the request for additional information would be necessary. 

On the transient level, the test data were compared with the results of the calculation using 

FFTBM [7]. The selection of 23 parameters to be compared closely follows earlier studies [13] 

and [15]. For each parameter, the FFTBM algorithm is run in time window 0 – 1477 s (whole 

transient) to get average amplitudes (AA) and weighted frequencies (WF). These results are 

further combined using corresponding weighting factors to obtain total average amplitude (AAtot) 

and total weighted frequency (WFtot). In [7], the authors discuss the accuracy of code 

predictions as follows: 

 AAtot  0.3 characterize very good code predictions 

 0.3  AAtot  0.5 characterize good code predictions 
41

 The acceptability factor K is then defined for total average amplitude K = 0.4, and for 

primary system pressure K = 0.1. The acceptable code predictions are those with 

AAtot  0.4, and AAPSP  0.1. 

Table 6 R5/M33 Test-5a reference case steady state data evaluation – PCS #1 

Experiment Reference case 

(steady state) Measurement 

accuracy (2) 

Is cntrlvar value 

in the interval 

Channel ID <-2; +2>? Value 

(μ) Cntrlvar Value 

YA01P01 15.74 106 15.62 0.22 TRUE 

YA01P03 15.81 104 15.61 0.23 TRUE 

YA01P08 15.76 105 15.63 0.06 FALSE 

YA01P09 15.67 103 15.60 0.06 FALSE 

YA01T02 292 114 291.9 3 TRUE 

YA01T03 321 108 321.8 3 TRUE 

YA01T26 294 111 291.9 3.7 TRUE 

YA01T32 293 110 291.9 2 TRUE 

YA01T33 293 112 291.9 2 TRUE 

YA01DP01 1.9 115 1.8 0.5 TRUE 

YA01DP02 8.7 116 8.7 0.5 TRUE 

YA01DP03b 4.2 117 3.5 0.5 FALSE 

YA01DP04 -22.4 118 -21.8 1.5 TRUE 

YA01DP05 -23.1 119 -24.1 0.7 FALSE 

YA01DP08 4.5 121 4.5 0.3 TRUE 

YA01DP09b -5.9 122 -9.9 0.7 FALSE 

YA01DP10 0.6 123 0.5 0.5 TRUE 

YA01DP11 -0.9 124 -1.0 0.7 TRUE 

YA01DP13 -20.5 125 -20.5 0.7 TRUE 

YA01DP14 -19.6 126 -19.8 0.7 TRUE 

The results of FFTBM application on the reference case are summarized in Table 7. Since 

AAtot < 0.4 and AA1 = AA2 =0.1, the reference case code predictions are acceptable and can be 

followed by the uncertainty and sensitivity studies. 
42

Table 7 Test-5a R5/M33 reference case FFTBM results 

N Parameter ID Exp ID Cal AA 

1 Upper plenum pressure YC01P16 cntrlvar-11 0.10 

2 Pressurizer pressure YP01P01 cntrlvar-7 0.10 

3 Accumulator #1 pressure TH01P01 cntrlvar-297 0.11 

4 Accumulator #4 pressure TH04P01 cntrlvar-300 0.12 

5 Steam generator #1 secondary side 

pressure 

YB01P01 cntrlvar-511 0.18 

6 Steam generator #2 secondary side 

pressure 

YB02P01 cntrlvar-521 0.18 

7 Core inlet coolant temperature YC01T06 tempf04101 0.12 

8 Core outlet coolant temperature YC01T04b tempf07009 0.51 

9 HPIS flow TJ04F01 mflowj-681 0.40 

10 Integrated break flow MBr cntrlvar-702 0.11 

11 Primary coolant mass M1k cntrlvar-580 0.66 

12 Accumulator #1 level TH01L01 cntrlvar-197 0.03 

13 Accumulator #4 level TH04L01 cntrlvar-198 0.04 

14 Differential pressure across downcomer YC01DP03 cntrlvar-16 0.97 

15 Differential pressure across downcomer YC01DP04 cntrlvar-17 0.98 

16 Differential pressure across the 

downcomer outlet and core simulator inlet YC01DP06 cntrlvar-19 0.99 

17 Differential pressure across the core DP_CORE cntrlvar-657 1.22 

18 Differential pressure across the upper 

plenum YC01DP13 cntrlvar-26 1.30 

19 Differential pressure across the upper 

plenum outlet and downcomer inlet YC01DP16 cntrlvar-29 0.77 

20 Peak cladding temperature 

max 

(YC01T09 - 

YC01T123) 

cntrlvar-656 0.46 

21 Cladding temperature – core bottom YC01T113 httemp65300320 0.70 

22 Cladding temperature – core center YC01T55 httemp65301020 0.53 

23 Cladding temperature – core top YC01T39 httemp65301220 0.51 

 Total average amplitude - AAtot - - 0.37 
43

Appendix A contains the figures with graphical comparisons of measured and RELAP5/MOD3.3 

calculated results. 

Figure A.1 compares the data for integrated break mass flows. In the experiment, this 

parameter was not measured directly, but a specific algorithm for measured data evaluation was 

applied. The Henry-Fauske critical flow model was applied in calculations with Cd = 0.83 in the 

reference case. 

Figure A.2 and Figure A.3 show good agreement between measured and calculated primary 

system pressures. Consequently, the timing of the accumulator injection is well predicted in the 

model (see Figures A.4 – A.7). 

Actions of HPIS and LPIS were simulated during the experiment. Coolant was injected to the 

upper plenum simulator, to the downcomer simulator, to the intact loop (No. 1) hot leg, and to 

the broken loop (No. 3) hot and cold legs. As an example, Figure A.8 compares the data for one 

injection line. In the R5/M33 model, all five injection lines liquid mass flows were modelled as 

boundary conditions to match integrated HPIS/LPIS mass flow in the experiment. 

Figure A.9 shows the comparison of the primary system coolant mass between evaluated in 

experiment and R5/M33 results. The experimental data is the product of the specific algorithm 

using data measured directly. According the experimenters the accuracy of this evaluation is 

± 50 kg. 

Figures A.10 – A.13 show the comparison of the secondary system pressure (at the steam 

generator steam dome) between measured and R5/M33 results. With reverse heat transfer 

across the tube bundle, the calculated data agrees well with the measurement up to 

approximately 600 s. 

Figures A.14 – A.15 show the comparison of the core simulator inlet and outlet coolant 

temperatures between measured and R5/M33 results. Perhaps the most challenging 

phenomena to reproduce by the code are those related to the upper plenum ECCS injection 

(from two accumulators and one line of HPIS/LPIS). The measured data show that subcooled 

liquid reached the core simulator outlet region up to approximately 120 s. In the attempt to 

capture the phenomena, the CCFL models were activated at two axial levels of the RPV model. 

The Kutateladze CCFL model was activated at the junction just below the injection points. Wallis 

CCFL models were activated at the junctions representing the flow through the upper core plate 

simulator. 

Figures A.16 – A.18 show the comparison of differential pressures across the downcomer and 

lower plenum simulators between measured and R5/M33 results. With a direct downcomer 

simulator ECCS injection it was difficult again to reproduce the measured values more correctly. 

The Wallis CCFL model was activated at the junction just below the injection point. 

Figures A.19 – A.20 show the comparison of differential pressures across the core and upper 

plenum & upper head simulators between measured and R5/M33 results. Qualitatively, the 

liquid phase distribution is predicted correctly during the accumulator injection period only for 

the upper plenum & upper head simulator. 

Figure A.21 compares the data for pressure drop across the downcomer and the upper plenum 

simulators. 

Two periods of FRSB cladding heatup were observed during the experiment. First one begun 

approximately 50 s after the stop of accumulator injection when FRSB liquid level decreased. As 
44

shown in Figure A.22, in contrast with experimental data, the code predicts the first heatup to 

occur also in the lower third region of FRSB. As shown in Figures A.23 – A.25, the timing of the 

first FRSB cladding heatup and quench is predicted quite correctly for the two other cladding 

temperatures selected for comparison. Please note that measurement channel selection was 

based on [15]. During the experiment, the first FRSB cladding heatup was terminated by the 

continuing HPIS/LPIS injections which gradually lead to the primary system coolant mass partial 

recovery. The second FRSB cladding heatup occurred after the stop of HPIS/LPIS injection 

(operator action) when primary coolant mass inventory substantially decreased again as a result 

of continuing boil-off. The transient was terminated by the facility operator when peak cladding 

temperature reached the predetermined value. During the experiment, the second heatup was 

not recorded in the bottom third FRSB region and, this is correctly predicted by the R5/M33 

model. As for R5/M33 model predictions of FRSB cladding temperatures in mid and top FRSB 

regions, the start of the second heatup is delayed by approximately 80 s. 

3.3.2 Uncertainty Study 

In the performed uncertainty study, the fully probabilistic methods with propagation of 

uncertainties of code input parameters are applied. 

For the probabilistic methods, two types of uncertainties can be distinguished: 

 The uncertainties of the first type are directly modelled by uncertain input parameters 

(e.g. physical models). 

 The uncertainties of the second type are not modelled via input parameters. They are 

uncertainties due to the nodalization, treatment of (relative) deficiencies of the code, 

scaling effect, and user effect. 

Three different main steps are used to establish input uncertain parameters: 

 Identification of the main phenomena. 

 Identification of the sensitive parameters for these phenomena. 

 Establishment of the list of the parameters taken into account. 

This method is performed by expert judgment. The aim is to consider all the potentially 

influential parameters. As far as possible, the specificities of PSB-VVER are taken into account. 

Previous studies have been considered [17], [18]. 

3.3.2.1 Uncertain Input Parameters 

The uncertainty of an input variable (geometrical data, physical model constant, etc.) can be 

modelled by a different method. The uncertainty of an input variable can be described directly 

by its pdf. Such an uncertain input variable can have dimensions. If the dimensionless 

uncertainty parameter of multiplicative form is used (the nominal value of the parameter 

being 1), that is the multiplier. 

The following sources of uncertainties are considered: 

 Physical models. These uncertainties belong to the used code (RELAP5/MOD3.3). 
45

 Initial and boundary conditions. Uncertainties belong to the experimental equipment 

(PSB-VVER) and the modelled test Test-5a. 

 Geometric modelling of experimental equipment by means of the RELAP5/MOD3.3 

code. It is fixed to reference the case that respects the RELAP5/MOD3.3 user’s 

guidelines in the nodalization of the PSB-VVER equipment and makes the best account 

of this kind of solved process. Nodalization is performed by respecting the rules of the 

user’s guidelines manual code [19], [20]. 

 3-D flow modelling is applied by means of pressure vessel modelling in input 

RELAP5/MOD3.3 dates. 

Uncertainties of the scaling effect and user’s effect are not considered. The scaling effect is not 

serious because the evaluated Test-5a was realized on the modelled experimental equipment 

(PSB-VVER). 

3.3.2.1.1 Probability Distribution Function 

The range of variation of an input uncertain parameter is specified by means of the probability 

distribution function (pdf). One of the following statistical distributions is used: 

 Normal distribution (N) 

 Lognormal distribution (LN) 

 Uniform distribution (U) 

 Discrete distribution (D) 

the indicated ranges of variation for parameters with normal and lognormal distribution 

correspond to 2.5% and 97.5% percentiles, i.e.   1.96  and e ± 1.96

, respectively. These 

percentiles are given to specify the probability density functions. 

Discrete distribution is the case of alternative physical models (choice among two, three or more 

correlations to describe the same physical phenomenon). 

A list of 31 considered uncertain input parameters is introduced in Table 8. 
46

Table 8 List and description of uncertain input parameters 

phenomena or 

main physical 

laws 

parameter

number

parameter description 

type of pdf imposed range 

of variation; 

uncertainty

reference; 

comments; 

uncertain method 

flow rates 

repartition in 

the 

circuit/pressure 

drops 

1 form loss coefficient 

(active core) LN cf=0.1 

(0.05; 0.2) 

preliminary study;

expert judgement 

flow rate at the 

break 

2 

m break discharge coefficient U Cd = 0.83 

(0.664; 0.996) 

preliminary study;

expert judgement 

3 

m 

thermal nonequilibrium 

constant D 

discrete 

parameters: 

control weight 

0.14 0.60; 

0.0 0.40 

[19] A9.17, p.209; 

preliminary study;

expert judgement 

4 

m 

liquid entrainment alternative 

models D 

discrete 

parameters: 

control weight 

0 0.25; 

1 0.25; 

2 0.25; 

3 0.25 

[19], A9.6, p.79; 

preliminary study;

expert judgement 

fuel thermal 

behaviour 

5 initial core power N 1512 ±15 kW 

(1497; 1527) 

[4], Table B.1; 

YC01N01 

6 multiplier 

cladding thermal conductivity 

N c = 1.0 

(0.9; 1.1) [21], Table A.7 

7 multiplier 

cladding specific heat capacity N c = 1.0 (0.9; 1.1) [21], Table A.8 

8 multiplier 

MgO thermal conductivity N c = 1.0 (0.9; 1.1) 

[1], Fig. 2.3; 

preliminary study;

expert judgement 

9 multiplier 

MgO specific heat capacity N c = 1.0 (0.9; 1.1) 

[1], Table 2.3; 

preliminary study;

expert judgement 

pump 

behaviour 

10 

m 

multiplier 

pump two-phase head 

difference multiplier 

U 1 ± 0.2 

(0.8; 1.2) 

[9], Vol. III, 2.3.2, 

p. 118; 

expert judgement 
47

phenomena or 

main physical 

laws 

parameter

number

parameter description 

type of pdf imposed range 

of variation; 

uncertainty

reference; 

comments; 

uncertain method 

heat transfers 

in the rewetted 

zone 

11 

m 

multiplier 

heat transfer fouling factor 

(± 18%) 

N 1 ± 18% 

(0.82; 1.18) 

[20], Vol. IV, 

4.2.3.4.1, p. 133; 

expert judgement 

critical heat flux 12 

m 

multiplier 

CHF correlation (PG-p) N gspf = 1.0 

(0.72; 1.28) 

[20], Vol. IV, 4.3, 

p. 192-202 

primary circuit 

thermal losses 13 heat transfer coefficient on outside surfaces of piping LN =3.5 Wm-2K-1

(1.75; 7) 

preliminary study;

expert judgement 

counter current 

flow limitation 

(CCFL) 

14 

m 

upper core plate: 

c of Wallis correlation 

(m = 1) 

U c=2.1 ± 0.1 

(2.0; 2.2) 

preliminary study;

expert judgement 

15 

m 

upper plenum: 

c of Kutateladze correlation 

(m = 1) 

U c=0.8± 0.2 

(0.6; 1) 

preliminary study;

expert judgement 

16 

m 

steam generator inlet: 

c of Kutateladze correlation 

(m = 1) 

U c=1.0 ± 0.2 

(0.8; 1.2) 

preliminary study;

expert judgement 

17 

m 

downcomer: 

c of Wallis correlation 

(m = 1) 

U c=1.0 ± 0.2 

(0.8; 1.2) 

preliminary study;

expert judgement 

accumulator 

data 

18 accumulator initial liquid level N 0  0.07 m 

(-0.07; 0.07) 

[4], Table B.1 

TH0nL01n=1,3,4 

19 friction form loss in the 

accumulator line acc

LN acc = 30 

(15; 60) 

preliminary study;

expert judgement 

20 accumulator initial liquid 

temperature N 0 ± 3.4 °C (-3.4; 3.4) 

[4], Table B.1 

TH0nT01n=1,3,4 

21 accumulator initial liquid 

pressure 

N 0  0.03 MPa 

(-0.03; 0.03) 

[4], Table B.1 

TH0nP01n=1,3,4 

pressurizer 

data 

22 initial level N 6.5 ± 0.5 m 

(6; 7) 

[4], Table B.1 

YP01L02 

23 initial pressure N 

15.58  0.06 

MPa 

(15.52; 15.64) 

[4], Table B.1 

YP01P01 

24 friction form loss in the surge 

line LN press = 3.9 

(1.95; 7.8) 

preliminary study;

expert judgement 

initial 

conditions in 

systems 

25 initial intact loop mass flow 

rate N ± 0.1 kg/s (-0.1; 0.1) 

[4], Table B.1 

YA0nF01b 

n=1,4 
48

phenomena or 

main physical 

laws 

parameter

number

parameter description 

type of pdf imposed range 

of variation; 

uncertainty

reference; 

comments; 

uncertain method 

26 initial pressure on 

parogenerator secondary side N 7.6 ± 0.2 MPa (7.4; 7.8) 

[4], Table B.1 

YB01P01 

expert judgement 

27 initial upper-head mean 

temperature N 304 ± 3 °C (301; 307) 

[4], Table B.1 

YC01T05 

data related to 

injections 

(HPSI/LPSI) 

28 liquid injection variation N  0.004 kg/s [4], Table B.1 

TJ01F01 

29 liquid injections temperature N 50  5 °C 

(45; 55) 

[4], Figure A.10 

RL09T01 

containment 

pressure 

30 multiplier 

cold leg discharge pressure N c=1.0 ± 0.02 (0.98; 1.02) 

[4], Table B.1; 

YE11P02 

31 multiplier 

hot leg discharge pressure N c=1.0 ± 0,02 (0.98; 1.02) 

[4], Table B.1; 

YE12P02 

3.3.2.1.2 Physical Model Uncertainties 

Uncertainties of the RELAP5/MOD3.3 physical models are evaluated on the basis of the code 

documentation [19], [20] and comparison of code calculations results with experimental data 

[16], [17]. 

In Table 8 the number of parameter describing an uncertainty of the physical model is followed 

by the lower case letter m. 

Physical model uncertainties are described by means of the probability distribution functions 

(pdf) for 10 uncertain parameters. Two of them, the thermal nonequilibrium constant and the 

liquid entrainment (parameter 2 and parameter 3, respectively), are described by alternative 

models, i.e. discrete distributions are used. Normal distributions are used for multipliers of the 

heat transfer fouling factor and the CHF correlation grid spacer factor (parameter 11 and 

parameter 12, respectively). 

Uniform distributions are used for the break discharge coefficient and for the pump two-phase 

head multiplier (parameter 2 and parameter 10, respectively) and for constants used in 4 

uncertain parameters of CCFL models (parameters 14 – 17). 

3.3.2.1.3 Uncertainties of Initial and Boundary Conditions 

These uncertainties are specified with help of PSB-VVER descriptions [1] and Test-5a 

experimental data report [4]. 21 uncertain parameters involve: 

 local pressure drop of active core (parameter 1) and primary circuit thermal losses 

(parameter 13), 

 fuel thermal behaviour (parameters 5-9), 
49

 accumulator data (parameters 18-21), 

 pressurizer data (parameters 22-24), 

 initial conditions in systems (parameters 25-27), 

 data related to active ECCS (HPSI/LPSI) (parameters 28, 29), 

 containment pressure (parameters 30, 31). 

3.3.2.2 PSB-VVER Test-5a Experimental Data 

PSB-VVER Test-5a experiment [4] is briefly described in section 3.1 above. Key experimental 

data are compared with R5/M33 reference case results in section 3.3.1 above and in Appendix 

A. In the context of the uncertainty study, key experimental data are presented in integrated 

form in Figures B.1 and B.2. Experimental time dependent values of pressures in accumulator 

(ACC), pressurizer (PRZ), upper plenum (UPP), steam generator (SG), power of FRSB (N) and 

primary coolant mass (PCM) are shown in Figure B.1. Experimental time dependent values of 

the integrated break flow (IBF), the primary coolant mass (PCM) and the maximum cladding 

temperature (maxT) are shown in Figure B.2. 

3.3.2.3 Uncertainty Method and Uncertainty Analysis Results 

The propagation of uncertainties of code input parameters is evaluated as follows. 

Uncertain code output parameters are statistically evaluated as  = 95% tolerance limit with a 

confidence level  of 95% on the basis of code results obtained in a propagation step. The 

propagation step involved: n code runs performed by varying simultaneously the values of all 

the uncertain input parameters according to their pdf. The number n of performed code runs in 

order to obtain (95%, 95%) tolerance limits is given by Wilks’ formula [22]. The n values of the 

uncertain output parameter are ordered: Y(1) < Y(2) ....< Y(n-1) < Y(n). For Wilks’ formula at the 

first order is n = 59. 

In such a case the minimal value Y(1) is the lower (95%, 95%) tolerance limit (lower uncertainty 

bound) and the majoring value Y(59) is the upper (95%, 95%) tolerance limit (upper uncertainty 

bound). An uncertainty band of the code output parameter is a difference Y(59) – Y(1). 

Two kinds of uncertain output parameters are evaluated: scalar and time trends output 

parameters. 

3.3.2.3.1 Scalar Output Parameters 

Uncertainty bounds are estimated for 7 scalar output parameters: 

 Accumulator injection time tacc (s) 

 Cladding first overheating time t1 (s) 

 Maximum peak cladding temperature time tmax (s) 

 Complete core quenching time t2 (s) 

 Cladding first overheating time interval t2 - t1 (s) 

 Cladding last overheating time t3 (s) 

 Maximum peak cladding temperature maxT (°C) 
50

Lower and upper uncertainty bounds for 7 single-valued parameters are in Table 9. 

Statistical bounds envelope experimental values for 5 parameters, including the maximum peak 

cladding temperature that is safety important. Only the experimental maximum peak cladding 

temperature time tmax = 398.5 s is higher than the calculated value of an upper uncertainty 

bound 395 s and cladding last overheating time t3 = 1265.3 s is lower than the calculated value 

of an lower uncertainty bound 1268.5 s. 
51

Table 9 One-sided (0.95; 0.95) tolerance bounds for the single-valued parameters 

output uncertain parameter 

lower 

uncertainty 

bound 

reference 

calculation 

value 

experimental 

value 

upper 

uncertainty 

bound 

accumulator injection time

tacc (s) 10.3 11.1 10.7 12.6 

cladding first overheating 

time*)

t1 (s) 

108.2 155.5 181 238.5 

maximum peak cladding 

temperature time 

tmax (s) 

236.5 280 398.5 395 

complete core quenching 

time*)

t2 (s) 

332.7 521.5 556.7 673 

cladding first overheating 

time interval 

t2 - t1 (s) 

159.2 366 375.7 542.9 

cladding last overheating 

time*)

t3 (s) 

1268.5 1310.2 1265.3 1385.6 

maximum peak cladding 

temperature

maxT (°C) 

276.6 457.2 428.7 638.2 

*) Cladding overheating t1, t3 and quenching t2 times are specified in the moment when the 

maximum peak temperature of rods cladding crosses the value 200 °C.

3.3.2.3.2 Time Trends Output Parameters 

Time dependent uncertainty bands are estimated for 4 time trends output parameters: 

 the maximum cladding temperature (maxT) 

 the upper plenum pressure (UPP) 

 the primary coolant mass (PCM) 

 the integrated break flow (IBF) 

Time dependent uncertainty bands of 4 time trends output parameters are presented for the 

maximum cladding temperature (maxT) in Figure B.3, for the upper plenum pressure (UPP) in 

Figure B.4, for the primary coolant mass (PCM) in Figure B.5, and for the integrated break flow 

in Figure B.6. 

From the physical point of view it is convenient to evaluate the results of uncertainty analysis in 

two time intervals. 
52

Time interval up to circa 650 s: 

Some typical aspects of LBLOCA process are observed. Uncertainty analysis results as follows: 

 Time dependent band of the maximum cladding temperature (maxT) envelope 

experimental values (Figure B.3). 

 Time dependent band of the upper plenum pressure (UPP) envelope experimental 

values with exception the first 6 s of process (Figure B.4). 

Time interval over 650 s up to circa 1000 s: 

Time dependent values of the upper plenum pressure (lower than 0.25 MPa) predetermine 

process running. 

Uncertainty analysis results show that experimental values of the upper plenum pressure and 

the maximum cladding temperature exceed their upper uncertainty bounds. 

The reason is lower calculated values of the primary coolant mass (PCM) against experimental 

values (Figure B.5) that are also connected with the higher calculated values of the integrated 

break flow (IBF) against experimental values (Figure B.6). For both parameters, such 

disagreements last during the whole process. The original cause is very probably overestimated 

calculated values of the critical discharge from cold leg in the process beginning (up to circa 

36 s). 

3.3.3 Sensitivity Study 

The sensitivity analysis of calculated results is performed with help of the code 

STATISTICA [23]. 

In the analysis the Spearman’s rank correlation coefficient is applied both for the single-values 

parameters and the time trends output parameters. 

3.3.3.1 Single-value Output Parameters Sensitivity Results 

The determined sensitivity measures for the single-value output parameters are introduced in 

Table 10 and are shown in Appendix B in Figures B.7 - B.12. 

Values of the Spearman’s rank correlation coefficients, which are significant and can be counted 

as reliable, and the corresponding input uncertain parameter numbers are written in bold in 

Table 10. 
53

Table 10 Sensitivity analysis of the single valued output parameters – Spearman’s 

correlation coefficients 

Input 

uncertain 

parameter 

number 

Output uncertain parameter 

tacc

Accumulator 

injection 

time 

t1

Cladding 

first 

overheating 

time 

t2 – t1 

Cladding 

first 

overheating 

time 

interval 

t3

Cladding 

last 

overheating 

time 

tmax 

Maximum 

peak 

cladding 

temperature 

time 

maxT 

Maximum 

peak 

cladding 

temperature

1 0.081 0.188 -0.107 -0.008 0.010 -0.052 

2 -0.965 -0.125 0.075 -0.245 -0.173 0.137 

3 0.040 0.055 0.037 0.251 0.110 0.037 

4 -0.262 -0.109 0.082 -0.083 0.050 0.105 

5 0.228 -0.131 0.058 -0.071 -0.093 0.083 

6 -0.116 -0.193 0.129 0.035 -0.095 -0.019 

7 0.081 0.250 -0.260 0.148 0.111 -0.271 

8 0.003 0.273 -0.175 0.107 -0.123 -0.143 

9 0.208 0.014 -0.139 -0.032 -0.142 -0.122 

10 0.072 0.032 0.002 -0.074 -0.025 -0.059 

11 -0.034 -0.119 -0.008 0.281 -0.048 0.144 

12 0.181 0.075 -0.161 -0.037 -0.067 -0.026 

13 0.023 0.049 -0.208 0.124 0.054 -0.129 

14 0.103 0.023 0.216 -0.126 0.142 0.217 

15 -0.069 0.160 -0.482 0.011 -0.234 -0.494 

16 0.223 -0.212 0.185 0.342 -0.063 0.145 

17 -0.035 -0.114 0.145 -0.183 0.109 0.116 

18 -0.027 -0.131 0.219 0.098 0.059 0.303 

19 0.106 -0.006 0.200 -0.006 0.145 0.134 

20 -0.047 -0.065 -0.109 0.066 -0.201 0.013 

21 0.010 0.123 -0.060 -0.016 -0.152 -0.069 

22 0.146 -0.022 -0.016 -0.206 0.141 -0.045 

23 -0.170 0.016 -0.086 -0.026 -0.032 -0.020 

24 -0.055 -0.099 0.073 -0.044 0.111 0.078 

25 0.032 0.180 -0.169 -0.063 -0.054 -0.116 

26 0.195 0.253 -0.238 0.172 -0.136 -0.165 

27 -0.018 -0.051 0.072 -0.037 -0.081 0.041 

28 -0.040 0.068 -0.008 0.057 -0.023 -0.061 

29 0.093 0.071 -0.013 -0.103 0.162 0.015 

30 -0.020 -0.155 -0.043 -0.035 -0.320 0.049 

31 0.097 -0.007 -0.076 0.033 0.129 -0.054 
54

3.3.3.2 Time Trends Output Parameters Sensitivity Results 

The sensitivity analyses of calculated time trends of the output parameters are performed for 

4 important measured or evaluated parameters during Test 5a. They are: 

 the maximum cladding temperature (maxT), 

 the upper plenum pressure (UPP), 

 the primary coolant mass (PCM), 

 the integrated break flow (IBF). 

The sensitivity analyses of 4 time trends output parameters are carried out with time steps. 

Values of the Spearman’s rank correlation coefficients considering all 31 uncertain input 

parameters are shown in figures of Appendix B: 

 Figures B.13 – B.20 for the maximum cladding temperature (maxT), 

 Figures B.21 – B.28 for the upper plenum pressure (UPP), 

 Figures B.29 – B.36 for the primary coolant mass (PCM), 

 Figures B.37 – B.44 for the integrated break flow (IBF). 

3.3.3.3 Sensitivity Code Results for Uncertain Input Parameters 

On the basis of the Spearman’s rank correlation coefficients shown in Figures B.13 – B.44 the 

uncertainties of 17 input uncertain parameters are evaluated as significant at least for one from 

time trends output parameters maxT, UPP, PCM and IBF. Time trends of the Spearman’s 

correlation coefficients of output parameters maxT, UPP, PCM and IBF for each from these 17 

significant input parameters are shown in Figures B.45 – B.61. The list of 17 input uncertain 

parameters and corresponding figures numbers follows in Table 11. 
55

Table 11 List of significant input uncertain parameters for the time trends output 

parameters maxT, UPP, PCM and IBF 

Parameter number Uncertain input parameter Figure No. 

2 

m 

break discharge coefficient Figure B.45 

3 

m 

thermal nonequilibrium constant Figure B.46 

5 initial core power Figure B.47 

7 cladding specific heat capacity Figure B.48 

10 

m 

pump two-phase head difference multiplier Figure B.49 

13 heat transfer coefficient on outside surfaces of piping Figure B.50 

14 

m 

upper core plate: 

c of Wallis correlation Figure B.51 

15 

m 

upper plenum: 

c of Kutateladze correlation Figure B.52 

16 

m 

steam generator inlet: 

c of Kutateladze correlation Figure B.53 

17 

m 

downcomer: 

c of Wallis correlation Figure B.54 

18 accumulator initial liquid level Figure B.55 

19 friction form loss in the accumulator line acc Figure B.56 

22 initial level in pressurizer Figure B.57 

26 initial pressure on steam generator secondary side Figure B.58 

28 liquid injection variation Figure B.59 

29 liquid injections temperature (HPSI/LPSI) Figure B.60 

30 cold leg discharge pressure Figure B.61 
56

3.4 Conclusions on RELAP5/MOD3.3 Code Assessment 

Section 3 of this report contains results obtained in the assessment of RELAP5/MOD3.3 code 

against the PSB-VVER Test-5a experiment. In addition to the traditional structure of thermalhydraulic code assessment reports, i.e. facility description, test description, thermal-hydraulic 

code facility model description, and the thermal-hydraulic code facility model verification and 

validation, this report also presents the results of the uncertainty and sensitivity study. 

Methodically, the authors repeated the contribution of NRI-1 team in recent BEMUSE project [3], 

[13], and [16]. 

As for as Test-5a LB LOCA experiment carried out at PSB-VVER test facility, it is important to 

note that Test-5a initial FRSB power was scaled down to 15% of prototypical. This was the 

reason that the measured cladding temperatures showed qualitatively different trends when 

compared to typical PWR LB LOCA. 

In the RELAP5/MOD3.3 PSB-VVER model validation step, it is shown that in the reference case 

calculation the results meet the acceptance criteria when compared to data on the steady state 

level and on the transient level. In the latter case, the application of FFTBM [7] showed that total 

average amplitude for 23 parameters selected for quantitative comparison was AAtot = 0.37, 

with average amplitudes for primary system pressure equal to 0.1, which enabled the 

conclusion that the model reproduced the experimental data with good accuracy. 

Non-parametric statistics with application of Wilks formula of first order was used in the 

uncertainty study. This method requires carrying out at least 59 code runs with random values 

of uncertain input parameters in order to evaluate one-sided lower and upper (0.95; 0.95) 

tolerance bounds for output parameters (peak cladding temperatures and other measured data). 

In the uncertainty study, 31 uncertainty input parameters were considered. Uncertainty bands 

were evaluated for selected scalar parameters as shown in Table 9 and for selected vector 

parameters (time trends) as shown in Figures B.3 – B.6. Uncertainty bounds envelop the 

experimental data for the start of accumulator injection; for the start of first FRSB cladding 

heatup, and for the duration of first FRSB cladding heatup. Uncertainty bounds do not envelop 

the experimental data for timing of peak cladding temperature during the first FRSB cladding 

heatup (upper uncertainty bound is by 3.5 s lower than in experiment), and for timing of final 

FRSB cladding heatup (lower uncertainty bound is by 3.2 s higher than in experiment). 

Calculated lower and upper (0.95; 0.95) uncertainty bounds for peak cladding temperature 

during first FRSB cladding heatup envelop the measured value. This is the key finding when 

considering the licensing point of view. 

From the time trends uncertainty results, the maximum cladding temperature uncertainty band 

as the most important result from the licensing point of view is discussed. Uncertainty band 

defined by one-sided lower and upper (0.95; 0.95) uncertainty bounds envelops experimental 

data up to approximately 700 s. After that, up to approximately 1000 s, with FRSB quenched, 

maximum cladding temperatures follow the saturation temperature and the uncertainty band lies 

slightly below the measured data. During that period, the calculated primary system pressure is 

lower than the measured one. In time interval approximately from 1200 s to 1300 s, the upper 

uncertainty bound is slightly lower than measured values of maximum cladding temperature due 

to calculated timing of final FRSB cladding heatup. 

The sensitivity study was carried out on 60 code runs with random values of uncertain input 

parameters. Sensitivities of code output parameters on code physical models uncertainty 
57

parameters and on uncertainties of initial and boundary conditions were evaluated. Based on 

Spearman’s rank correlation coefficient, 17 out of 31 uncertain input parameters were evaluated 

as influential. 

59

4. TRACE V5.0 CODE ASSESSMENT 

As for small-break LOCA analyses, the TRACE V5.0 code developmental assessment [24] is 

based on LOFT, ROSA IV, Semiscale, and BETHSY small-break LOCA tests. The results of the 

independent TRACE V5.0 code assessment against small-break LOCA test performed at PSBVVER test facility [5] are reported below. 

First two steps in the thermal-hydraulic system code assessment process are defined in [2] as: 

 Formulate assessment matrices for each class of transient studied, e.g., large-break 

LOCA, small-break LOCA. 

 Select key parameters for these classes. 

When addressing these steps the authors please the reader to refer to [14]: “The construction of 

VVER Thermal Hydraulic Code Validation Matrix follows the logic of the CSNI Code Validation 

Matrices (CCVM) [12]. Similar to the CCVM it is an attempt to collect together in a systematic 

way the best sets of available test data for VVER specific code validation, assessment and 

improvement, including quantitative assessment of uncertainties in the modelling of phenomena 

by the codes.” Since the PSB-VVER test facility had not been in operation at the time the report 

[14] was compiled, the phenomena vs. system test large-break LOCA matrix refers PSB-VVER 

as “expected to be suitable”. Test facility vs. phenomena, specifically the suitability of test data 

for the code assessment is further discussed in [5]. The selection of key parameters to be 

compared is based on [13] and [15]. 

4.1 PSB-VVER Test-3 Description 

The general objectives of PSB-VVER small-break LOCA experiment are listed in [5]: 

 Investigate the thermal-hydraulic response of the VVER-1000 primary system to the 

small break LOCA. 

 Obtain experimental data for validation of thermal-hydraulic codes applied to SB LOCA 

analysis of VVER-1000 NPP. 

 Comparison of the experimental data obtained on PSB-VVER with the data previously 

obtained on LOBI, BETHSY, SPES and LSTF facilities, for the purpose of determining 

the scaling effect on the accident behaviour. Hence, the test scenario closely followed 

that of BL-34 test performed on LOBI test facility [25]. 

Internally at EREC, the test is denoted as CL-4.1-03. In the context of PSB-VVER OECD 

Project the test is denoted as Test-3. 

Test-3 specific PSB-VVER facility configuration is discussed in Table 12. 
60

Table 12 Test-3 specific PSB-VVER test facility configuration 

Subsystem Description 

Pressurizer Surgeline was connected to the hot leg loop No. 4 

Core bypass Two orifices with a diameter of 7 mm were installed at the 

inlet/outlet of the core bypass 

Steam generators Under initial steady state operation all SGs were connected to each 

other by the steam header. The steady state pressure was adjusted 

by steam dump valve RA06S01 located at common steam line. 

Feedwater Under initial steady state operation the SG secondary side coolant 

levels were adjusted by RL01S06, RL02S06, RL03S06, and 

RL04S06 valves. The feedwater temperature was close to 170 °C. 

Steam dump to 

atmosphere (ADS) 

ADS line was connected to each steam generator. A throttle 

channel with l/d = 10; d = 12.1 mm was installed in each ADS line. 

Accumulators ACC No. 2 and ACC No. 4 surgelines were connected to the 

downcomer simulator. 

HPIS N/A. 

LPIS Three lines of LPIS were connected to cold legs of PCS loops 

No. 1, No. 3, and No. 4. 

Break unit Vertically upward oriented break unit was located in the horizontal 

section of PCS loop No. 4 cold leg between downcomer simulator 

inlet nozzle and RCP. Break nozzle was smooth edged channel (R 

= 6 mm) with a diameter of 10 mm and a length of 100 mm. 

Upper plenum Under steady state operation the upper plenum simulator warm-up 

line was opened. The line was closed 2 minutes prior the initiation 

of the transient. 

The PSB-VVER DAS was recording the measured parameters with frequency of 5 Hz 

throughout the test. 

Test-3 specific PSB-VVER facility initial conditions are presented in Table 13. 
61

Table 13 Test-3 specific PSB-VVER test facility initial conditions 

Parameter Measurement Value Measurement 

accuracy 

Primary coolant system 

Upper plenum pressure, MPa YC01P17 15.71 ± 0.06 

Hot leg inlet coolant temperature, °C YA01T03 310 ± 3 

YA02T03 308 ± 3 

YA03T03 311 ± 3 

YA04T03 308 ± 3 

Cold leg outlet coolant temperature, 

°C 

YA01T02 283 ± 3 

YA02T02 283 ± 3 

YA03T02 282 ± 3 

YA04T02 282 ± 3 

FRSP power, kW YC01N01 1129 ± 15 

BP power, kW YC01N02 14.9 ± 0.7 

Pressurizer level, m YP01L02 3.05 ± 0.3 

Secondary coolant system 

SG pressure, Mpa YB01P01 6.88 ± 0.05 

YB02P01 6.91 ± 0.05 

YB03P01 6.93 ± 0.05 

YB04P01 6.88 ± 0.05 

SG level, m YB01L01 1.90 ± 0.08 

YB02L01 1.91 ± 0.08 

YB03L01 1.94 ± 0.08 

YB04L01 1.90 ± 0.08 

Accumulators 

ACC pressure, MPa TH02P01 4.08 ± 0.03 

TH04P01 4.14 ± 0.03 

ACC level, m TH02L01 4.58 ± 0.07 

TH04L01 4.60 ± 0.07 
62

Test-3 specific PSB-VVER facility boundary conditions are presented in Table 14. 

Table 14 Test-3 specific PSB-VVER test facility boundary conditions 

Time, s Event 

0 Break simulation 

4.1 Primary system pressure drops to 13 MPa: SCRAM 

4.1 SCRAM, FRSB and BP power controlled to simulate decay heat 

4.1 SG feedwater isolation, start of RL01-04S06 valve closing 

4.1 Main stem line isolation, start of RA06S01 valve closing 

4.1 RCPs trip, coastdown was not simulated 

8.0 RCPs rotor speed drops to zero 

9.9 SG feedwater isolation, RL04S06 valve closed 

11.5 SG feedwater isolation, RL01S06 valve closed 

13.4 SG feedwater isolation, RL02S06 valve closed 

13.8 SG feedwater isolation, RL03S06 valve closed 

17.5 Steam line isolation signal, valve RA06S01 closed 

97 First core dryout 

405 Second core dryout 

406 Start of ACC-4 injection 

414 Start of ACC-2 injection 

1365 Stop of ACC-4 injection 

1452 Stop of ACC-2 injection 

2057 Final core dryout 

2432 Start of LPIS injection (TJ01F02) 

2432 Start of LPIS injection (TJ03F02) 

2434 Start of LPIS injection (TJ04F02) 

2593 FRSB power switched off – end of the experiment 

In PSB-VVER Test-3 the small-break LOCA accident, vertically upward oriented cold leg No. 4 

break was simulated with boundary conditions simulating a coincidental loss of offsite power. 

The FRSB and core bypass power levels were controlled to simulate decay heat power. The 

ECCS actions were simulated. Three periods of FRSB cladding heatup were observed based 

on thermocouple measurements. The first one, just a minor deviation from the saturation 

temperature, was monitored only at thermocouples located above the axial position of YC01T40 

measurement (3.002 m above the FRSB bottom). The second, more distinctive FRSB heatup 

was also monitored only by thermocouples located 3.002 m above the FRSB bottom. This 

heatup was terminated by the accumulator injection. The third (final) heatup was monitored by 

the thermocouples located above 1.151 m above the FRSB bottom. This heatup was terminated 

by LPIS. 
63

4.2 TRACE V5.0 Thermal-Hydraulic Model of PSB-VVER Test Facility 

As already described in section 3.2, the facility database was compiled in an MS-ACCESS 

environment. Facility design data and thermal-hydraulic characteristics as reported in the frame 

of OECD/NEA PSB-VVER Project were organized in two tables with a total of 600 records. 

In the case of the TRACE model of PSB-VVER test facility development, the Symbolic Nuclear 

Analysis Package (SNAP) [8] was applied. At the first step, the reference case steady state run 

input deck of R5/M33 T-H model of PSB-VVER test facility (see section 3.3.1 above) was 

imported to SNAP using FileImportRELAP ASCII technique. At the second step, the 

R5/M33 model was converted to TRACE model of the facility using ToolsConvert to TRACE 

technique. This process ended up with revision 0 TRACE V5.0 model of PSB-VVER test facility. 

ToolsCheck Model protocol summarized that there were 15 errors in the model, each of 

“elevation loop closure” type. All these errors were corrected in revision 1 of the model. 

The model verification started with revision 2 of the model. As inherited from the R5/M33 model 

of the PSB-VVER test facility, the initial and boundary conditions were set up to simulate Test5a of OECD/NEA PSB-VVER project, i.e. the large-break LOCA experiment, 2x100% cold leg 

break with initial power scaled further down to 15% of prototypical [4]. The “null transient” 

method was applied to reproduce the test specific initial conditions. As the acceptance criterion, 

the code calculated data were requested to be within corresponding channel measurement 

accuracy intervals. The large-break LOCA transient was analyzed in code restart mode with 

specification of test specific trips in order to simulate the boundary conditions of the experiment. 

As the acceptance criterion, the calculated data were requested to reproduce the measured 

data qualitatively (phenomena and trends), with only key parameters (e.g. primary system 

pressure) being compared quantitatively. The acceptance criteria were fulfilled in revision 55 of 

the TRACE model of PSB-VVER test facility. In each revision, typically, one change to the 

model was introduced and its effect was checked in the steady state and transient runs. Major 

changes related to the application of TRACE VESSEL component for the reactor pressure 

vessel simulator model started from revision 10. 

The model validation started with revision 56 of the model. The initial and boundary conditions 

of the reference case calculations (steady state run and transient run) were set up for the facility 

model in question to simulate Test-3 of OECD/NEA PSB-VVER project, i.e. the small-break 

LOCA experiment, 4.1 % cold leg break with initial power scaled further down to 15% of 

prototypical [5]. This case is described in detail in Chapter 4.3 below. Please note that the 

validation case had been completed with revision 60 of the TRACE model of PSB-VVER test 

facility. 

In SNAP model editor, the thermal-hydraulic views of the TRACE V5.0 PSB-VVER model were 

reorganized as shown in Figures 27 to 32. 

As summarized in SNAP model editor navigator, TRACE V5.0 model of PSB-VVER facility 

consists of 338 hydraulic components, 1163 control systems, 156 thermal components, and 

1730 connections. 

The nodalization of reactor pressure vessel model is illustrated in Figure 27. VESSEL 

component (c010) with 4 axial levels, 2 radial rings, and 4 azimuthal sectors is used to model 

the upper part of the downcomer. Single junction components (c01n; n=1, 4) are used to model 

vessel (c010) to vessel (c015) connections for each azimuthal sector. VESSEL component 

(c015) with 10 axial levels, 1 radial ring, and 4 azimuthal sectors is used to model the middle 

part of the downcomer. Single junction components (c02n; n=1, 4) are used to model vessel 
64

(c015) to vessel (c020) connections for each azimuthal sector. VESSEL component (c020) with 

3 axial levels, 2 radial rings, and 4 azimuthal sectors is used to model the lower part of the 

downcomer. PIPE component (c037) with 3 cells is used to model a tube connecting the 

downcomer simulator and the lower plenum simulator. Multiple junction components (not shown 

in this view) are used to simulate axial and azimuthal flows between the downcomer inlet 

annular sectors. VESSEL component (c025) with 2 axial levels, 2 radial rings, and 4 azimuthal 

sectors is used to model the lower plenum simulator. Single junction components (c03n; n=1, 4) 

are used to model vessel (c025) to vessel (c030) connections for each azimuthal sector. 

VESSEL component (c030) with 21 axial levels, 1 radial ring, and 4 azimuthal sectors is used to 

model the core and upper plenum simulator. Single junction components (c04n; n=1, 4) are 

used to model vessel (c030) to vessel (c035) connections for each azimuthal sector in inner 

radial ring. Similarly, single junction components (c04n; n=6, 9) are used to model vessel (c030) 

to vessel (c035) connections for each azimuthal sector in outer radial ring. VESSEL component 

(c035) with 2 axial levels, 2 radial rings, and 4 azimuthal sectors is used to model the zone with 

nozzles. Single junction components (c05n; n=1, 4) are used to model vessel (c035) to vessel 

(c040) connections for each azimuthal sector in c035 inner radial ring. Similarly, single junction 

components (c05n; n=6, 9) are used to model vessel (c035) to vessel (c040) connections for 

each azimuthal sector in c035 outer radial ring. VESSEL component (c040) with 6 axial levels, 1 

radial ring, and 4 azimuthal sectors is used to model the upper head simulator. Two identical 

PIPE components (c076, c077) with 7 cells and one PIPE component (c078) with 22 cells are 

used to model the core bypass simulator. A part of pipe (c078) vertical section is heated. Heat 

structures (106n2; n=1, 4) each with 14 axial cells and 13 radial nodes are used to model the 

fuel rod simulator bundle. During Test-3 FRSB was heated uniformly along the radial and axial 

directions. Each of four fuel rod simulator heat structure represents 42 actual fuel rod simulators 

that communicate the energy with the corresponding azimuthal sectors of VESSEL (c030), from 

axial level 2 to axial level 15. Axial conduction flag was activated. A number of heat structures 

are used to model downcomer simulator, core simulator, upper plenum and upper head 

simulators piping walls in order to account for stored heat and eventually for heat losses to 

environment. Wallis CCFL model is set up and activated at axial edges 16 of VESSEL (c030) 

component. This is to account for CCFL effects at the upper core plate. 
65

Figure 27 TRACE V5.0 reactor pressure vessel model 
66

The nodalization of PCS loops is illustrated in Figure 28. PIPE components (cn05; n=1, 4) each 

with 14 cells are used to model the hot legs. PIPE components (cn10; n=1, 4) each with 1 cell 

are used to model steam generator inlets. Steam generator primary side models consist of hot 

header model, PIPE components (cn15; n=1, 4) each with 10 cells, horizontal tube bundle 

simulator, PIPE components (cn2r; n=1, 4; r=1, 9) each with 11 cells, and cold header model, 

PIPE components (cn31; n=1, 4) each with 10 cells. In the PSB-VVER test facility, the steam 

generator simulator consists of 34 nearly horizontal coiled tubes. In TRACE model, the 

individual tubes are grouped to have nine tube rows along the height of the steam generator 

vessel. PIPE component (cn33; n=1, 4) each with 1 cell are used to model steam generator 

outlets. 

Figure 28 TRACE V5.0 primary coolant system model – Loop No. 4 

PIPE components (cn35, n=1, 4) each with 44 cells are used to represent the loop seals. 

Reactor coolant pumps are modelled with PUMP components (cn36, n=1, 4) each with 2 cells. 

The intact loop cold legs are modelled using PIPE components (cn40, n=1, 4) each with 2 cells, 

valve components (cn41, n=1, 4), and pipe components (cn42, n=1, 4) with 14 control volumes. 

In Test-3, the break was simulated to occur at cold leg No. 4, so the nodalization of the broken 

loop cold leg follows standard guidelines to model the single ended cold leg break. The actions 

of LPIS are modelled using FILL components. The injection points are Test-3 specific. 
67

Heat structures associated with those hydraulic components model piping walls. 

The nodalization of the pressurizer and the pressurizer surgeline models is illustrated in 

Figure 29. Only the loop No. 4 hot leg connected surgeline is modelled using PIPE component 

(c507) with 4 cells, VALVE component (c508), and PIPE component (c509) with 23 cells. The 

model of the common part of the pressurizer surgeline begins with the PIPE component (c510) 

with 1 cell and continues using the PIPE component (c511) with 4 cells. The PIPE component 

(c515) with 24 cells is used to model the pressurizer vessel. VALVE component (c519), 

connected to the top of the pressurizer, and the BREAK component (c520) are used to adjust 

the desired steady state pressurizer pressure. Similarly, the mass flow controlled single junction 

component (c529), and the BREAK (c530) component are used to adjust the desired steady 

state pressurizer coolant level. Mass flow controlled single junction components (c541 and 

c549), and BREAK components (c540 and c550) are used to achieve the desired steady state 

surgeline coolant temperature. In transient runs, the steady state control component gains are 

set to zero. 

The nodalization of the accumulators and accumulator surgelines is illustrated in Figure 30. 

Hydro-accumulators (TH10B01 and TH30B01) were isolated during the Test-3. Two active 

hydro accumulators (TH20B01 and TH40B01) were connected to the downcomer. The check 

valves (c6n2; n=1, 4) are modelled explicitly. The model also accounts for the actual locations of 

Test-3 specific LPIS injection. 
68

Figure 29 TRACE V5.0 pressurizer model 
69

Figure 30 TRACE V5.0 emergency core cooling system model 

The nodalization of SG secondary side and associated steam and feedwater lines is illustrated 

in Figure 31. The combination of single cell PIPE components is used to model the steam 

generator secondary side. Considering the top view, the central part of the vessel where the 

coiled tube bundle is located is treated as the “boiler” section, whereas the peripheral annular 

part is treated as the “downcomer” section. The lateral cross flow paths are modelled to 

simulate the internal recirculation of the secondary coolant. Nine rows of PIPE components 

correspond to the nine row tube bundle simulated on the steam generator primary side. The 

combination of three single PIPE components (e.g. c460, c470, and c471) is used to account for 

the steam-water separation. The upper parts of steam generator vessels are modelled using the 

pipe components (cn75; n=1, 4) each with 3 cells. In PSB-VVER facility, there are two main 

steam lines per one steam generator. The 1.5 MWt steam line is designed to remove the heat 

generated by the 1.5 MWt FRSB power. If 10 MWt FRSB power (to follow 1:300 scale) were 

used for the particular test, the 10 MWt steam line would have been used. During the PSBVVER experiments carried out within the frame of OECD/NEA PSB-VVER project only the 
70

1.5 MWt FRSB power systems were available. However, in order to account for unisolatable 

parts of 10 MWt steam lines and also for future use of the model, the 10 MWt steam lines are 

modelled and discussed here. 

Figure 31 TRACE V5.0 secondary coolant system model – Loop No. 4 

The 1.5 MWt steam line models consist of PIPE components (cn84; n=1, 4), VALVE 

components (cn85; n=1, 4), PIPE components (cn86; n=1, 4), and PIPE components (cn88; 

n=1, 4). Actually, the later components are parts of 10 MWt steam lines. 

The 10 MWt steam line models consist of PIPE components (cn92; n=1, 4), VALVE 

components (cn91; n=1, 4), PIPE components (cn90; n=1, 4), and VALVE components (cn89; 

n=1, 4). Valves in the 10 MWt steam lines were closed throughout the Test-3. 

The lengths and therefore the nodalization of individual steam lines vary accordingly to the loop 

specific design of the PSB-VVER test facility. 

The combinations of FILL components (cn95; n=1, 4), and pipe components (cn97; n=1, 4) are 

used to model individual steam generator feedwater lines. In this, the facility steam generator 

feedwater system model is highly simplified. 
71

VALVE components (cn81; n=1, 4) are used to simulate actions of power operated relief valves 

or safety valves. The test specific trip logic has to be set up to model the steam dump to 

atmosphere. BREAK components (cn82; n=1, 4) are used to model the boundary conditions. 

The nodalization of the main steam line is illustrated in Figure 32. Main steam line is simulated 

using two PIPE components. VALVE component (c957) models the steam dump valve 

RA06S01. 

A number of heat structure components are used to simulate heat transfer in steam generators 

and heat losses from the pipe walls to the environment. 

There are 380 control blocks used in the model. They are set up to simulate I&C actions, to 

drive the model to the desired steady state, to facilitate the comparison of calculated data with 

measured data (e.g. differential pressures), etc. 

Figure 32 TRACE V5.0 main steam line model 
72

4.3 TRACE V5.0 PSB-VVER Test Facility Model Validation 

Third step in the thermal-hydraulic system code assessment process is defined in [2] as: 

 Perform assessment calculations. Compare the test data with the results of calculations 

for the key parameters. 

This study follows the concept of OECD/NEA BEMUSE Project, specifically Phase II [13]. In 

this, the reference case was run in two steps, i.e. the steady state run and the transient run to 

enable code data evaluation on the steady state level and on the transient level. 

4.3.1 Reference Case 

Similarly as in Phase II of the BEMUSE project, reference case in PSB-VVER Test-3 post test 

analysis was run and the test data were compared with the results of the calculation in two 

steps, on the steady state level, and on the transient level. A 500 second “null transient” code 

calculation was run to simulate the test specific initial conditions. Code control blocks were set 

up to force the code to reach predefined primary system coolant mass flows and secondary 

system steam line pressures. In total 127 measured and calculated parameters characterizing 

facility initial conditions were compared. In close to 80% of cases, the calculated value at 500 

seconds was within the measurement accuracy interval <-2; +2>, where  is the measured 

value of the parameter and 2 is the measurement accuracy of the parameter as reported in [5]. 

Table 15 illustrates this steady state data evaluation process in case of the pressurizer model. 

Please refer to Table 2 and Figure 8 for the measurement channel identification. Please note 

that pressures (P) are in MPa, coolant temperatures (T) are in °C, and differential pressures 

(DP) are in kPa. 

Table 15 TRACE V5.0 Test-3 reference case steady state data evaluation – PRZ 

Experiment Reference case 

(steady state) 

Measurement 

accuracy (2) 

Is cntrlvar value 

in the interval 

<-2; +2>? 

Channel ID Value (μ) Control 

block Value 

YP01P01 15.52 -506 15.52 0.07 TRUE 

YP01DP02 -25.91 -508 -23.73 1.3 FALSE 

YP01DP03 -37.75 -509 -37.32 1 TRUE 

YP01L02 3.05 -597 3.05 0.3 TRUE 

The calculated data for the pressurizer pressure (cb_506) match exactly the measured data 

(YP01P01) and similarly, the calculated data for the pressurizer level (cb_507) match exactly 

the measured data (YP01L02) due to corresponding steady state calculation controls. 
73

In case of the differential pressure across the lower part of the pressurizer vessel and upper part 

of the surgeline (YP01DP02) compared with control block 508, the acceptance criterion is not 

fulfilled, but the result is tolerated due to following reasons. The lower experimental value 

indicates the presence of colder water in this section. The location of pressurizer heaters and its 

actions before the initiation of the transient contradicts the presence of colder water in the 

pressurizer vessel. There was only one measurement of coolant temperature over the surgeline. 

The model was initialized to reproduce that measurement. Table 16 illustrates the steady state 

data evaluation process in the case of the reactor pressure vessel model. 
74

Table 16 TRACE V5.0 Test-3 reference case steady state data evaluation – RPV 

Experiment Reference case 

(steady state) 

Measurement 

accuracy (2) 

Is cntrlvar value 

in the interval 

<-2; +2> 

Channel ID Value (μ) Control 

block Value 

YC01T01 285 -2 285.56 4.3 TRUE 

YC01T02 282 -3 285.56 3.7 TRUE 

YC01T03 292 -4 288.97 3 FALSE 

YC01T04b 311 -56 312.12 4.2 TRUE 

YC01T05 292 -59 293.84 3 TRUE 

YC01P16 15.51 -33 15.53 0.05 TRUE 

YC01P17 15.6 -29 15.54 0.06 TRUE 

YC01P18 15.58 -9 15.54 0.04 TRUE 

YC01DP01 -8.71 -10 -8.63 0.2 TRUE 

YC01DP02 -4.02 -12 -3.69 0.2 FALSE 

YC01DP03 -25.7 -14 -25.92 0.7 TRUE 

YC01DP04 -11.82 -16 -11.15 1 TRUE 

YC01DP05 -8.83 -18 -8.94 0.2 TRUE 

YC01DP06 8.67 -20 8.47 0.2 TRUE 

YC01DP07 -7.38 -22 -7.23 1 TRUE 

YC01DP08 -7.9 -24 -7.81 0.2 TRUE 

YC01DP09 -4.97 -26 -4.58 0.2 FALSE 

YC01DP10 -8.29 -28 -7.88 1 TRUE 

YC01DP11 -5.96 -30 -5.47 0.2 FALSE 

YC01DP12 -5.83 -32 -5.89 0.2 TRUE 

YC01DP13 -11.08 -34 -11.17 0.2 TRUE 

YC01DP14 -10.72 -36 -10.38 1.1 TRUE 

YC01DP15 -17.67 -38 -17.04 1.1 TRUE 

YC01DP16 21.3 -39 17.75 1 FALSE 

YC01DP17 -35.98 -42 -31.54 1 FALSE 

*DP_CORE -28.458 -58 -27.50 1 TRUE 

*DP_UPUH -51.127 -60 -49.96 1.1 FALSE 

YC01N01 1129 -57 1129.00 15 TRUE 

YC01N02 14.9 -6 14.9 0.4 TRUE 
75

Please note that pressures (P) are in MPa, coolant temperatures (T) are in °C, differential 

pressures (DP) are in kPa, and powers (N) are in kW. 

In case of coolant temperature at the core bypass outlet (YC01T03 vs. cb_004), the acceptance 

criterion is missed by 0.03 K and the case is tolerated. In case of the differential pressure across 

the upper downcomer (YC01DP02 vs. cb_12), the acceptance criterion is not fulfilled. The case 

is further evaluated as follows. The sum of two successive differential pressure measurements 

YC01DP01 and YC01DP02 was compared to the sum of control blocks cb_010 and cb_012. 

The acceptance criterion is fulfilled for this derived parameter. Moreover, in the verification step 

steady state data (PSB-VVER Test-5a) evaluation for YC01DP01 and YC01DP02, the 

TRUE/FALSE results were reversed. In case of the differential pressure over the core section 

(YC01DP09 vs. cb_026) the acceptance criterion is not fulfilled. The case is tolerated because 

there is no evidence about the differences in the individual spacer grid design. When the 

derived parameter comparison technique is applied, i.e. (*DP_CORE vs. cb_058), the 

acceptance criterion is fulfilled. The “FALSE” case of YC01DP16 is tolerated since if the sum of 

differential pressures from YC01DP02 to YC01DP13 (18.54 kPa) was compared to cb_039, the 

acceptance criterion is fulfilled. 
76

Table 17 illustrates the steady state data evaluation process in case of the primary coolant 

system model. Measured and calculated data of loop No. 1 are compared. The results of 

reference case steady state data evaluation of remaining loops are similar. The “FALSE” cases 

are attributed to the fact that thermal-hydraulic characteristics of primary system loops are 

considered identical in the model. 

Table 17 TRACE V5.0 Test-3 reference case steady state data evaluation – PCS 

Experiment Reference case 

(steady state) 

Measurement 

accuracy (2) 

Is cntrlvar value 

in the interval 

<-2; +2>? 

Channel ID Value (μ) Control 

block Value 

YA01P01 15.58 -106 15.54 0.22 TRUE 

YA01P03 15.54 -118 15.53 0.23 TRUE 

YA01P08 15.57 -121 15.55 0.06 TRUE 

YA01P09 15.5 -103 15.52 0.06 TRUE 

YA01T02 283 -114 285.64 3 TRUE 

YA01T03 310 -108 311.89 3 TRUE 

YA01T26 285 -111 285.64 3.7 TRUE 

YA01T32 285 -110 285.63 0.5 FALSE 

YA01T33 285 -112 285.64 0.5 FALSE 

YA01DP01 1.88 -115 1.75 0.2 TRUE 

YA01DP02 9.11 -116 9.03 0.2 TRUE 

YA01DP03 2.95 -8 3.21 0.1 FALSE 

YA01DP04 -22.19 -43 -22.15 0.3 TRUE 

YA01DP05 -24.08 -44 -24.51 1.1 TRUE 

YA01DP06 -15.3 -45 -15.99 1 TRUE 

YA01DP08 4.55 -46 4.51 1 TRUE 

YA01DP09 -10.73 -47 -11.06 1 TRUE 

YA01DP10 0.41 -48 0.41 0.2 TRUE 

YA01DP11 -0.99 -124 -1.08 0.2 TRUE 

YA01DP13 -20.38 -49 -20.75 1 TRUE 

YA01DP14 -20.42 -50 -20.26 1 TRUE 
77

Table 18 illustrates the steady state data evaluation process in case of the secondary coolant 

system model. 

Table 18 TRACE V5.0 Test-3 reference case steady state data evaluation – SG 

Experiment 

Reference case 

(steady state) 

Measurement 

accuracy (2) 

Is cntrlvar value 

in the interval 

<-2; +2>? 

Channel ID Value (μ) Control 

block Value 

YB01P01 6.88 -511 6.91 0.05 TRUE 

YB02P01 6.91 -521 6.92 0.05 TRUE 

YB03P01 6.93 -531 6.92 0.05 TRUE 

YB04P01 6.88 -541 6.91 0.05 TRUE 

On the transient level, the test data were compared with the results of the calculation using 

FFTBM [7]. The selection of 22 parameters to be compared closely follows earlier studies [13] 

and [15]. For each parameter, the FFTBM algorithm is run in time window 0 – 2590 s (whole 

transient) to get average amplitudes (AA) and weighted frequencies (WF). These results are 

further combined using corresponding weighting factors to obtain total average amplitude (AAtot) 

and total weighted frequency (WFtot). In [7], the authors discuss the accuracy of code 

predictions as follows: 

 AAtot  0.3 characterize very good code predictions 

 0.3  AAtot  0.5 characterize good code predictions 

 The acceptability factor K is then defined for total average amplitude K = 0.4, and for 

primary system pressure K = 0.1. The acceptable code predictions are those with 

AAtot  0.4, and AAPSP  0.1. 
78

The results of FFTBM application on the reference case are summarized in Table 19. Since 

AAtot < 0.4 and AA1 = AA2 =0.1, the reference case code predictions are acceptable. 

Table 19 Test-3 TRACE V5.0 reference case FFTBM results 

N Parameter ID Exp ID Cal AA 

1 Upper plenum pressure YC01P16 cb_033 0.080 

2 Pressurizer pressure YP01P01 cb_506 0.060 

3 Accumulator #2 pressure TH02P01 pn620A04 0.103 

4 Accumulator #4 pressure TH04P01 cb_300 0.090 

5 Steam generator #1 secondary side 

pressure YB01P01 cb_511 0.390 

6 Steam generator #3 secondary side 

pressure YB03P01 cb_531 0.384 

7 Core inlet coolant temperature YC01T02 cb_003 0.155 

8 Core outlet coolant temperature YC01T04b cb_056 0.499 

9 Integrated break flow MBr cb_725 0.229 

10 Primary coolant mass M1k cb_580 0.251 

11 Accumulator #2 level TH02L01 cb_204 0.100 

12 Accumulator #4 level TH04L01 cb_200 0.088 

13 Differential pressure across downcomer YC01DP03 cb_014 0.824 

14 Differential pressure across downcomer YC01DP04 cb_016 0.952 

15 Differential pressure across the 

downcomer outlet and core simulator inlet YC01DP06 cb_020 0.899 

16 Differential pressure across the core DP_CORE cb_058 0.765 

17 Differential pressure across the upper 

plenum DP_UPUH cb_060 0.370 

18 Differential pressure across the upper 

plenum outlet and downcomer inlet YC01DP16 cb_039 0.895 

19 Peak cladding temperature YC01T11 cb_pct 0.451 

20 Cladding temperature – core bottom YC01T113 cb_bot 0.163 

21 Cladding temperature – core center YC01T55 cb_mid 0.539 

22 Cladding temperature – core top YC01T39 cb_top 0.322 

 Total average amplitude - AAtot - - 0.306 
79

Appendix C contains the figures with graphical comparisons of measured and TRACE 

calculated results. 

Figure C.1 compares the data for integrated break mass flows. In the experiment, this 

parameter was not measured directly, but a specific algorithm for measured data evaluation was 

applied. In agreement with experimental data evaluation, the code predicts first vaporization at 

break at 108 s. After that, experimental data evaluation shows the break flow stagnation for 

approximately 50 s. In the reference case calculation with Cd = 0.85, no attempt to reproduce 

this phenomenon was applied. Approximately from 160 s until the end of the transient, the 

difference between the results slightly increases. The experimental data evaluation does not 

account for break mass flow increase following the LPIS injection. 

Figure C.2 and Figure C.3 show good agreement between measured and calculated primary 

system pressures. Consequently, the timing of the accumulator injection is well predicted in the 

model (see Figures C.4 – C.7). 

Figure C.8 shows the comparison of the primary system coolant mass between that evaluated 

in experiment and TRACE results. The experimental data is the product of the specific algorithm 

using data measured directly. According to the experimenters the accuracy of this evaluation is 

± 50 kg. As shown in Table 19, parameter No. 10, the average amplitude is 0.251 which 

characterizes very good code prediction. 

Figures C.9 – C.12 show the comparison of the secondary system pressure (at the steam 

generator steam dome) between measured and TRACE results. Approximately from 180 s in 

the transient the secondary system pressure exceeds the primary system pressure. The heat 

transfer across the steam generator tube bundle is modelled in TRACE, however, in the 

reference case; the heat losses from steam line pipes to environment were not simulated. As 

shown in Table 19, parameters No. 5 and No. 6, the average amplitudes are just below the 

acceptance limit (K = 0.4) for good code predictions. The heat loss model from secondary 

system pipe walls to the environment is considered for further refining the reference case results 

before the uncertainty and sensitivity study. 

Figure C.13 shows the comparison of the core simulator inlet coolant temperature between 

measured and TRACE results. The measured data follows the saturation temperature. During 

the period of the accumulator injection, the TRACE data shows that subcooled liquid reached 

the core simulator inlet. 

Figure C.14 compares core simulator outlet coolant temperatures. The measured data show 

three periods of steam superheat during the final FRSB dryout. TRACE data reproduce the 

phenomenon qualitatively well. Quantitatively, as shown in Table 19, parameter No. 8, the 

average amplitude is 0.499 just on the edge (0.5) of good code prediction characteristics. Due 

to the weighting factors applied when calculating AAtot, this parameter is by far the largest 

contributor (0.047) to AAtot = 0.306. 

Figures C.15 – C.17 show the comparison of differential pressures across the downcomer and 

lower plenum simulators between measured and TRACE results. Qualitatively, the code 

predictions agree well with the measurements except cb_016 compared to YC01DP04 on 

Figure C.16, where the calculated liquid loss after 2200 s is more significant. 

Figures C.18 – C.20 show the comparison of differential pressures across other parts of the 

reactor pressure vessel simulator between measured and TRACE results. Qualitatively, the 

code predictions agree with the measurements. Quantitatively, good code prediction is reached 

only for cb_060 compared to sum of YC01DP11 to YC01DP15, as shown I Table 19, parameter 
80

No. 17, which contributes 0.006 to AAtot. The individual contribution of other selected 

differential pressure parameters to AAtot varies from 0.012 to 0.014. 

In case of SB LOCA events, the transient is affected by the loop seal clearing processes. If at 

least one loop seal remains cleared, then coolant level depression in the core region is limited. 

Figures C.21 – C.28 compare the differential pressures in loop seal descending and ascending 

legs for all the loops. After the accumulator injection is terminated, the loop seal No. 4 remains 

clear. Please note that loop No. 4 is the broken loop. 

Finally, the measured and calculated cladding temperatures are compared on Figures C.29 –

 C.32. Similarly to approach adopted in [13] and [15], the representative measurements are 

selected for the bottom, middle, and top third of the FRSB. In addition, the measurement 

channel data with maximum cladding temperature recorded during the transient (YC01T11) is 

compared to the corresponding TRACE data. TRACE predicts only temporary and limited 

heatup during the final heatup period observed in experiment. Two axial level calculated data 

are compared to YC01T113 thermocouple readings, which show no FRSB heatup at that 

elevation. Only the final heatup was recorded by thermocouples located close to the middle of 

FRSB. The readings of YC01T55 thermocouple measurements are representative and 

compared to each of four azimuthal sectors calculated data for the given axial position. TRACE 

predicts the start of the final heatup later and the prediction of its termination agrees well with 

the measurement. Azimuthal sector No. 4 (adjacent to the broken loop) is slightly favoured for 

cooling. The second and third (final) heatup were recorded by the thermocouples located close 

to the center of the top third of FRSB. The readings of YC01T39 thermocouple measurements 

are representative and compared to each of four azimuthal sectors calculated data for the given 

axial position. TRACE predicts the second heatup to begin earlier and to be of minimal and 

shortly lasted deviation from the saturation temperature. TRACE predicts the final heatup to 

begin later. The code prediction of its termination agrees well with the measurement. Measured 

and calculated peak cladding temperatures are shown in Figure C.32. Measured data document 

all three periods of cladding heatup at the top of the FRSB. First measured heatup was brief in 

time and of very small overheat with respect to saturation temperature. This heatup is not 

predicted by the code. During the experiment, the second dryout begun at about 400 s. With 

correct timing, but lower in magnitude, the corresponding FRSB cladding heatup is predicted by 

the code. During the experiment, the third dryout begun at about 2060 s. Delayed by 80 s, and 

lower in magnitude, the corresponding FRSB cladding heatup at that elevation is predicted by 

the code. During the experiment, the third dryout was terminated by the LPIS injection. The 

timing of this event is well predicted by the code. 

4.3.2 Preliminary sensitivity study 

This section is titled “preliminary” sensitivity study in order to distinguish between activities being 

carried out before and after the uncertainty study. Before the uncertainty study is started, 

preliminary sensitivity study may be carried out, e.g. [13] with the aim to determine the code 

response to the variation of values of input parameters which may be considered in a later 

uncertainty study. 

If a full scale preliminary sensitivity study were performed within the project reported here, the 

selection of code input parameters would have corresponded closely to the case reported 

in [13]. In this project the preliminary sensitivity study is limited to one uncertain parameter, 

namely the discharge coefficient in the critical flow model. As shown in comment to Figure C.1, 

the reference case Cd = 0.85. In the sensitivity run with Cd = 1.0, the code response, in 
81

particular the prediction of peak cladding temperature, is evaluated. The transient was analyzed 

to 2500 s, and the key results are compared in Figure C.33. The final dryout started earlier in 

the sensitivity calculation and its timing closely matched the experimental data. Calculated peak 

cladding temperature is expected to be overpredicted by approximately 50 K the measured 

value since cladding temperature in azimuthal sector #4 has already reached its maximum and 

taking into account the delay in the timing of peak cladding temperatures in other azimuthal 

sectors as in Figure C.32. 

4.4 Conclusions on TRACE V5.0 Code Assessment 

The independent code assessment reported here is based on TRACE V5.0 PSB-VVER thermalhydraulic model validation against small break LOCA test data. The experiment denoted as 

Test-3 [5] was carried out within the frame of OECD/NEA PSB Project. The conclusions are 

drawn to each of four major steps considered: 

 PSB-VVER model conversion 

 PSB-VVER model verification 

 PSB-VVER model validation 

 Preliminary sensitivity study 

SNAP [8] was successfully applied for R5/M33 PSB-VVER model conversion to TRACE V5.0 

PSB-VVER model. In addition SNAP Model Editor was successfully applied in TRACE V5.0 

PSB-VVER model verification step for gradual abandonment of RELAP5/MOD3.3 inherited 1D 

component modelling technique toward TRACE 3D component modelling technique. The 

number of TRACE V5.0 PSB-VVER model versions (45) developed during the model 

verification step may appear high. However, the “one change at the time” technique which had 

been intentionally preferred by new TRACE code user finally paid off. In the validation step only 

4 TRACE V5.0 PSB-VVER model versions were needed for refining the initial and boundary 

conditions of the test selected. 

In both the model verification and the model validation steps, the model was shown to be 

qualified on the steady state level. In close to 80% of the 120 parameters compared, the 

acceptance criterion was fulfilled. Other cases were tolerated primarily due to the modelling 

approach that was followed, i.e. no consideration of differences in the primary loop hydraulic 

characteristics which may be test dependent. 

In the verification step, the model was shown to reproduce main phenomena of the large break 

LOCA test qualitatively and quantitatively well. However, according the FFTBM criteria applied it 

was not possible to conclude that the model was qualified on the transient level. In this case the 

total average amplitude (AAtot = 0.41) that exceeded the acceptability factor (K = 0.4). The main 

challenge to the TRACE V5.0 PSB-VVER model in reproducing large break LOCA test data 

more accurately was identified in modelling upper plenum ECCS injection related phenomena. 

Less accurate predictions of differential pressures across the upper plenum and core simulators 

lead to the lesser accuracy of cladding temperatures during the transient. 

In the validation step, the model was shown to be qualified on the transient level. The FFTBM 

applied acceptance criteria were fulfilled with conclusion of good code predictions. 
82

Preliminary sensitivity study shown that model is capable of reproducing key safety parameter 

(peak cladding temperature) when considering the uncertainty of critical discharge coefficient. 

TRACE V5.0 PSB-VVER thermal-hydraulic model was validated for small break LOCA analyses 

against the experimental data from Test-3. This independent validation case may contribute to 

the TRACE V5.0 code assessment. 
83

5. REFERENCES 

1. MELIKHOV, O. I., et al. ”Report about PSB-VVER Description (including 

measurement system)”, OECD PSB-VVER Project PSB-03, Elektrogorsk, 2003. 

2. TRACE V5.0 Assessment Manual, Main Report, USNRC Division of Risk Assessment 

and Special Projects, March 2010. 

3. BEMUSE Phase VI Report, Status report on the area, classification of the methods, 

conclusions and recommendations; NEA/CSNI/R(2011)4, March 2011. 

4. ELKIN, I. V., et al. “Experimental Data Report Test 5a (Version 1)”, OECD PSB-VVER 

Project PSB-31, Elektrogorsk, 2008. 

5. MELIKHOV, O. I., et al. “Post-Test Full Experimental Data Report (Test-3)”, OECD 

PSB-VVER Project PSB-18, Elektrogorsk, 2004. 

6. D’ Auria, F., Review of methods of Calculation, Assumptions and Acceptance Criteria 

for the Analyses, IAEA Regional Workshop on Evaluation and Licensing of NPP 

Modifications, Jožef Stefan Institute, Ljubljana, 21-25 June, 1999. 

7. Prošek, A., D’ Auria, F., Mavko, B., Review of quantitative accuracy assessments with 

fast Fourier transform based Method (FFTBM), Technical Note, Nucl. Eng. Des., 217, 

179-206, 2002. 

8. Symbolic Nuclear Analysis Package, Version 2.0.7, August 15, 2011 

© 2002-2011 Applied Programming Technology, Inc. 

9. RELAP5/MOD3.3 Code Manual, Volume III: DEVELOPMENTAL ASSESSMENT 

PROBLEMS, Information Systems Laboratories, NUREG/CR-5535/Rev 1-Vol III, 

December 2001. 

10. Douglas L. Reeder: “LOFT System and Test Description (5.5 FT Nuclear Core)”, 

NUREG/CR-0247, TREE-1208, July 1978. 

11. Paul D. Bayless, Janice M. Divine: “Experiment Data Report for LOFT Large Break 

Loss-of-Coolant Experiment L2-5”, NUREG/CR-2826, INEL, August 1982. 

12. “CSNI Integral Test Facility Validation Matrix for the Assessment of Thermal-Hydraulic 

Codes for LWR LOCA and Transients”, Committee on the Safety of Nuclear 

Installations, OECD Nuclear Energy Agency, NEA/CSNI/R(96)17 (also referenced as: 

OCDE/GD(97)12, July 1996, France. 

13. BEMUSE Phase II Report, Re-analysis of the ISP-13 Exercise, Post Test Analysis of 

the LOFT L2-5 Test Calculation; NEA/CSNI/R(2006)2, May 2006. 

14. Validation Matrix for the Assessment of Thermal-Hydraulic Codes for VVER LOCA 

and Transients, A Report by the OECD Support Group on the VVER ThermalHydraulic Code Validation Matrix; NEA/CSNI/R(2001)4, June 2001. 

15. Del Nevo, A., et al. Benchmark on OECD NEA PSB-VVER Project Test 5a: LB-LOCA 

Transient in PSB-VVER Facility, DIMNP NT 638(08) Rev. 0, Pisa, November 2008. 
84

16. “BEMUSE Phase III Report, Uncertainty and Sensitivity Analysis of the LOFT L2-5 

Test”, Nuclear Energy Agency, Committee on the Safety of Nuclear Installations, 

NEA/CSNI/R(2007)4, 

Coordinators: A. de Crécy, P. Bazin (CEA), 

Authors (Organizations): P. Bazin, A. de Crécy (CEA, France), H. Glaeser, T. Skorek 

(GRS, Germany), J. Joucla, P. Probst (IRSN, France), B. Chung (KAERI, South 

Korea), M. Kyncl, R. Pernica (NRI-1, Czech Republic), J. Macek, R. Meca (NRI-2, 

Czech Republic), R. Macian (Switzerland), F. D’Auria, A. Petruzzi (UNIPI, Italy), 

M. Perez, F. Reventos (UPC, Spain), K. Fujioka (JNES, Japan). 

17. A. de Crécy et al. “Uncertainty and sensitivity analysis of the LOFT L2-5 test: Results 

of the BEMUSE programme”. Nuclear Engineering and Design 238 (2008), 3561-3578 

18. M. Perez et al., Uncertainty and sensitivity analysis of a LBLOCA in a PWR Nuclear 

Power Plant: Results of the Phase V of the BEMUSE programme. Nuclear 

Engineering and Design 241 (2011) 4206– 4222. 

19. “RELAP5/MOD3.3 Code Manual”, Volume II: Appendix A Input Requirements, 

Information Systems Laboratories, Inc., NUREG/CR-5535/Rev P4-Vol II App A, 

October 2010. 

20. “RELAP5/MOD3.3 Code Manual”, Volume I – Volume VIII, Information Systems 

Laboratories, Inc,. NUREG/CR-5535/Rev P4, October 2010. 

21. Bemuse Phase IV Report, “Simulation of a LB-LOCA in ZION Nuclear Power Plant – 

Appendices A to D”, Nuclear Energy Agency, Committee on the Safety of Nuclear 

Installations, NEA/CSNI/R(2008)6/vol2, November 2008. 

22. Sachs L., “Applied Statistics“, Springer-Verlag, New York, 1982. 

23. StatSoft, Inc. (2006). Electronic Statistics Textbook. Tulsa, OK: StatSoft. WEB: 

http://www.statsoft.com/textbook/stathome.html 

24. TRACE V5.0 Assessment Manual, Appendix C: Integral Effects Tests, USNRC 

Division of Risk Assessment and Special Projects, March 2010. 

25. Annunziato, A., Quick Look Report on LOBI-MOD2 Test BL-34 (LOBI/BETHSY 

Counterpart Test), J.R.C., LQC 91-60. 
85

APPENDIX A 

List of Figures 

Fig. A.1 PSB-VVER Test-5a Integrated Break Flow 

Fig. A.2 PSB-VVER Test-5a Upper Plenum Pressure 

Fig. A.3 PSB-VVER Test-5a Pressurizer Pressure 

Fig. A.4 PSB-VVER Test-5a Accumulator #1 Pressure 

Fig. A.5 PSB-VVER Test-5a Accumulator #1 Level 

Fig. A.6 PSB-VVER Test-5a Accumulator #4 Pressure 

Fig. A.7 PSB-VVER Test-5a Accumulator #4 Level 

Fig. A.8 PSB-VVER Test-5a HPIS Mass Flow 

Fig. A.9 PSB-VVER Test-5a Primary Coolant Mass 

Fig. A.10 PSB-VVER Test-5a Steam Generator #1 Pressure 

Fig. A.11 PSB-VVER Test-5a Steam Generator #2 Pressure 

Fig. A.12 PSB-VVER Test-5a Steam Generator #3 Pressure 

Fig. A.13 PSB-VVER Test-5a Steam Generator #4 Pressure 

Fig. A.14 PSB-VVER Test-5a Core Simulator Inlet Liquid Temperature 

Fig. A.15 PSB-VVER Test-5a Core Simulator Outlet Coolant Temperature 

Fig. A.16 PSB-VVER Test-5a Downcomer Simulator (6310 – 2790 mm) Differential 

Pressure 

Fig. A.17 PSB-VVER Test-5a Downcomer Simulator (2790 – 1010 mm) Differential 

Pressure 

Fig. A.18 PSB-VVER Test-5a Lower Plenum Simulator (1010 – 1915 mm) 

Differential Pressure 

Fig. A.19 PSB-VVER Test-5a Core Simulator (5440 – 1915 mm) Differential 

Pressure 

Fig. A.20 PSB-VVER Test-5a Upper Plenum & Upper Head Simulator (12525 –

 5440 mm) Differential Pressure 

Fig. A.21 PSB-VVER Test-5a Downcomer Simulator to Upper Plenum Simulator 

(6870 – 8650 mm) Differential Pressure 

Fig. A.22 PSB-VVER Test-5a FRSB Bottom Third Cladding Temperature 

Fig. A.23 PSB-VVER Test-5a FRSB Middle Third Cladding Temperature 

Fig. A.24 PSB-VVER Test-5a FRSB Top Third Cladding Temperature 

Fig. A.25 PSB-VVER Test-5a FRSB Peak Cladding Temperature 

87

Fig. A.1: PSB-VVER Test-5a Integrated Break Mass Flow 
88

Fig. A.2: PSB-VVER Test-5a Upper Plenum Pressure 
89

Fig. A.3: PSB-VVER Test-5a Pressurizer Pressure 
90

Fig. A.4: PSB-VVER Test-5a Accumulator #1 Pressure 
91

Fig. A.5: PSB-VVER Test-5a Accumulator #1 Level 
92

Fig. A.6: PSB-VVER Test-5a Accumulator #4 Pressure 
93

Fig. A.7: PSB-VVER Test-5a Accumulator #4 Level 
94

Fig. A.8: PSB-VVER Test-5a HPIS Mass Flow 
95

Fig. A.9: PSB-VVER Test-5a Primary Coolant Mass 
96

Fig. A.10: PSB-VVER Test-5a Steam Generator #1 Pressure 
97

Fig. A.11: PSB-VVER Test-5a Steam Generator #2 Pressure 
98

Fig. A.12: PSB-VVER Test-5a Steam Generator #3 Pressure 
99

Fig. A.13: PSB-VVER Test-5a Steam Generator #4 Pressure 
100

Fig. A.14: PSB-VVER Test-5a Core Simulator Inlet Liquid Temperature 
101

Fig. A.15: PSB-VVER Test-5a Core Simulator Outlet Coolant Temperature 
102

Fig. A.16: PSB-VVER Test-5a Downcomer Simulator (6310 – 2790 mm) Differential 

Pressure 
103

Fig. A.17: PSB-VVER Test-5a Downcomer Simulator (2790 – 1010 mm) Differential 

Pressure 
104

Fig. A.18: PSB-VVER Test-5a Lower Plenum Simulator (1010 – 1915 mm) Differential 

Pressure 
105

Fig. A.19: PSB-VVER Test-5a Core Simulator (5440 – 1915 mm) Differential Pressure 
106

Fig. A.20: PSB-VVER Test-5a Upper Plenum & Upper Head Simulator (12525 –

 5440 mm) Differential Pressure 
107

Fig. A.21: PSB-VVER Test-5a Downcomer Simulator to Upper Plenum Simulator 

(6870 – 8650 mm) Differential Pressure 
108

Fig. A.22: PSB-VVER Test-5a FRSB Bottom Third Cladding Temperature 
109

 

Fig. A.23: PSB-VVER Test-5a FRSB Middle Third Cladding Temperature 
110

Fig. A.24: PSB-VVER Test-5a FRSB Top Third Cladding Temperature 
111

Fig. A.25: PSB-VVER Test-5a FRSB Peak Cladding Temperature 

113

APPENDIX B 

List of Figures 

Fig. B.1 PSB-VVER Test-5a measured data: pressures in upper plenum (UPP), 

pressurizer (PRZ), accumulator (ACC), steam generator (SG), core power (N) 

and primary circuit mass (PCM). 

Fig. B.2 PSB-VVER Test-5a measured data: maximum cladding temperature (maxT), 

primary coolant mass (PCM) and integrated break flow (IBF). 

Fig. B.3 Uncertainty analysis for the maximum cladding temperature (maxT). 

Fig. B.4 Uncertainty analysis for the upper plenum pressure (UPP). 

Fig. B.5 Uncertainty analysis for the primary coolant mass (PCM). 

Fig. B.6 Uncertainty analysis for the integrated break flow (IBF). 

Fig. B.7 Influence on the accumulator injection time (tacc) per uncertain input parameter. 

Fig. B.8 Influence on the cladding first overheating time (t1) per uncertain input parameter. 

Fig. B.9 Influence on the cladding first overheating time interval (t2 - t1) per uncertain input 

parameter. 

Fig. B.10 Influence on the maximum peak cladding temperature time (tmax) per uncertain 

input parameter. 

Fig. B.11 Influence on the complete core quenching time (t3) per uncertain input parameter. 

Fig. B.12 Influence on the maximum peak cladding temperature maxT (°C) per uncertain 

input parameter. 

Fig. B.13 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 1 to 4: Spearman’s correlation coefficients. 

Fig. B.14 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 5 to 9: Spearman’s correlation coefficients. 

Fig. B.15 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 10 to 13: Spearman’s correlation coefficients. 

Fig. B.16 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 14 to 17: Spearman’s correlation coefficients. 

Fig. B.17 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 18 to 21: Spearman’s correlation coefficients. 

Fig. B.18 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 22 to 24: Spearman’s correlation coefficients. 

Fig. B.19 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 25 to 27: Spearman’s correlation coefficients. 

Fig. B.20 Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 28 to 31: Spearman’s correlation coefficients. 

Fig. B.21 Sensitivity results for the upper plenum pressure (UPP) for the parameters 1 to 4: 

Spearman’s correlation coefficients. 

Fig. B.22 Sensitivity results for the upper plenum pressure (UPP) for the parameters 5 to 9: 

Spearman’s correlation coefficients. 
114

Fig. B.23 Sensitivity results for the upper plenum pressure (UPP) for the parameters 

10 to 13: Spearman’s correlation coefficients. 

Fig. B.24 Sensitivity results for the upper plenum pressure (UPP) for the parameters 

14 to 17: Spearman’s correlation coefficients. 

Fig. B.25 Sensitivity results for the upper plenum pressure (UPP) for the parameters 

18 to 21: Spearman’s correlation coefficients. 

Fig. B.26 Sensitivity results for the upper plenum pressure (UPP) for the parameters 

22 to 24: Spearman’s correlation coefficients. 

Fig. B.27 Sensitivity results for the upper plenum pressure (UPP) for the parameters 

25 to 27: Spearman’s correlation coefficients. 

Fig. B.28 Sensitivity results for the upper plenum pressure (UPP) for the parameters 

28 to 31: Spearman’s correlation coefficients. 

Fig. B.29 Sensitivity results for the primary coolant mass (PCM) for the parameters 1 to 4: 

Spearman’s correlation coefficients. 

Fig. B.30 Sensitivity results for the primary coolant mass (PCM) for the parameters 5 to 9: 

Spearman’s correlation coefficients. 

Fig. B.31 Sensitivity results for the primary coolant mass (PCM) for the parameters 

10 to 13: Spearman’s correlation coefficients. 

Fig. B.32 Sensitivity results for the primary coolant mass (PCM) for the parameters 

14 to 17: Spearman’s correlation coefficients. 

Fig. B.33 Sensitivity results for the primary coolant mass (PCM) for the parameters 

18 to 21: Spearman’s correlation coefficients. 

Fig. B.34 Sensitivity results for the primary coolant mass (PCM) for the parameters 

22 to 24: Spearman’s correlation coefficients. 

Fig. B.35 Sensitivity results for the primary coolant mass (PCM) for the parameters 

25 to 27: Spearman’s correlation coefficients. 

Fig. B.36 Sensitivity results for the primary coolant mass (PCM) for the parameters 

28 to 31: Spearman’s correlation coefficients. 

Fig. B.37 Sensitivity results for the integrated break flow (IBF) for the parameters 1 to 4: 

Spearman’s correlation coefficients. 

Fig. B.38 Sensitivity results for the integrated break flow (IBF) for the parameters 5 to 9: 

Spearman’s correlation coefficients. 

Fig. B.39 Sensitivity results for the integrated break flow (IBF) for the parameters 10 to 13: 

Spearman’s correlation coefficients. 

Fig. B.40 Sensitivity results for the integrated break flow (IBF) for the parameters 14 to 17: 

Spearman’s correlation coefficients. 

Fig. B.41 Sensitivity results for the integrated break flow (IBF) for the parameters 18 to 21: 

Spearman’s correlation coefficients. 

Fig. B.42 Sensitivity results for the integrated break flow (IBF) for the parameters 22 to 24: 

Spearman’s correlation coefficients. 

Fig. B.43 Sensitivity results for the integrated break flow (IBF) for the parameters 25 to 27: 

Spearman’s correlation coefficients. 

Fig. B.44 Sensitivity results for the integrated break flow (IBF) for the parameters 28 to 31: 

Spearman’s correlation coefficients. 

Fig. B.45 Sensitivity code results (maxT, UPP, PCM and IBF) for the break discharge 

coefficient (parameter 2): Spearman’s correlation coefficients. 
115

Fig. B.46 Sensitivity code results (maxT, UPP, PCM and IBF) for the thermal 

nonequilibrium constant (parameter 3): Spearman’s correlation coefficients. 

Fig. B.47 Sensitivity code results (maxT, UPP, PCM and IBF) for the initial core power 

(parameter 5): Spearman’s correlation coefficients. 

Fig. B.48 Sensitivity code results (maxT, UPP, PCM and IBF) for the cladding specific heat 

capacity (parameter 7): Spearman’s correlation coefficients. 

Fig. B.49 Sensitivity code results (maxT, UPP, PCM and IBF) for the 2-phase degradation 

input of the pump (parameter 10): Spearman’s correlation coefficients. 

Fig. B.50 Sensitivity code results (maxT, UPP, PCM and IBF) for the heat transfer 

coefficient on outside surfaces of piping (parameter 13): Spearman’s correlation 

coefficients. 

Fig. B.51 Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL upper core 

plate: c of Wallis correlation (parameter 14): Spearman’s correlation coefficients. 

Fig. B.52 Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL upper plenum:

c of Kutateladze correlation (parameter 15): Spearman’s correlation coefficients. 

Fig. B.53 Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL steam 

generator tubes inlet: c of Kutateladze correlation (parameter 16): Spearman’s 

correlation coefficients. 

Fig. B.54 Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL downcomer: 

c of Wallis correlation (parameter 17): Spearman’s correlation coefficients. 

Fig. B.55 Sensitivity code results (maxT, UPP, PCM and IBF) for the accumulator initial 

liquid level (parameter 18): Spearman’s correlation coefficients. 

Fig. B.56 Sensitivity code results (maxT, UPP, PCM and IBF) for the friction form loss in 

the accumulator line (parameter 19): Spearman’s correlation coefficients. 

Fig. B.57 Sensitivity code results (maxT, UPP, PCM and IBF) for the pressurizer initial level 

(parameter 22): Spearman’s correlation coefficients. 

Fig. B.58 Sensitivity code results (maxT, UPP, PCM and IBF) for the initial pressure on 

steam generator secondary side (parameter 26): Spearman’s correlation 

coefficients. 

Fig. B.59 Sensitivity code results (maxT, UPP, PCM and IBF) for the liquid injection flow 

variation (HPSI/LPSI) (parameter 28): Spearman’s correlation coefficients. 

Fig. B.60 Sensitivity code results (maxT, UPP, PCM and IBF) for the liquid injections 

temperature (parameter 29): Spearman’s correlation coefficients. 

Fig. B.61 Sensitivity code results (maxT, UPP, PCM and IBF) for the cold leg discharge 

pressure (parameter 30): Spearman’s correlation coefficients. 
116

Fig. B.1 PSB-VVER Test-5a measured data: pressures in upper plenum (UPP), pressurizer 

(PRZ), accumulator (ACC), steam generator (SG), core power(N) and 

primary circuit mass (PCM). 

Pressure (MPa): UPP, PRZ, ACC, SG; 

Power: N(%); Primary Coolant Mass: PCM(%)

0

2

4

6

8

10

12

14

16

18

20

22

-10 0 10 20 30 40 50 60 70 80 90 100

time (s)

pressure (MPa)

0%

20%

40%

60%

80%

100%

percentage (-)

UPP

PRZ

ACC

SG

N(%)

PCM(%)

Pressure (MPa): UPP, PRZ, ACC, SG; 

Power: N(%); Primary Coolant Mass: PCM(%)

0

1

2

3

4

5

6

7

8

9

10

100 300 500 700 900 1100 1300 1500

time (s)

pressure (MPa)

0%

5%

10%

15%

20%

25%

30%

35%

40%

45%

50%

percentage (-)

UPP

PRZ

ACC

SG

N(%)

PCM(%)
117

Fig. B.2. PSB-VVER Test-5a measured data: maximum cladding temperature (maxT), 

primary coolant mass (PCM) and integrated break flow (IBF). 

Fig. B.3. Uncertainty analysis for the maximum cladding temperature (maxT). 

maxT: Maximum cladding temperature

100

200

300

400

500

600

0 200 400 600 800 1000 1200 1400

time (s)

temperature (°C)

experimental value reference calculation low er uncertainty bound upper uncertainty bound

maxT(°C): Maximum cladding temperature

IBF: Integrated break flow

PCM: Primary coolant mass

0

100

200

300

400

500

-50 200 450 700 950 1200 1450

time (s)

tacc=11.56 s; t1=193.5 s; tmax=440.6 s; t2=590.2 s; t3=1300.5 s 

maxT (°C)

0

1

2

3

4

5

IBF/PCM(t=0) PCM/PCM(t=0)

maxT(°C)

IBF/PCM(t=0)

PCM/PCM(t=0)
118

UPP: Upper plenum pressure

0

2

4

6

8

10

12

14

16

-10 0 10 20 30 40 50 60 70 80 90 100

time (s)

pressure (MPa)

experimental value reference calculation low er uncertainty bound upper uncertainty bound

UPP: Upper plenum pressure

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1.0

0 200 400 600 800 1000 1200 1400

time (s)

pressure (MPa)

experimental value reference calculation low er uncertainty bound upper uncertainty bound

Fig. B.4. Uncertainty analysis for the upper plenum pressure (UPP). 
119

Fig. B.5. Uncertainty analysis for the primary coolant mass (PCM). 

Fig. B.6. Uncertainty analysis for theintegrated break flow (IBF). 

PCM: Primary coolant mass

0

100

200

300

400

500

600

700

800

900

0 200 400 600 800 1000 1200 1400

time (s)

mass (kg)

experimental value reference calculation low er uncertainty bound upper uncertainty bound

IBF: Integrated break flow

0

500

1000

1500

2000

2500

3000

3500

0 200 400 600 800 1000 1200 1400

time (s)

mass (kg)

experimental value reference calculation low er uncertainty bound upper uncertainty bound
120

Fig. B.7. Influence on the accumulator injection time (tacc) per uncertain input parameter. 

Fig. B.8. Influence on the cladding first overheating time (t1) per uncertain input 

parameter. 

Sensitivity analysis for the time of start of accumulator injection

 (tacc)

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31

parameter number

Spearman correlation coefficient

Sensitivity analysis for the time of beginning of first cladding heatup 

(t1)

-0.4

-0.2

0.0

0.2

0.4

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31

parameter number

Spearman correlation coefficient
121

Fig. B.9. Influence on the cladding first overheating time interval (t2 - t1) per uncertain 

input parameter. 

Fig. B.10. Influence on the maximum peak cladding temperature time (tmax) per uncertain 

input parameter. 

Sensitivity analysis for the time interval of cladding overheating

 (t2- t1)

-0.6

-0.4

-0.2

0.0

0.2

0.4

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31

parameter number

Spearman correlation coefficient

Sensitivity analysis for the time of maximum cladding temperature 

(tmax)

-0.4

-0.2

0.0

0.2

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31

parameter number

Spearman correlation coefficient
122

Fig. B.11. Influence on the complete core quenching time (t3) per uncertain input 

parameter. 

Fig. B.12. Influence on the maximum peak cladding temperature maxT per uncertain 

input parameter. 

Sensitivity analysis for the time of beginning of last cladding heatup 

(t3)

-0.4

-0.2

0.0

0.2

0.4

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31

parameter number

Spearman correlation coefficient

Sensitivity analysis for the maximum cladding temperature 

(maxT)

-0.6

-0.4

-0.2

0.0

0.2

0.4

1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31

parameter number

Spearman correlation coefficient
123

Fig. B.13. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 1 to 4: Spearman’s correlation coefficients. 

Fig. B.14. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 5 to 9: Spearman’s correlation coefficients. 

maxT: sensitivities of 1 to 4

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

1

2

3

4

maxT: sensitivities of 5 to 9

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

5

6

7

8

9
124

Fig. B.15. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 10 to 13: Spearman’s correlation coefficients. 

Fig. B.16. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 14 to 17: Spearman’s correlation coefficients. 

maxT: sensitivities of 10 to 13

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

10

11

12

13

maxT: sensitivities of 14 to 17

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

14

15

16

17
125

Fig. B.17. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 18 to 21: Spearman’s correlation coefficients. 

Fig. B.18. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 22 to 24: Spearman’s correlation coefficients. 

maxT: sensitivities of 18 to 21

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

18

19

20

21

maxT: sensitivities of 22 to 24

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

22

23

24
126

Fig. B.19. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 25 to 27: Spearman’s correlation coefficients. 

Fig. B.20. Sensitivity results for the maximum peak cladding temperature (maxT) for the 

parameters 28 to 31: Spearman’s correlation coefficients. 

maxT: sensitivities of 25 to 27

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

25

26

27

31

maxT: sensitivities of 28 to 31

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

28

29

30

31
127

Fig. B.21. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

1 to 4: Spearman’s correlation coefficients. 

Fig. B.22. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

5 to 9: Spearman’s correlation coefficients. 

UPP: sensitivities of 1 to 4

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

1

2

3

4

UPP: sensitivities of 5 to 9

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

5

6

7

8

9
128

Fig. B.23. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

10 to 13: Spearman’s correlation coefficients. 

Fig. B.24. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

14 to 17: Spearman’s correlation coefficients. 

UPP: sensitivities of 10 to 13

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

10

11

12

13

UPP: sensitivities of 14 to 17

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

14

15

16

17
129

Fig. B.25. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

18 to 21: Spearman’s correlation coefficients. 

Fig. B.26. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

22 to 24: Spearman’s correlation coefficients. 

UPP: sensitivities of 18 to 21

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

18

19

20

21

UPP: sensitivities of 22 to 24

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

22

23

24
130

Fig. B.27. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

25 to 27: Spearman’s correlation coefficients. 

Fig. B.28. Sensitivity results for the upper plenum pressure (UPP) for the parameters 

28 to 31: Spearman’s correlation coefficients. 

UPP: sensitivities of 28 to 31

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

28

29

30

31

UPP sensitivities of 25 to 27

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

25

26

27
131

Fig. B.29. Sensitivity results for the primary coolant mass (PCM) for the parameters 

1 to 4: Spearman’s correlation coefficients. 

Fig. B.30. Sensitivity results for the primary coolant mass (PCM) for the parameters 

5 to 9: Spearman’s correlation coefficients. 

PCM: sensitivities of 1 to 4

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

1

2

3

4

PCM: sensitivities of 5 to 9

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

5

6

7

8

9
132

Fig. B.31. Sensitivity results for the primary coolant mass (PCM) for the parameters 

10 to 13: Spearman’s correlation coefficients. 

Fig. B.32. Sensitivity results for the primary coolant mass (PCM) for the parameters 

14 to 17: Spearman’s correlation coefficients. 

PCM: sensitivities of 10 to 13

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

10

11

12

13

PCM: sensitivities of 14 to 17

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

14

15

16

17
133

Fig. B.33. Sensitivity results for the primary coolant mass (PCM) for the parameters 

18 to 21: Spearman’s correlation coefficients. 

Fig. B.34. Sensitivity results for the primary coolant mass (PCM) for the parameters 

22 to 24: Spearman’s correlation coefficients. 

PCM: sensitivities of 18 to 21

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

18

19

20

21

PCM: sensitivities of 22 to 24

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

1.0

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

22

23

24
134

Fig. B.35. Sensitivity results for the primary coolant mass (PCM) for the parameters 

25 to 27: Spearman’s correlation coefficients. 

Fig. B.36. Sensitivity results for the primary coolant mass (PCM) for the parameters 

28 to 31: Spearman’s correlation coefficients. 

PCM sensitivities of 28 to 31

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

28

29

30

31

PCM sensitivities of 25 to 27

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

25

26

27
135

Fig. B.37. Sensitivity results for the integrated break flow (IBF) for the parameters 1 to 4: 

Spearman’s correlation coefficients. 

Fig. B.38. Sensitivity results for the integrated break flow (IBF) for the parameters 5 to 9: 

Spearman’s correlation coefficients. 

IBF: sensitivities of 1 to 4

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

1.0

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

1

2

3

4

IBF: sensitivities of 5 to 9

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

5

6

7

8

9
136

Fig. B.39. Sensitivity results for the integrated break flow (IBF) for the parameters 

10 to 13: Spearman’s correlation coefficients. 

Fig. B.40. Sensitivity results for the integrated break flow (IBF) for the parameters 

14 to 17: Spearman’s correlation coefficients. 

IBF: sensitivities of 14 to 17

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

14

15

16

17

IBF: sensitivities of 10 to 13

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

10

11

12

13
137

Fig. B.41. Sensitivity results for the integrated break flow (IBF) for the parameters 

18 to 21: Spearman’s correlation coefficients. 

Fig. B.42. Sensitivity results for the integrated break flow (IBF) for the parameters 

22 to 24: Spearman’s correlation coefficients. 

IBF: sensitivities of 18 to 21

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

18

19

20

21

IBF: sensitivities of 22 to 24

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

22

23

24
138

Fig. B.43. Sensitivity results for the integrated break flow (IBF) for the parameters 

25 to 27: Spearman’s correlation coefficients. 

Fig. B.44. Sensitivity results for the integrated break flow (IBF) for the parameters 

28 to 31: Spearman’s correlation coefficients. 

IBF: sensitivities of 28 to 31

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

28

29

30

31

IBF: sensitivities of 25 to 27

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

25

26

27
139

Fig. B.45. Sensitivity code results (maxT, UPP, PCM and IBF) for the break discharge 

coefficient (parameter 2): 

Spearman’s correlation coefficients. 

Fig. B.46. Sensitivity code results (maxT, UPP, PCM and IBF) for the thermal 

nonequilibrium constant (parameter 3): 

Spearman’s correlation coefficients. 

parameter 2

-1.0

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

1.0

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parameter 3

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
140

Fig. B.47. Sensitivity code results (maxT, UPP, PCM and IBF) for the initial core power 

(parameter 5): 

Spearman’s correlation coefficients. 

Fig. B.48. Sensitivity code results (maxT, UPP, PCM and IBF) for the cladding specific 

heat capacity (parameter 7): 

Spearman’s correlation coefficients. 

parameter 5

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parameter 7

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
141

Fig. B.49. Sensitivity code results (maxT, UPP, PCM and IBF) for the 2-phase degradation 

input of the pump (parameter 10): 

Spearman’s correlation coefficients. 

Fig. B.50. Sensitivity code results (maxT, UPP, PCM and IBF) for the heat transfer 

coefficient on outside surfaces of piping (parameter 13): 

Spearman’s correlation coefficients. 

parameter 10

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parameter 13

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
142

Fig. B.51. Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL upper core 

plate; c of Wallis correlation (parameter 14): 

Spearman’s correlation coefficients. 

Fig. B.52. Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL upper plenum; 

c of Kutateladze correlation (parameter 15): 

Spearman’s correlation coefficients. 

parameter 14

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parameter 15

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
143

Fig. B.53. Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL steam 

generator tubes inlet; c of Kutateladze correlation (parameter 16): 

Spearman’s correlation coefficients. 

Fig. B.54. Sensitivity code results (maxT, UPP, PCM and IBF) for the CCFL downcomer; c 

of Wallis correlation (parameter 17): 

Spearman’s correlation coefficients. 

parameter 17

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parameter 16

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
144

Fig. B.55. Sensitivity code results (maxT, UPP, PCM and IBF) for the accumulator initial 

liquid level (parameter 18): 

Spearman’s correlation coefficients. 

Fig. B.56. Sensitivity code results (maxT, UPP, PCM and IBF) for the friction form loss in 

the accumulator line (parameter 19): 

Spearman’s correlation coefficients. 

parameter 18

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parametr 19

-0.6

-0.4

-0.2

0.0

0.2

0.4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
145

Fig. B.57. Sensitivity code results (maxT, UPP, PCM and IBF) for the pressurizer initial 

level (parameter 22): 

Spearman’s correlation coefficients. 

Fig. B.58. Sensitivity code results (maxT, UPP, PCM and IBF) for the initial pressure on 

steam generator secondary side (parameter 26): 

Spearman’s correlation coefficients. 

parameter 22

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

1.0

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parameter 26

-0.8

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
146

Fig. B.59. Sensitivity code results (maxT, UPP, PCM and IBF) for the liquid injection flow 

variation (HPSI/LPSI) (parameter 28): 

Spearman’s correlation coefficients. 

Fig. B.60. Sensitivity code results (maxT, UPP, PCM and IBF) for the liquid injections 

temperature (parameter 29): 

Spearman’s correlation coefficients. 

parameter 29

-0,6

-0,4

-0,2

0,0

0,2

0,4

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

parametr 28

-0.4

-0.2

0.0

0.2

0.4

0.6

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF
147

Fig. B.61. Sensitivity code results (maxT, UPP, PCM and IBF) for the cold leg discharge 

pressure (parameter 30): 

Spearman’s correlation coefficients. 

parameter 30

-0.6

-0.4

-0.2

0.0

0.2

0.4

0.6

0.8

0 200 400 600 800 1000 1200 1400

time (s)

sensitivity

maxT

UPP

PCM

IBF

149

APPENDIX C 

List of Figures 

Fig. C.1 PSB-VVER Test-3 Integrated Break Flow 

Fig. C.2 PSB-VVER Test-3 Upper Plenum Pressure 

Fig. C.3 PSB-VVER Test-3 Pressurizer Pressure 

Fig. C.4 PSB-VVER Test-3 Accumulator #2 Pressure 

Fig. C.5 PSB-VVER Test-3 Accumulator #2 Level 

Fig. C.6 PSB-VVER Test-3 Accumulator #4 Pressure 

Fig. C.7 PSB-VVER Test-3 Accumulator #4 Level 

Fig. C.8 PSB-VVER Test-3 Primary Coolant Mass 

Fig. C.9 PSB-VVER Test-3 Steam Generator #1 Pressure 

Fig. C.10 PSB-VVER Test-3 Steam Generator #2 Pressure 

Fig. C.11 PSB-VVER Test-3 Steam Generator #3 Pressure 

Fig. C.12 PSB-VVER Test-3 Steam Generator #4 Pressure 

Fig. C.13 PSB-VVER Test-3 Core Simulator Inlet Coolant Temperature 

Fig. C.14 PSB-VVER Test-3 Core Simulator Outlet Coolant Temperature 

Fig. C.15 PSB-VVER Test-3 Downcomer Simulator (6310 – 2790 mm) Differential 

Pressure 

Fig. C.16 PSB-VVER Test-3 Downcomer Simulator (2790 – 1010 mm) Differential 

Pressure 

Fig. C.17 PSB-VVER Test-3 Lower Plenum Simulator (1010 – 1915 mm) Differential 

Pressure 

Fig. C.18 PSB-VVER Test-3 Core Simulator (5440 – 1915 mm) Differential Pressure 

Fig. C.19 PSB-VVER Test-3 Upper Plenum & Upper Head Simulator (12525 – 5440 mm) 

Differential Pressure 

Fig. C.20 PSB-VVER Test-3 Downcomer Simulator to Upper Plenum Simulator (6870 –

 8650 mm) Differential Pressure 

Fig. C.21 PSB-VVER Test-3 Loop Seal #1 Descending Leg (7150 – 3810 mm) 

Differential Pressure 

Fig. C.22 PSB-VVER Test-3 Loop Seal #1 Ascending Leg (7065 – 4370 mm) Differential 

Pressure 

Fig. C.23 PSB-VVER Test-3 Loop Seal #2 Descending Leg (7345 – 3795 mm) 

Differential Pressure 

Fig. C.24 PSB-VVER Test-3 Loop Seal #2 Ascending Leg (7065 – 4315 mm) Differential 

Pressure 
150

Fig. C.25 PSB-VVER Test-3 Loop Seal #3 Descending Leg (7155 – 3780 mm) 

Differential Pressure 

Fig. C.26 PSB-VVER Test-3 Loop Seal #3 Ascending Leg (7065 – 4300 mm) Differential 

Pressure 

Fig. C.27 PSB-VVER Test-3 Loop Seal #4 Descending Leg (7170 – 3800 mm) 

Differential Pressure 

Fig. C.28 PSB-VVER Test-3 Loop Seal #4 Ascending Leg (7065 – 4350 mm) Differential 

Pressure 

Fig. C.29 PSB-VVER Test-3 FRSB Bottom Third Cladding Temperature 

Fig. C.30 PSB-VVER Test-3 FRSB Middle Third Cladding Temperature 

Fig. C.31 PSB-VVER Test-3 FRSB Top Third Cladding Temperature 

Fig. C.32 PSB-VVER Test-3 FRSB Peak Cladding Temperature 

Fig. C.33 PSB-VVER Test-3 FRSB Peak Cladding Temperature – Sensitivity to Critical 

Discharge Coefficient 
151

Fig. C.1: PSB-VVER Test-3 Integrated Break Flow 
152

Fig. C.2: PSB-VVER Test-3 Upper Plenum Pressure 
153

Fig. C.3: PSB-VVER Test-3 Pressurizer Pressure 
154

Fig. C.4: PSB-VVER Test-3 Accumulator #2 Pressure 
155

Fig. C.5: PSB-VVER Test-3 Accumulator #2 Level 
156

Fig. C.6: PSB-VVER Test-3 Accumulator #4 Pressure 
157

Fig. C.7: PSB-VVER Test-3 Accumulator #4 Level 
158

Fig. C.8: PSB-VVER Test-3 Primary Coolant Mass 
159

Fig. C.9: PSB-VVER Test-3 Steam Generator #1 Pressure 
160

Fig. C.10: PSB-VVER Test-3 Steam Generator #2 Pressure 
161

Fig. C.11: PSB-VVER Test-3 Steam Generator #3 Pressure 
162

Fig. C.12: PSB-VVER Test-3 Steam Generator #4 Pressure 
163

Fig. C.13: PSB-VVER Test-3 Core Simulator Inlet Coolant Temperature 
164

Fig. C.14: PSB-VVER Test-3 Core Simulator Outlet Coolant Temperature 
165

Fig. C.15: PSB-VVER Test-3 Downcomer Simulator (6310 – 2790 mm) Differential 

Pressure 
166

Fig. C.16: PSB-VVER Test-3 Downcomer Simulator (2790 – 1010 mm) Differential 

Pressure 
167

Fig. C.17: PSB-VVER Test-3 Lower Plenum Simulator (1010 – 1915 mm) Differential 

Pressure 
168

Fig. C.18: PSB-VVER Test-3 Core Simulator (5440 – 1915 mm) Differential Pressure 
169

Fig. C.19: PSB-VVER Test-3 Upper Plenum & Upper Head Simulator (12525 –

 5440 mm) Differential Pressure 
170

Fig. C.20: PSB-VVER Test-3 Downcomer Simulator to Upper Plenum Simulator (6870 –

 8650 mm) Differential Pressure 
171

Fig. C.21: PSB-VVER Test-3 Loop Seal #1 Descending Leg (7150 – 3810 mm) 

Differential Pressure 
172

Fig. C.22: PSB-VVER Test-3 Loop Seal #1 Ascending Leg (7065 – 4370 mm) 

Differential Pressure 
173

Fig. C.23: PSB-VVER Test-3 Loop Seal #2 Descending Leg (7345 – 3795 mm) 

Differential Pressure 
174

Fig. C.24: PSB-VVER Test-3 Loop Seal #2 Ascending Leg (7065 – 4315 mm) 

Differential Pressure 
175

Fig. C.25: PSB-VVER Test-3 Loop Seal #3 Descending Leg (7155 – 3780 mm) 

Differential Pressure 
176

Fig. C.26: PSB-VVER Test-3 Loop Seal #3 Ascending Leg (7065 – 4300 mm) 

Differential Pressure 
177

Fig. C.27: PSB-VVER Test-3 Loop Seal #4 Descending Leg (7170 – 3800 mm) 

Differential Pressure 
178

Fig. C.28: PSB-VVER Test-3 Loop Seal #4 Ascending Leg (7065 – 4350 mm) 

Differential Pressure 
179

Fig. C.29: PSB-VVER Test-3 FRSB Bottom Third Cladding Temperature 
180

Fig. C.30: PSB-VVER Test-3 FRSB Middle Third Cladding Temperature 
181

Fig. C.31: PSB-VVER Test-3 FRSB Top Third Cladding Temperature 
182

Fig. C.32: PSB-VVER Test-3 FRSB Peak Cladding Temperature 
183

Fig. C.33: PSB-VVER Test-3 FRSB Peak Cladding Temperature – Sensitivity to Critical 

Discharge Coefficient 




UNITED STATES 

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OFFICIAL BUSINESS
NUREG/IA-0435 Assessment of RELAP5/MOD3.3 and TRACE V5.0 Computer Codes 

Against LOCA Test Data from PSB-VVER Test Facility