NUREG/IA-0479 

Prepared by: 

Andrej Prošek 

Jožef Stefan Institute

Jamova cesta 39 

Sl-1000 Ljubljana, 

Slovenia

Kirk Tien, NRC Project Manager 

Division of Systems Analysis 

Office of Nuclear Regulatory 

Research U.S. Nuclear Regulatory 

Commission Washington, DC 

20555-0001 

Manuscript Completed: November 2016

Date Published: June 2017

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by 

U.S. Nuclear Regulatory Commission

RELAP5 and TRACE Calculations of 

LOCA in PWR 

NUREG/IA-0479

Prepared by: 

Andrej Prošek 

Jožef Stefan Institute

Jamova cesta 39 

Sl-1000 Ljubljana, 

Slovenia

Kirk Tien, NRC Project Manager 

Division of Systems Analysis 

Office of Nuclear Regulatory 

Research U.S. Nuclear Regulatory

Commission Washington, DC 

20555-0001 

Manuscript Completed: November 2016

Date Published: June 2017

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by 

U.S. Nuclear Regulatory Commission

RELAP5 and TRACE Calculations of 

LOCA in PWR 

iii

ABSTRACT

The accident at the Fukushima Dai-ichi nuclear power plant in 2011 demonstrated that external 

events could cause loss of all safety systems. In the Europe stress tests were performed and 

the need was identified to further improve the safety of the existing operating reactors. 

Therefore the safety upgrade programs were started. The objective of this study was to 

demonstrate that developed input model of two-loop pressurized water reactor (PWR) for 

TRACE thermal-hydraulic systems code can be used for independent calculations to be 

compared with RELAP5 computer code calculations. For demonstration the response of PWR 

to loss-of-coolant accident (LOCA) break spectrum from 10.16 cm (4 inch) to 30.48 cm (12 inch) 

was simulated. Only passive accumulators were assumed available. For calculations the latest 

TRACE Version 5.0 Patch 4 and RELAP5/MOD3.3 Patch 4 using both break flow models were 

used. The results showed that RELAP5 calculations using different break flow models are rather 

similar, therefore also other parameters are similar. The accumulators discharge was faster in 

TRACE calculation than in RELAP5 calculations. It can be concluded that different accumulator 

discharge influencing the break flow seems to be the largest contributor to the differences in the 

results between RELAP5 and TRACE.

v 

CONTENTS

Page

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

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

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

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

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

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

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

2. METHODS USED............................................................................................................... 3

2.1 LOCA Scenario Description......................................................................................... 3

2.2 Computer Codes Used ................................................................................................ 3

2.3 RELAP5 Input Model ................................................................................................... 3

2.4 TRACE Input Model..................................................................................................... 4

2.5 Initial and Boundary Conditions ................................................................................... 5

2.6 Simulated LOCA Break Cases .................................................................................... 7

3. RESULTS........................................................................................................................... 9

3.1 LOCA with 10.16 cm Break Size.................................................................................. 9

3.2 LOCA with 12.7 cm Break Size ..................................................................................13

3.3 LOCA with 15.24 cm Break Size.................................................................................18

3.4 LOCA with 20.32 cm Break Size.................................................................................22

3.5 LOCA with 30.48 cm Break Size.................................................................................26

3.6 Discussion of Results .................................................................................................31

4. CONCLUSIONS................................................................................................................33

5. REFERENCES..................................................................................................................35

vii

FIGURES

Page

Figure 1 RELAP5 Krško NPP Hydraulic Components View . .................................................. 4

Figure 2 TRACE Krško NPP Hydraulic Components View ...................................................... 5

Figure 3 Pressurizer Pressure (10.16 cm)............................................................................... 9

Figure 4 Cold Leg no. 1 Temperature (10.16 cm)...................................................................10

Figure 5 Hot Leg no. 1 Temperature (10.16 cm) ....................................................................10

Figure 6 Break Flow (10.16 cm).............................................................................................10

Figure 7 Core Collapsed Liquid Level (10.16 cm)...................................................................11

Figure 8 Core Exit Temperature (10.16 cm)...........................................................................11

Figure 9 Fuel Cladding Temperature (10.16 cm)....................................................................11

Figure 10 RCS Mass (10.16 cm)..............................................................................................12

Figure 11 Integrated Break Flow (10.16 cm) ............................................................................12

Figure 12 Mass Injected by Accumulators (10.16 cm)..............................................................12

Figure 13 Accumulator no. 1 Flow (10.16 cm)..........................................................................13

Figure 14 Accumulator no. 2 Flow (10.16 cm)..........................................................................13

Figure 15 Pressurizer Pressure (12.7 cm)................................................................................14

Figure 16 Cold Leg no. 1 Temperature (12.7 cm).....................................................................14

Figure 17 Hot Leg no. 1 Temperature (12.7 cm) ......................................................................14

Figure 18 Break Flow (12.7 cm)...............................................................................................15

Figure 19 Core Collapsed Liquid Level (12.7 cm).....................................................................15

Figure 20 Core Exit Temperature (12.7 cm).............................................................................15

Figure 21 Fuel Cladding Temperature (12.7 cm)......................................................................16

Figure 22 RCS Mass (12.7 cm)................................................................................................16

Figure 23 Integrated Break Flow (12.7 cm) ..............................................................................16

Figure 24 Mass Injected by Accumulators (12.7 cm)................................................................17

Figure 25 Accumulator no. 1 Flow (12.7 cm)............................................................................17

Figure 26 Accumulator no. 2 Flow (12.7 cm)............................................................................17

Figure 27 Pressurizer Pressure (15.24 cm)..............................................................................18

Figure 28 Cold Leg no. 1 Temperature (15.24 cm)...................................................................18

Figure 29 Hot Leg no. 1 Temperature (15.24 cm) ....................................................................19

Figure 30 Break Flow (15.24 cm).............................................................................................19

Figure 31 Core Collapsed Liquid Level (15.24 cm)...................................................................19

Figure 32 Core Exit Temperature (15.24 cm)...........................................................................20

Figure 33 Fuel Cladding Temperature (15.24 cm)....................................................................20

Figure 34 RCS Mass (15.24 cm)..............................................................................................20

Figure 35 Integrated Break Flow (15.24 cm) ............................................................................21

Figure 36 Mass Injected by Accumulators (15.24 cm)..............................................................21

Figure 37 Accumulator no. 1 Flow (15.24 cm)..........................................................................21

Figure 38 Accumulator no. 2 Flow (15.24 cm)..........................................................................22

Figure 39 Pressurizer Pressure (20.32 cm)..............................................................................22

Figure 40 Cold Leg no. 1 Temperature (20.32 cm)...................................................................23

Figure 41 Hot Leg no. 1 Temperature (20.32 cm) ....................................................................23

Figure 42 Break Flow (20.32 cm).............................................................................................23

Figure 43 Core Collapsed Liquid Level (20.32 cm)...................................................................24

Figure 44 Core Exit Temperature (20.32 cm)...........................................................................24

Figure 45 Fuel Cladding Temperature (20.32 cm)....................................................................24

Figure 46 RCS Mass (20.32 cm)..............................................................................................25

Figure 47 Integrated Break Flow (20.32 cm) ............................................................................25
viii

Figure 48 Mass Injected by Accumulators (20.32 cm)..............................................................25

Figure 49 Accumulator no. 1 Flow (20.32 cm)..........................................................................26

Figure 50 Accumulator no. 2 Flow (20.32 cm)..........................................................................26

Figure 51 Pressurizer Pressure (30.48 cm)..............................................................................27

Figure 52 Cold Leg no. 1 Temperature (30.48 cm)...................................................................27

Figure 53 Hot Leg no. 1 Temperature (30.48 cm) ....................................................................28

Figure 54 Break Flow (30.48 cm).............................................................................................28

Figure 55 Core Collapsed Liquid Level (30.48 cm)...................................................................28

Figure 56 Core Exit Temperature (30.48 cm)...........................................................................29

Figure 57 Fuel Cladding Temperature (30.48 cm)....................................................................29

Figure 58 RCS Mass (30.48 cm)..............................................................................................29

Figure 59 Integrated Break Flow (30.48 cm) ............................................................................30

Figure 60 Mass Injected by Accumulators (30.48 cm)..............................................................30

Figure 61 Accumulator no. 1 Flow (30.48 cm)..........................................................................30

Figure 62 Accumulator no. 2 Flow (30.48 cm)..........................................................................31

Figure 63 Mass Discharged by Accumulators (Ref. 7)..............................................................32

Figure 64 Integrated Accumulators Injection Flow (Fig. C.8-65 from Ref. 12) ...........................32
ix

TABLES

Page

Table 1 Initial and Boundary Conditions.................................................................................... 6

Table 2 LOCA Scenario Cases Simulated with RELAP5 and TRACE....................................... 7

xi

EXECUTIVE SUMMARY

The accident at the Fukushima Dai-ichi nuclear power plant in 2011 demonstrated that external 

events could cause loss of all safety systems. In the Europe stress tests were performed and 

the need was identified to further improve the safety of the existing operating reactors. 

Therefore the safety upgrade programs were started. The objective of this study was to 

demonstrate that developed input model of two-loop pressurized water reactor (PWR) for 

TRACE thermal-hydraulic systems code has the capability for independent assessment of 

RELAP5 computer code calculations. Namely, in the frame of the safety upgrade program the 

RELAP5 analyses have been performed by the plant to determine pressure and flow 

requirements for alternative safety injection pump for design extension conditions A (DEC-A) 

loss of coolant accidents (LOCA). For demonstration the response of PWR to loss-of-coolant 

accident (LOCA) was simulated. The break spectrum consists of 30.48 cm (12 inch), 20.32 cm 

(8 inch), 15.24 cm (6 inch), 12.7 cm (5 inch) and 10.16 cm (4 inch) equivalent diameter cold leg 

breaks. The initiating event was opening of the valve simulating the break. The reactor trip on 

(compensated) low pressurizer pressure (12.99 MPa) further caused the turbine trip. The safety 

injection (SI) signal was generated on the low-low pressurizer pressure signal at 12.27 MPa. On 

SI signal no active safety systems started (e.g. high pressure safety injection pumps and low 

pressure safety injection pumps and motor driven auxiliary feedwater pumps). Only passive 

components were assumed available, i.e. accumulators. All these LOCA scenarios with above 

assumptions lead to the core heatup. In this way the time available before significant core 

heatup could be obtained.

For calculations the latest TRACE Version 5.0 Patch 4 using extension of Ransom and Trapp 

critical flow model (default) and RELAP5/MOD3.3 Patch 4 using Henry-Fauske critical flow 

model (default) and Ransom-Trapp critical flow model (Option 50) were used.

The results showed that RELAP5 calculations using different break flow models are rather 

similar, therefore also other parameters are similar. When comparing TRACE results to 

RELAP5 results, the accumulators discharge was consistently faster in TRACE calculations 

than in RELAP5 calculations. Therefore the calculated TRACE break flow was also larger than 

RELAP5 calculated break flow during this period. This further leads to qualitative differences at 

the 30.48 cm break size scenario. It can be concluded that the different accumulator discharge 

influencing the break flow seems to be the largest contributor to the differences in the results 

between RELAP5 and TRACE.

xiii

ACKNOWLEDGMENTS 

The authors acknowledge the financial support from Krško Nuclear Power Plant and Slovenian 

Nuclear Safety Administration within CAMP program (project no. POG-U3-KE-R4/104/12 – also 

NEK no.: 3120118) and from the state budget by the Slovenian Research Agency (program 

no. P2-0026).

xv 

ABBREVIATIONS

ACC accumulator

AFW auxiliary feedwater

ASCII American Standard Code for Information Interchange

CVCS chemical and volume control system

DEC-A design extension conditions A

ECCS emergency core cooling system

HPSI high pressure safety injection

LOCA loss-of-coolant accident

LPSI low pressure safety injection

MFW main feedwater

MSIV main steam isolation valve

NPP nuclear power plant

PORV power operated relief valve

PRZ pressurizer

PWR Pressurized Water Reactor

RCP reactor coolant pump

RCS reactor coolant system

RELAP Reactor Excursion and Leak Analysis Program

RPV reactor pressure vessel

SG steam generator

SI safety injection

SL surge line

SNAP Symbolic Nuclear Analysis Package

WR wide range

1 

1. INTRODUCTION

Slovenian Krško nuclear power plant is a one unit plant with pressurized water reactor (PWR), a 

two-loop Westinghouse design with thermal power 1994 MW. In the frame of Krško Safety 

Upgrade Program the RELAP5 calculations have been also used to define requirements for 

alternative safety injection pump. To support independent assessment of Krško Safety Upgrade 

Program there was a need to make code comparison, therefore TRACE code has been 

proposed for comparison calculations.

To define requirements for safety injection pump, loss of coolant accidents (LOCA) were 

simulated in a two-loop PWR by Krško nuclear power plant (NPP). Therefore in this study 

independent analyses of loss-of-coolant accident (LOCA) break spectrum by RELAP5 and 

TRACE computer codes have been performed for comparison purposes. In Section 2 first the 

LOCA scenarios are described. Then the RELAP5 and TRACE thermal-hydraulic system 

computer codes are briefly described, followed by input model description for both computer 

codes. Then the initial and boundary conditions, resulting from steady state calculations are 

presented. Five break sizes ranging from 10.16 cm to 30.48 cm equivalent diameter break size 

in cold leg were simulated and for each break size three calculations were performed, one with 

TRACE using default option for critical flow model and two with RELAP5 using Henry-Fauske 

and Ransom-Trapp critical flow models. Then, results of the LOCA calculations are presented in

Section 3, including discussion of the result. Finally, main conclusions are drawn.

3 

2. METHODS USED

2.1 LOCA Scenario Description

In the LOCAs simulated at the beginning of transient only passive components were assumed 

available: two accumulators, pressurizer safety valves (not needed during LOCAs), and steam 

generator safety valves (not needed during LOCAs). All the LOCA scenarios simulated with 

above assumptions lead to the core heatup.

The initiating event is opening of the valve simulating the break in the cold leg with reactor 

operating at 100% power. The reactor trip on (compensated) low pressurizer pressure (12.99 

MPa) further causes the turbine trip. The safety injection (SI) signal is generated on the low-low 

pressurizer pressure signal at 12.27 MPa. On SI signal no active safety systems start (e.g. high 

pressure safety injection (HPSI) pumps and low pressure safety injection (LPSI) pumps and 

motor driven (MD) AFW pumps). When pressure drops below 49.55 bars, both accumulators 

start to inject. Larger is the break size, faster is the accumulator discharge. When both 

accumulators are emptied, the reactor coolant system (RCS) mass inventory is again 

decreasing, resulting in core uncovering. The core starts to heat up and the calculations are 

terminated at 2100 K, if calculation is not aborted earlier.

2.2 Computer Codes Used

At the time of calculations the latest RELAP5 and TRACE thermal hydraulic system codes were 

used: U.S. NRC RELAP5/MOD3.3 Patch 4 (Ref. 1) and TRACE Version 5.0 Patch 4 (Ref. 2), 

respectively. In June 2016 new RELAP5/MOD3.3 Patch 5 (Ref. 3) was released as a result of 

maintenance of the code, without any new critical flow models. The RELAP5/MOD3.3 Patch 4 

has built in two models for critical flow: Henry-Fauske critical flow model which is default and 

Ransom-Trapp critical flow model (Option 50 need to be used). The TRACE has built in as default 

the critical flow model which is extension of Ransom and Trapp critical flow model.

2.3 RELAP5 Input Model

To perform the analyses, the base RELAP5 input model of Krško NPP has been used. Krško 

NPP is a two loop PWR, Westinghouse type, with reactor power uprated to 1994 MW. The input 

model has been validated by plant transients (e.g. Ref. 4). It has been used for several safety 

analyses including reference calculations for Krško full scope simulator verification (Refs. 5 and

6). The base model consists of 469 control volumes, 497 junctions and 378 heat structures with 

2107 radial mesh points. When imported ASCI file into SNAP, the hydraulic components view 

has been generated semi automatically (hydraulic components with connections generated 

automatically, annotations and layout manually). In terms of SNAP this gives 304 hydraulic 

components and 108 heat structures. Hydraulic components in SNAP consist of both volumes 

and junctions, where pipe with more volumes is counted as one component. Each heat structure 

in SNAP connected to pipe is counted as one component in SNAP and not as many heat 

structures as pipe volumes like counted in RELAP5 output file. This explains the difference in 

numbers of heat structures in Figure 1 and that reported in RELAP5 output file.
4 

Figure 1 RELAP5 Krško NPP Hydraulic Components View

Modeling of the primary side includes the reactor pressure vessel (RPV), both loops (LOOP 1 

and 2), the pressurizer (PRZ) vessel, pressurizer surge line (SL), pressurizer spray lines and 

valves, two pressurizer power operated relief valves (PORVs) and two pressurizer safety 

valves, chemical and volume control system (CVCS) charging and letdown flow, and reactor 

coolant pump (RCP) seal flow. Emergency core cooling system (ECCS) piping includes high 

pressure safety injection (HPSI) pumps, accumulators (ACCs), and low pressure safety injection 

(LPSI) pumps. The secondary side consists of the SG secondary side, main steam line, main 

steam isolation valves (MSIVs), SG relief and safety valves, and main feedwater (MFW) piping. 

The turbine valve is modeled by the corresponding logic. The turbine is represented by time 

dependent volume. The MFW and AFW (auxiliary feedwater) pumps are modeled as time 

dependent junctions.

2.4 TRACE Input Model

The one-dimensional TRACE plant input model was obtained from an existing RELAP5/MOD3.3 

plant input deck (Ref. 9). The conversion of the RELAP5 input model to TRACE input model 

was performed using SNAP (Ref. 8) and following the JSI RELAP5 to TRACE conversion 

method. A detailed description regarding the conversion procedure can be found in Ref. 7. 

Several modifications were manually brought to the TRACE input model during the conversion 
5 

process, mostly related to Heat Structures boundary conditions, Accumulator model option and 

Hydraulic connections of Pipe components that originated from RELAP5 Branch components. 

Several Control Block Data have been modified too. For more details refer to Ref. 10. 

TRACE input model is shown in Figure 2. The number of SNAP hydraulic components is 473 

and the number of heat structures is 108.

Figure 2 TRACE Krško NPP Hydraulic Components View

2.5 Initial and Boundary Conditions

Table 1 shows initial and boundary conditions at the beginning of simulation. Initial values 

and boundary conditions are given for both loops (where applicable). It can be seen that RELAP5 

initial and boundary conditions are a bit closer to reference PWR values than TRACE initial and 

boundary conditions. When looking TRACE values there is significant deviation in steam 

generator levels only due to separator model problems.
6 

Table 1 Initial and Boundary Conditions 

Parameter (unit) Reference PWR RELAP5/MOD3.3 TRACE

Value

Core power (MW) 1994 1994 1994

Pressurizer pressure (MPa) 15.513 15.513 15.512

Pressurizer level (%) 55.7 55.8 55.24

Average RCS temperature no. 1 (K) 578.15 578.15 579.26

Average RCS temperature no. 2 (K) 578.15 578.06 579.33

Cold leg temperature no. 1 (K) 558.75 559.51 561.34

Cold leg temperature no. 2 (K) 558.75 559.32 561.48

Hot leg temperature no. 1 (K) 597.55 596.79 597.22

Hot leg temperature no. 2 (K) 597.55 596.79 597.22

Cold leg flow no. 1 (kg/s) 4694.7 4721.2 4888.9

Cold leg flow no. 2 (kg/s) 4694.7 4719.6 4886.5

Steam generator pressure no. 1 (MPa) 6.281 6.438 6.619

Steam generator pressure no. 2 (MPa) 6.281 6.415 6.635

Steam generator NR level no. 1 (%) 69.3 69.3 6.2

Steam generator NR level no. 2 (%) 69.3 69.3 6.5

Steam flow no. 1 (kg/s) 544.5 541.3 539.9

Steam flow no. 2 (kg/s) 544.5 544.5 532.3

Main feedwater temperature (K) 492.6 492.8 493.7

Namely, besides expected steam flow there appears also liquid mass flow of 84.1 kg/s and 123.7 

kg/s in steam line no. 1 and 2, respectively. The reason seems to be separator model in TRACE 

which needs improvement (bug report has been sent in December 2015). The plant TRACE input 

model of SG may also need some improvements. The problems of separator model to correctly 

separate steam and liquid at all possible boundary conditions caused that artificial steam 

generator level control during steady state calculation could not fill the steam generator levels to 

PWR reference values. As already mentioned, the liquid is flowing also to steam lines. Such 

separator model is serious limitation to perform calculation of any transient occurring on the 

secondary side. In LOCA calculations the influence of the secondary side is typically smaller than 

in the secondary side initiated transients because due to break the primary side empties and the 

natural circulation is terminated. Due to this fact and information that new version of TRACE will 

be released soon it was decided to perform comparison calculations with the current version, 

being aware of TRACE separator model limitation. It was judged that in spite of this deficiency 

LOCA calculations could be performed for larger breaks while at smaller break sizes the influence 

of secondary side on the primary side is expected to be larger and larger.
7 

2.6 Simulated LOCA Break Cases

The breaks simulated were 10.16 cm (4 inch), 12.7 cm (5 inch), 15.24 cm (6 inch), 20.32 

cm (8 inch) and 30.48 cm (12 inch) equivalent diameter cold leg breaks. For each break size three 

simulations were performed, two by RELAP5 and one by TRACE as can be seen from Table 2. 

In case of 30.48 cm break size additional TRACE calculation was performed, in which the TRACE 

accumulator discharge flow was tuned to RELAP5 accumulator discharge flow. In all simulations 

default values for break flows were used.

Table 2 LOCA Scenario Cases Simulated with RELAP5 and TRACE

Break size diameter RELAP5/MOD3.3 

using HF critical flow 

model

RELAP5/MOD3.3 

using RT critical flow 

model

TRACE using 

extended RT critical 

flow model

10.16 cm (4 inch) 4_R5-HF 4_R5-RT 4_TRACE

12.7 cm (5 inch) 5_R5-HF 5_R5-RT 5_TRACE

15.24 cm (6 inch) 6_R5-HF 6_R5-RT 6_TRACE

20.32 cm (8 inch) 8_R5-HF 8_R5-RT 8_TRACE

30.48 cm (12 inch) 12_R5-HF 12_R5-RT 12_TRACE, 

12_TRACE (ACC)*

* - TRACE accumulator discharge flow tuned to RELAP5 accumulator discharge flow

9 

3. RESULTS

The results of LOCA break spectrum calculations are shown in Figures 3 through 62. For each 

break size the following parameters are shown: pressurizer pressure, cold leg no. 1 

temperature, leg no. 1 temperature, break flow, core collapsed liquid level, core exit 

temperature, fuel cladding temperature, RCS mass, integrated break flow, mass injected by 

accumulators, accumulator no. 1 flow and accumulator no. 2 flow. Finally, in the discussion 

section additional Figures 63 and 63 are shown to explain the accumulator discharge flow.

3.1 LOCA with 10.16 cm Break Size

As has been indicated, at 10.16 cm break size (see Figures 3 through 14) there is some 

pressure plateau in primary pressure in RELAP5 calculations. It means that in this period 

secondary side is important. Due to explained limitations of TRACE separator model, the 

secondary side pressure drops and primary pressure follows it. Therefore the time sequence of 

further events in TRACE is faster. It should be noted that in the BETHSY LOCA calculations 

performed in the past (see Ref. 7), the mass discharged through accumulators was also much 

faster in TRACE simulation comparing to RELAP5. Both these facts resulted in earlier core 

heatup comparing to both RELAP5 calculations. Due to faster accumulator discharge the break 

flow significantly changes (increases) in the time period of accumulator discharging, influencing 

further the pressure, RCS mass, core collapsed liquid level, hot and cold leg temperatures, core 

exit temperature and fuel cladding temperature.

On the other hand, the influence of break flow model in case of RELAP5 calculations is more

significant during initial blowdown. Therefore, slightly earlier heatup is predicted by HenryFauske (HF) critical flow model, which is higher in the initial period. 

Figure 3 Pressurizer Pressure (10.16 cm) 
10

Figure 4 Cold Leg no. 1 Temperature (10.16 cm) 

Figure 5 Hot Leg no. 1 Temperature (10.16 cm) 

Figure 6 Break Flow (10.16 cm) 
11

Figure 7 Core Collapsed Liquid Level (10.16 cm) 

Figure 8 Core Exit Temperature (10.16 cm) 

Figure 9 Fuel Cladding Temperature (10.16 cm) 
12

Figure 10 RCS Mass (10.16 cm) 

Figure 11 Integrated Break Flow (10.16 cm) 

Figure 12 Mass Injected by Accumulators (10.16 cm) 
13

Figure 13 Accumulator no. 1 Flow (10.16 cm) 

Figure 14 Accumulator no. 2 Flow (10.16 cm) 

3.2 LOCA with 12.7 cm Break Size

In case of 12.7 cm break size (see Figures 15 through 26) the pressure plateau is shorter and 

also the accumulator influence on RCS mass is smaller, therefore also the difference between 

TRACE and RELAP5 simulations is smaller (see RCS mass, core collapsed liquid level, hot and 

cold leg temperatures, core exit temperature and fuel cladding temperature).

In all calculated case the core heatup started in about 30 minutes.
14

Figure 15 Pressurizer Pressure (12.7 cm) 

Figure 16 Cold Leg no. 1 Temperature (12.7 cm) 

Figure 17 Hot Leg no. 1 Temperature (12.7 cm) 
15

Figure 18 Break Flow (12.7 cm) 

Figure 19 Core Collapsed Liquid Level (12.7 cm) 

Figure 20 Core Exit Temperature (12.7 cm) 
16

Figure 21 Fuel Cladding Temperature (12.7 cm) 

Figure 22 RCS Mass (12.7 cm) 

Figure 23 Integrated Break Flow (12.7 cm) 
17

Figure 24 Mass Injected by Accumulators (12.7 cm) 

Figure 25 Accumulator no. 1 Flow (12.7 cm) 

Figure 26 Accumulator no. 2 Flow (12.7 cm) 
18

3.3 LOCA with 15.24 cm Break Size

In case of 15.24 cm break size (see Figures 27 through 38) the pressure plateau is even shorter 

and the difference between TRACE and RELAP5 simulations is smaller than in 12.7 cm break 

size case. The difference is now mainly due to faster accumulator emptying in case of TRACE 

calculation. This has further influence on pressurizer pressure, break flow, RCS mass, core 

collapsed liquid level, hot and cold leg temperatures, core exit temperature and fuel cladding 

temperature.

In all calculated case the core heatup started in about 20 minutes. 

Figure 27 Pressurizer Pressure (15.24 cm) 

Figure 28 Cold Leg no. 1 Temperature (15.24 cm) 
19

Figure 29 Hot Leg no. 1 Temperature (15.24 cm) 

Figure 30 Break Flow (15.24 cm) 

Figure 31 Core Collapsed Liquid Level (15.24 cm) 
20

Figure 32 Core Exit Temperature (15.24 cm) 

Figure 33 Fuel Cladding Temperature (15.24 cm) 

Figure 34 RCS Mass (15.24 cm) 
21

Figure 35 Integrated Break Flow (15.24 cm) 

Figure 36 Mass Injected by Accumulators (15.24 cm) 

Figure 37 Accumulator no. 1 Flow (15.24 cm) 
22

Figure 38 Accumulator no. 2 Flow (15.24 cm) 

3.4 LOCA with 20.32 cm Break Size

In case of 20.32 cm break size (see Figures 39 through 50) the pressure plateau is negligible 

and the difference between TRACE and RELAP5 simulations is again very small. The 

differences started after accumulator injection. The period of accumulator emptying is short, but 

in the case of TRACE this period is significantly shorter comparing to RELAP5. 

The faster accumulator discharge (higher injection flow) caused higher break flow (much of the 

injected water is thus lost through the break), the pressure drops faster and there is also less 

mass in the RCS in that period. However, later the break flow is smaller due to lower pressure 

and so the TRACE calculation after approximately 450 s agrees well with RELAP5 calculations.

In all calculated case the core heatup started in about 15 minutes.

Figure 39 Pressurizer Pressure (20.32 cm) 
23

Figure 40 Cold Leg no. 1 Temperature (20.32 cm) 

Figure 41 Hot Leg no. 1 Temperature (20.32 cm) 

Figure 42 Break Flow (20.32 cm) 
24

Figure 43 Core Collapsed Liquid Level (20.32 cm) 

Figure 44 Core Exit Temperature (20.32 cm) 

Figure 45 Fuel Cladding Temperature (20.32 cm) 
25

Figure 46 RCS Mass (20.32 cm) 

Figure 47 Integrated Break Flow (20.32 cm) 

Figure 48 Mass Injected by Accumulators (20.32 cm) 
26

Figure 49 Accumulator no. 1 Flow (20.32 cm) 

Figure 50 Accumulator no. 2 Flow (20.32 cm) 

3.5 LOCA with 30.48 cm Break Size

Looking results for smaller breaks, it would be expected that TRACE results for 30.48 cm (see

Figures 51 through 62) would be even closer to RELAP5 results. However, in original TRACE 

calculation faster accumulator discharge so much influences the break flow that RCS mass (see 

Figure 58) was so low after the accumulators are emptied that another core heatup starts 1 

minute after break occurrence and the TRACE results are qualitatively very much different from 

RELAP5 results. Therefore it was decided to reduce the accumulator flow area in such a way to 

get comparable accumulator discharge between TRACE and RELAP5 (see Figures 60, 61 and 

62). As can be seen from Figure 54 showing results for 30.48 cm break size for reduced 

accumulator line area (label “TRACE (ACC)”), the TRACE break flow (and its integrated value 

shown in Figure 59) is now similar to RELAP5 calculations with some differences in the time of 

core heatup start. Slightly smaller break flow (Figure 54) means less RCS mass discharged

(Figure 58) and by this the core heatup is delayed (Figure 57). 
27

In case of RELAP5 using RT and HF critical flow model the core heatup occurred in 6 minutes 

and 8 minutes, respectively, while in TRACE in 11 minutes after break occurrence (not seen 

from Figure 57, but TRACE calculated results are available until 905 s). The results showed that 

accumulator emptying should be further studied.

Figure 51 Pressurizer Pressure (30.48 cm) 

Figure 52 Cold Leg no. 1 Temperature (30.48 cm) 
28

Figure 53 Hot Leg no. 1 Temperature (30.48 cm) 

Figure 54 Break Flow (30.48 cm) 

Figure 55 Core Collapsed Liquid Level (30.48 cm) 
29

Figure 56 Core Exit Temperature (30.48 cm) 

Figure 57 Fuel Cladding Temperature (30.48 cm) 

Figure 58 RCS Mass (30.48 cm) 
30

Figure 59 Integrated Break Flow (30.48 cm) 

Figure 60 Mass Injected by Accumulators (30.48 cm) 

Figure 61 Accumulator no. 1 Flow (30.48 cm) 
31

Figure 62 Accumulator no. 2 Flow (30.48 cm) 

3.6 Discussion of Results

In the performed TRACE LOCA analysis of 30.48 cm (case labeled “TRACE”) faster 

accumulator discharge (see Figure 60) so much influences the break flow (see Figure 54) that 

RCS mass was so low that heatup starts after 1 minute and is qualitatively very much different 

from RELAP5 calculation. Therefore it was decided to reduce the accumulator flow area in such 

a way to get comparable accumulator discharge between TRACE and RELAP5. As can be seen 

from Figures 61 and 62 showing results for 30.48 cm break size for reduced accumulator line 

area (label “TRACE (ACC)”), the TRACE break flow is now similar to RELAP5 calculations with 

differences in the time of core heatup start. In case of RELAP5 using RT and HF critical flow 

model the core heatup occurred in 6 minutes and 8 minutes, respectively, while in TRACE in 

11 minutes after break occurrence. 

As was already indicated the results showed that accumulator emptying should be further 

studied. Therefore the BETHSY 6.2TC calculation (Refs. 7 and 11) has been checked as shown 

in Figure 63. The accumulator discharge was faster in the case of TRACE comparing to 

RELAP5 calculation, which was in reasonable agreement with experimental data. The TRACE 

calculation of the same BETHSY 6.2TC test described in the TRACE code development 

assessment manual (Ref. 12) similarly shows that TRACE calculated accumulator discharge is 

faster than in the experiment as can be seen from Figure 64. In the Reference 12, page C-491 it 

is further explained: 

“Calculated and measured integrated accumulator injection flows into both loops are shown in

Figure C.8-65. In the test, the accumulator injection starts at 345 seconds and is isolated at

approximately 950 seconds. In the calculation, the accumulator injection begins at about 350

seconds and is isolated at about 650 seconds. As indicated in Reference 9, the combination of 

the larger amount of water inventory loss out the break and the earlier time that the 

accumulators complete their injection results in the earlier calculated core heatup at 

approximately 1330 seconds.”
32

Figure 63 Mass Discharged by Accumulators (Ref. 7) 

Figure 64 Integrated Accumulators Injection Flow (Fig. C.8-65 from Ref. 12) 

When comparing TRACE predictions in Figures 63 and 64, it can be seen that also in the case 

of TRACE assessment using TRACE Patch 01 code (Ref. 12), the accumulator discharge was 

predicted faster than in case of BETHSY 6.2TC experiment. Calculated accumulator water flow 

over prediction of TRACE for integral experiments presented in the Assessment Manual, 

Appendix C (Ref. 12), is seen in Figures C.5-99, C.5-100, C.5-204, C.5-205, C.6-16, C.6-17 and

C.6-37 of Reference 12, and indirectly from integrated accumulator flow in Figure C.8-65 of 

Reference 12. 
33

4. CONCLUSIONS

In the study it was demonstrated that developed input model of two-loop pressurized water reactor 

(PWR) for TRACE thermal-hydraulic systems code has the capability for independent assessment 

of RELAP5 computer code calculations. For demonstration the response of PWR to loss-ofcoolant accident (LOCA) was simulated for five break sizes ranging from 10.16 cm to 30.48 cm 

equivalent diameter break size in cold leg. Two RELAP5 code calculations were performed to see 

the influence of critical flow break model, while in case of TRACE default critical flow model was 

used.

The results showed that RELAP5 calculations using different break flow models are rather similar

between each other, therefore also other parameters are similar. The accumulators discharge 

was faster in TRACE calculations than in RELAP5 calculations for whole LOCA spectrum. 

Therefore the calculated TRACE break flow was also larger than RELAP5 calculated break flow 

during this period. This further influences the accident progression. In the case of smaller breaks 

also secondary side more significantly influences the primary pressure, but this could not be 

properly simulated with TRACE due to problems with separator component. It can be concluded 

that different accumulator discharge influencing the break flow seems to be the largest contributor 

to the differences between RELAP5 and TRACE for the performed LOCA calculations.

35

5. REFERENCES

1. USNRC, RELAP5/MOD3.3 Code Manual, Patch 04, Vols. 1 to 8, Information Systems

Laboratories, Inc. Idaho Falls, Idaho, prepared for United States Nuclear Regulatory

Commission (USNRC), 2010.

2. USNRC, TRACE V5.840 User’s Manual, Patch 04, Vols. 1 to 2, United States Nuclear

Regulatory Commission (USNRC), 2014.

3. USNRC, RELAP5/MOD3.3 Code Manual, Patch 05, Vols. 1 to 8, Information Systems

Laboratories, Inc. Idaho Falls, Idaho, prepared for United States Nuclear Regulatory

Commission (USNRC), 2016.

4. A. Prošek, B. Mavko, “Reactor trip analysis at Krško Nuclear Power Plant”, International

agreement report NUREG/IA, 0221, Office of Nuclear Regulatory Research, U.S. Nuclear

Regulatory Commission, Washington, 2010, pp. 1-57.

5. A. Prošek, I. Parzer, and B. Krajnc, “Simulation of hypothetical small-break loss-of-coolant

accident in modernized nuclear power plant”, Electrotechnical Review, 71(4), 2004, pp.

191-196.

6. I. Parzer, B. Mavko, and B. Krajnc, “Simulation of a hypothetical loss-of-feedwater

accident in a modernized nuclear power plant”, Journal of Mechanical Engineering, 49(9),

2003, pp. 430-444.

7. A. Prošek, O.-A. Berar, IJS procedure for RELAP5 to TRACE input model conversion

using SNAP, Proceedings of ICAPP'12. American Nuclear Society, USA, 2012.

8. APT, Symbolic Nuclear Analysis Package (SNAP), User's Manual, Applied Programming

Technology (APT), Inc., 2011.

9. I. Parzer, B. Mavko, Assessment of RELAP5/MOD3.3 against Single Main Steam Isolation

Valve Closure Events at the Krško Nuclear Power Plant, NUREG/IA-0223, 2010.

10. O.-A. Berar, A. Prošek, M. Leskovar, B. Mavko, Two-loop PWR RELAP5 to TRACE model

conversion and three dimensional vessel model development for coolant mixing analysis,

V: Proceedings of the 2014 International Congress on Advances in Nuclear Power Plants,

ICAPP 2014, Charlotte, North Carolina, April 6-9, 2014. Le Grange Park: American

Nuclear Society, 2014, pp. 1540-1546.

11. A. Prošek, O.-A. Berar, Advanced presentation of BETHSY 6.2TC Test results calculated

by RELAP5 and TRACE. Science and Technology of Nuclear Installations, vol. 2012,

2012, pp. 812130-812115.

12. USNRC, TRACE V5.0 ASSESSMENT MANUAL, Appendix C: Integral Effects Tests, June

18, 2008 (Accession Number ML120060172).







NUREG/IA-0479 June 2017 RELAP5 and TRACE Calculations of LOCA in PWR