NUREG/IA-0474

Steam Line Break Analysis Using 

RELAP5/MOD3.3 for Steam Generator 

Blowdown Load Assessment

Prepared by:

Han-Sik JUNG, Du-Ho HONG, Jong-In KIM, Ae-Ju CHEONG*, Kyung-Won LEE*

Doosan Heavy Industries & Construction

22, DoosanVolvo-ro, Seongsan-gu,

Changwon, Gyeongnam, 642-792, Korea

*Korea Institute of Nuclear Safety

62 Gwahak-ro, Yuseong-gu

Daejeon, 34142, Korea

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: February 2016 

Date Published: May 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

NUREG/IA-0474

Steam Line Break Analysis Using 

RELAP5/MOD3.3 for Steam Generator 

Blowdown Load Assessment

Prepared by: 

Han-Sik JUNG, Du-Ho HONG, Jong-In KIM, Ae-Ju CHEONG*, Kyung-Won LEE* 

Doosan Heavy Industries & Construction

22, DoosanVolvo-ro, Seongsan-gu,

Changwon, Gyeongnam, 642-792, Korea

*Korea Institute of Nuclear Safety

62 Gwahak-ro, Yuseong-gu

Daejeon, 34142, Korea

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: February 2016 

Date Published: May 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

iii

ABSTRACT

The RELAP5/MOD3.3 is generally used for best-estimate transient simulation of light water reactor 

coolant systems during postulated accidents in the Light Water Reactor (LWR). The RELAP5/MOD3.3 

code is based on a non-homogeneous, and non-equilibrium model for a two phase system that is solved 

by a fast, partially implicit numerical scheme to permit economical calculation of system transients. This 

code is suitable for the analysis of transients and postulated accidents in LWR systems, including both 

large- and small-break loss of coolant accidents, as well as for the full range of operational transients. 

For the evaluation of structural integrity for the steam generator in the Pressurized Water Reactor 

(PWR), the postulated accidents, such as the Steam Line Break (SLB) in the Advanced Power

Reactor (APR1400) at the Korean domestic plants, are considered Design Basis Events (DBE). In 

order to evaluate the structural integrity of a steam generator during the SLB, the data for the thermohydraulic velocity, density and pressure are needed. 

This study was performed to calculate thermal hydraulic parameters, such as thermo-hydraulic 

velocity, density and pressure, using the RELAP5/MOD3.3 code for the structural evaluation of the 

steam generator internals during the postulated SLB accidents.

The calculation results were verified by comparing with experimental data generated from the 

experimental facility ATLAS (Advanced Thermal-Hydraulic Test Loop for Accident Simulation)

v 

TABLE OF CONTENTS

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

LIST OF FIGURES..............................................................................................................................vii

LIST OF TABLES................................................................................................................................vii

EXECUTIVE SUMMARY .....................................................................................................................ix

ABBREVIATIONS AND ACRONYMS................................................................................................xi

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

2. DESCRIPTION OF THE ATLAS SG SLB EXPERIMENT............................................................ 3

2.1 Experimental Facilities............................................................................................................. 3

2.2 Instrumentation ........................................................................................................................ 7

2.3 Experimental Procedure.......................................................................................................... 9

2.4 Experimental Conditions.......................................................................................................... 9

3. RELAP5 INPUT MODEL..............................................................................................................11

3.1 ATLAS Facility Steam Generator RELAP5 Input Model.......................................................11

4. RELAP5 ANALYSIS RESULTS...................................................................................................15

4.1 Steady State Calculation .......................................................................................................15

4.2 Transient Calculation.............................................................................................................16

Accumulated Mass of Break Flow..............................................................................17

Flow Rate....................................................................................................................18

Temperature ...............................................................................................................19

Pressure......................................................................................................................20

Dynamic Pressure ......................................................................................................21

4.3 Sensitivity Study.....................................................................................................................22

Critical Flow Model......................................................................................................22

Break Valve Opening Time.........................................................................................23

Time Step Size............................................................................................................24

Nodalization ................................................................................................................25

5. CONCLUSIONS............................................................................................................................27

6. REFERENCES..............................................................................................................................29

vii

LIST OF FIGURES

Figure 2-1 Major Component of the ATLAS........................................................................................ 4

Figure 2-2 Steam Generator of the ATLAS......................................................................................... 5

Figure 2-3 Configuration of the Break Simulation System.................................................................. 6

Figure 2-4 Configuration of the Double Rupture Disc Assembly........................................................ 6

Figure 2-5 Locations of Measurement ID in the Steam Generator of the ATLAS.............................. 8

Figure 2-6 Detailed Locations of Dynamic Pressure Sensors Installed in the SG............................. 8

Figure 3-1 MARS-KS Code ATLAS Nodalization .............................................................................11

Figure 3-2 The ATLAS Steam Generation RELAP5/MOD3.3 Nodalization.....................................13

Figure 4-1 Accumulated Mass of Break Flow ...................................................................................17

Figure 4-2 Flow Rate at the Main Steam Line...................................................................................18

Figure 4-3 Temperature at the Hot Side Downcomer .......................................................................19

Figure 4-4 Temperature at the Cold Side Downcomer......................................................................20

Figure 4-5 Pressure at the Steam Dome ..........................................................................................20

Figure 4-6 Dynamic Pressure - 03 ....................................................................................................21

Figure 4-7 Dynamic Pressure - 03 for the Critical Flow Model .........................................................22

Figure 4-8 Dynamic Pressure - 03 for the Break Valve Opening Time............................................24

Figure 4-9 Dynamic Pressure - 03 for the Time Step Size ...............................................................25

Figure 4-10 Dynamic Pressure - 03 for the Nodalization....................................................................26

LIST OF TABLES

Table 2-1 List of Measuring Locations and Parameters.................................................................... 7

Table 2-2 Uncertainty Level of Instruments ....................................................................................... 7

Table 2-3 Sequence of Major Events................................................................................................. 9

Table 2-4 Actual Initial Parameters of the ATLAS Steady-State Condition.....................................10

Table 4-1 Steady State Results........................................................................................................15

Table 4-2 Sequence of Events.........................................................................................................16

ix

EXECUTIVE SUMMARY

The RELAP5/MOD3.3 is a computer code for best-estimate transient simulation of light water reactor 

systems during the postulated accidents. In this study, the possibility of the application of the 

postulated SLB accident analysis for a steam generator was investigated using the RELAP5/MOD3.3. 

For the verification and validation of the code analysis results, experimental tests were performed in 

the thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for 

Accident Simulation) at KEARI, located in Daejeon, Korea. The code analysis results for the thermal 

hydraulic response during the SLB accidents were compared to those of experimental tests. Some 

sensitivity studies were also performed.

From this study, major results of the analysis and experiments are summarized below.

• The RELAP5/MOD3.3 predicts well the thermal hydraulic behaviors of the ATLAS SLB

experimental test.

• The dynamic pressure estimated with the RELAP5/MOD3.3 is slightly conservative

compared to that of the ATLAS SLB experimental test data.

• The following conclusions are drawn from the results of the sensitivity studies.

• The Henry-Fauske critical flow model is recommended.

- The break valve opening time of 1.0 milli-sec is recommended.

- A maximum time-step-size less than or equal to or 10-4 sec is recommended.

- The nodalization effect in the components around the steam line throat is

- Negligible

xi

ABBREVIATIONS AND ACRONYMS

ATLAS Advanced Thermal-Hydraulic Test Loop for Accident Simulation

APR1400 Advanced Power Reactor 1400 MWe

SG Steam Generator

DAS Data Acquisition System

SLB Steam Line Break

KAERI Korea Atomic Energy Research Institute

MARS-KS Multi-dimensional Analysis of Reactor Safety

RCS Reactor Coolant System

RELAP Reactor Excursion and Leak Analysis Program

US NRC United States Nuclear Regulatory Commission

1 

1 INTRODUCTION

The light water reactor (LWR) transient analysis code, RELAP5, was developed at the Idaho National 

Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC). Uses of code

include analyses required to support rulemaking, licensing audit calculations, evaluation of accident 

mitigation strategies, evaluation of operator guidelines, and experiment planning analysis. RELAP5 

has also been used as the basis for a nuclear plant analyzer. 

The mission of the RELAP5/MOD3.3 development program was to develop a code version suitable 

for analysis of all transients and postulated accidents in LWR systems, including both large- and smallbreak loss-of-coolant accidents (LOCAs) as well as the full range of operational transients.

This study focuses on the applicability of the RELAP5/MOD3.3 for a guillotine break transient showing 

very sudden change of thermal hydraulic conditions in the system in a short period of time. That is, 

the code is analyzed to check whether or not it predicts well the rapid thermal hydraulic response in 

an SLB accident. 

From a structural integrity point of view, the pressure difference between the internals of the steam 

generator is a critical parameter because the pressure difference is the only structural load during the 

guillotine break accidents. To verify the pressure difference inside the steam generator, the dynamic 

pressure near the break line was investigated and analyzed in relation to experimental test data.

3 

2 DESCRIPTION OF THE ATLAS SG SLB EXPERIMENT 

Experimental tests were performed to verify the RELAP5/MOD3.3 results. The tests were performed 

at a special facility for testing the thermal hydraulic integral effect. Transducers were installed to check 

accurately the dynamic pressure data. The experimental tests proceeded to reach a steady state 

condition and then the break was simulated with data logging. During the test, the major thermalhydraulic parameters, such as dynamic and static pressures, local temperatures, and flow rates, were 

obtained in the course of an abrupt break of the steam generator steam line using the double rupture 

disk assembly. Also, the reproducibility of the test was checked by doing additional test cases 

observing the characteristics of the dynamic pressure during the tests. Details are shown in the 

following subsections.

2.1 Experimental Facilities

A thermal-hydraulic integral effect test facility, ATLAS (Advanced Thermal-hydraulic Test Loop for 

Accident Simulation), was used to perform tests for a steam line break in the steam generator. ATLAS 

has the same two-loop features as the APR1400 (Advanced Power Reactor 1400 MWe); but is a halfheight and 1/288-volume scaled test facility with respect to APR1400. The fluid system of the ATLAS 

consists of a primary system, a secondary system, a safety injection system, a break simulating 

system, a containment simulating system, and auxiliary systems. Figure 2-1 shows a 3-dimensional 

view of the ATLAS. 

The ATLAS has two steam generators and each steam generator consists of a lower plenum, a Utube assembly, middle and upper SG vessels, two downcomer pipes, and other internals as shown 

in Figure 2-2. 

A steam line break was simulated by installing a break spool piece in one of the steam line in the SG1. Figure 2-3 shows the configuration of the break simulation system for the steam line break of the

ATLAS steam generator. The break opening time is the most crucial factor influencing a blow-down 

load during a steam line break, so it should be simulated appropriately in the test. In order to make 

the break opening time as short as possible, a double rupture disc assembly was used in the test. 

Figure 2-4 shows the configuration of the double rupture disc assembly which consists of two rupture 

discs having different cutoff pressures for actuation. 

The double rupture disc assembly works as followings: rupture disc-2 will be opened at first when the 

pressure in the “intermediate region” is increased up to a specified actuation pressure by operators. 

Subsequently, rupture disc-1 will be opened within very short period of time (about 1 milli-sec) by the 

driving force resulting from opening rupture disc-2.
4 

Figure 2-1 Major Component of the ATLAS
5 

Figure 2-2 Steam Generator of the ATLAS
6 

Figure 2-3 Configuration of the Break Simulation System

Figure 2-4 Configuration of the Double Rupture Disc Assembly
7 

2.2 Instrumentation

In the ATLAS test facility, 1,300 instrument were installed for measurement of several thermalhydraulic parameters in the components. The measuring locations that would be affected most after 

the break were selected. Table 2-1 shows a list of the measuring locations, parameters, and tag name 

of sensors. Figure 2-5 shows the locations of measurement ID in the ATLAS steam generator. 

The dynamic pressure was measured using Kistler dynamic pressure transducers with a 

measurement frequency of 10000 Hz. In order to precisely estimate the rupture time of the discs, 

additional dynamic pressure transducers (Dynamic-P-04 and Dynamic-P-05) were installed in the 

pipe line of the break simulation system as shown in Figure 2-6. 

shows the analysis uncertainty levels of each group of instruments.

Table 2-1 List of Measuring Locations and Parameters

ID Location

Dynamic

Pressure

Static

Pressure

Fluid

Temperature

Mass

Flow rate

1 Outside of main steam line QV-MS1-01

2 Top of steam dome Dynamic-P-02 PT-SGSD1-01

3 Outlet region of steam 

separator

Dynamic-P-01 TF-SGSD1-03

4 Downcomer feedwater line TF-MF1-02 QV-MF1-02

5 Downcomer cold side TF-SGDC1-04

6 Downcomer hot side TF-SGDC1-02

7 Economizer feedwater line TF-MF1-01 QV-MF1-01

8 Inside of lower plenum from 

hot leg

TF-SGP1-01

9 Inside of lower plenum to cold 

leg

TF-SGP1-02

10 Hot leg side PT-HL1-01 QV-HL1-01B

11 Cold leg side QV-CL1A-01B

QV-CL1B-01B

Table 2-2 Uncertainty Level of Instruments

Items Unit Uncertainty

Static Pressure MPa 0.039 %

Dynamic Pressure bar 1.02 %

Differential Pressure kPa 0.23 %

Collapsed Water Level m 0.17 %

Temperature oC maximum 2.4 oC 

Flow rate kg/s 0.053 %
8 

Figure 2-5 Locations of Measurement ID in the Steam Generator of the ATLAS

Figure 2-6 Detailed Locations of Dynamic Pressure Sensors Installed in the SG
9 

2.3 Experimental Procedure

After a steady state condition was achieved in both primary and secondary systems, a break in the

steam generator steam line was simulated according to the following procedures.

- ATLAS main data is logged.

- After about 10 minutes from start of the ATLAS data logging, sub-DAS is started to acquire 

the dynamic pressure data in the steam generator.

- After a few sec from the start of the sub-DAS data logging, OV-MSIV1-01 valve is closed to 

block the main steam line from the break simulation line.

- After a few sec from closing of the OV-MSIV1-01 valve, nitrogen gas is supplied into the 

“Intermediate region” as shown in Figure 2-4 to rupture the disc-2. Subsequently, the disc-1 

is opened by the driving force resulting from rupture of the disc-2.

- After acquiring experimental data for about 30 sec, Sub-DAS and ATLAS data logging are 

stopped.

The major events of the tests is listed in Table 2-3. 

Table 2-3 Sequence of Major Events

Events Time (sec) Description

Heat Up ~ 18,000 Heating up process

Test Standby ~ 600 Steady state condition

Test Start

0.0 ATLAS main data logging start

600.0 Sub-DAS data logging start (Dynamic Pressure)

607.5 Blocking of main steam line

Break 613.0 SG-1 steam line break 100% open

Test End 630.0 Data recording stop

2.4 Experimental Conditions

The tests aimed to obtain major thermal-hydraulic parameters, such as dynamic and static pressures 

near the break location, local temperatures, and flow rates during the steam line break of the steam 

generator. Two tests, named SLB-DS-01 and SLB-DS-02, were performed using the double rupture 

disc assembly in order to simulate a sudden break of the steam generator steam line. The reason for 

performing SLB-DS-02 was to confirm the reproducibility of the SLB tests, and the reproducibility of 

the tests was confirmed by comparing the SLB test results. Therefore, only the SLB-DS-01 data was

used to compare with the RELAP5/MOD3.3 calculation results. 

Table 2-4 presents the actual initial conditions measured in these two tests. In 

Table 2-4, STDEV stands for standard deviations of each designated values. Even though there are 

some discrepancies between the target values and the measured values, the measured initial 

conditions of the primary and secondary systems are acceptable considering the standard deviations 

of each value, and the characteristics of the integral effect test.
10 

Table 2-4 Actual Initial Parameters of the ATLAS Steady-State Condition

Design 

Parameter

Target 

Value

Measured Value

Remark

SLB (Sensor ID) -DS-01

(Value(1)/STDEV(2)) 

SLB-DS-02

(Value(1)/STDEV(2))

REACTOR PRESSURE VESSEL

Normal

Power (MW) 1.56 1.630 / 0.0008 1.624 / 0.0006 Heat loss: about 80 kW

Pressurizer 

Pressure (MPa) 15.5 15.5 / 0.006 15.5 / 0.008 PT-PZR-01

Core Inlet

Temperature (oC) 290.7 289.8 / 0.17 290.0 / 0.15 TF-LP-02G18

Core Outlet

Temperature (oC) 324.2 325.6 / 0.14 325.7 / 0.13 TF-CO-07G14,18,21,25

STEAM GENERATOR

Steam Flow

Rate (kg/s) 0.444 0.382 / 0.0070.418 / 0.001

0.377 / 0.006

0.409 / 0.004

SG-1 (QV-MS1-01)

SG-2 (QV-MS2-01)

Feed Water Flow

Rate (kg/s) 0.444 0.434 / 0.0030.426 / 0.002

0.397 / 0.007

0.403 / 0.007

SG-1 (QV-MF1-01, 2)

SG-2 (QV-MF2-01, 2)

Feed Water

Temperature (oC) 232.2 234.2 / 0.16233.6 / 0.17

233.1 / 0.13

232.6 / 0.13

SG-1 (TF-MF1-01)

SG-2 (TF-MF2-01)

Steam

Pressure (MPa) 7.83 7.82 / 0.0097.82 / 0.009

7.84 / 0.005

7.84 / 0.005

SG-1 (PT-SGSD1-01)

SG-2 (PT-SGSD2-01)

Steam

Temperature (oC) 293.5 295.2 / 0.11295.4 / 0.09

295.4 / 0.08

295.5 / 0.08

SG-1 (TF-SGSD1-03)

SG-2 (TF-SGSD2-03)

PRIMARY PIPING

Cold Leg Flow

(kg/s) 2.0 1.94 / 0.016 1.90 / 0.017

Cold leg average

(QV-CL1A,1B,2A,2B01B)

Notes) 

(1) Average Value at -10 to 300 sec

(2) STDEV: Standard Deviation at -10 to 300 sec
11

3 RELAP5 INPUT MODEL

3.1 ATLAS Facility Steam Generator RELAP5 Input Model

The RELAP5/MOD3.3 input model was prepared from the MARS-KS code input model originally

created at KAERI. The MARS-KS code input model consists of a reactor pressure vessel, primary 

piping, steam generators, a pressurizer, steam lines, a safety injection system, feedwater and the 

turbine system, and reactor coolant pumps. Figure 3-1 shows the MARS-KS code input model 

nodalization. 

Figure 3-1 MARS-KS Code ATLAS Nodalization
12

To simulate the steam line break test, the RELAP5/MOD3.3 input model took one steam generator 

from the MARS-KS code input model and modified it accordingly for analysis. Nodalization of the 

RELAP5/MOD3.3 input model for the ATLAS steam line break is shown in Figure 3-2. The model is 

based on 78 volumes connected by 38 junctions and 15 heat structures. 

The primary side of the steam generator consists of primary inlet/outlet plenum and U-tubes.

- The primary inlet/outlet plenums (C330/C350) are modeled as BRANCH components and 

are connected to RCS loops (C300/C390/C391) and the primary side of the U-tubes (C340). 

RCS loop are modeled as TIME DEPENDENT VOLUME components and used as the 

boundary condition in the primary side of the steam generator. 

- The U-tubes consist of 12 volumes in a PIPE component.

The secondary side of the steam generator consists of economizer, evaporator, riser, downcomer, 

hot and cold side downcomer pipe, separator, bypass, steam dome, economizer feedwater line, 

donwcomer feedwater, main steam line and steam break line.

- The economizer (C630) is modeled as a PIPE component and has 2 volumes. 

- The evaporator is divided into the hot side evaporator (C640/C650) and the cold side 

evaporator (C651). The hot side evaporator is modeled as a PIPE component and has 6 

volumes and the cold side evaporator is also modeled as a PIPE component and has 4 

volumes. The hot/cold side evaporators (C650/C651) are also connected by cross flow

which is modeled as a MULTIPLE JUNCTION component. 

- The riser (C659) is modeled as a BRANCH component connected to the evaporator and 

the separator.

- The downcomer (C610) is modeled as an ANNULUS component and has 3 volumes. 

- The hot/cold side downcomer pipes (C624/C620) are modeled as PIPE components and 

have 3 and 2 volumes, respectively. 

- The separator (C660) is modeled as a SEPARATOR component.

- The bypass (C670) is modeled as a SINGLE VOLUME component.

- The steam dome (C680/C690) is modeled as BRANCH components. 

- The economizer feedwater line consists of an economizer box (C615) and economizer

feedwater pipes (C617/C618). The economizer box and the economizer feedwater pipes

are modeled as PIPE components and consists of 14 volumes. The economizer feedwater 

pipes are connected to TIME DEPENDENT VOLUME components (C700/C701) used as

boundary condition to the economizer feedwater line. 

- The downcomer feedwater (C604) is modeled as a TIME DEPENDENT VOLUME 

component used as boundary condition to the downcomer feedwater and connected to the 

downcomer by a TIME DEPENDENT JUNCTION component (J605).

- The main steam line (C800/C810/C694/C696/C698) is modeled as SINGLE VOLUME 

components and a TIME DEPENDENT VOLUME component used as boundary condition 

to the main steam line connected by a VALVE component. The main steam line consists of 

5 volumes. 

- The steam break line (C910/C920/C930/C940) is modeled as SINGLE VOLUME

components, a PIPE component, and a TIME DEPENDENT VOLUME component and has 

15 volumes. A TIME DEPENDENT VOLUME component (C940) provides the boundary 

condition as the atmosphere and is connected by a VALVE component (J931). 
13

Figure 3-2 The ATLAS Steam Generation RELAP5/MOD3.3 Nodalization

15

4 RELAP5 ANALYSIS RESULTS

4.1 Steady State Calculation

In order to achieve a stable initial condition, the steady state calculation was performed for 3000 sec. 

The following controllers were used for the first 3000 sec: 

- The downcomer feedwater flow rate proportional controller

- The economizer feedwater flow rate proportional controller

- The steam dome pressure proportional controller

The other controlled parameters (feedwater temperature, primary coolant temperature, primary 

coolant pressure, and primary coolant flow rate) were entered as boundary conditions. 

Table 4-1 shows the calculated parameters compared to the experiment. The calculated steam dome 

pressure was a little different from that of experimental value. Because the measured steam dome 

pressure was lower than the saturation pressure at the given steam dome temperature, the saturation 

pressure at that given steam dome temperature was used as the steam dome pressure in the steady 

state calculation. The calculated steam flow rate was a little different from that of experimental value. 

Although the calculated steam and feedwater flow rates were equal, their measured values were not 

equal. It is noted that the instrument used for the water flow rate is more accurate than for the steam 

flow rate. Therefore, the calculated steam and feedwater flow rates were adjusted to the 

experimentally measured feedwater flow rate.

Once a stable condition was obtained, the downcomer and economizer feedwater controllers were

de-activated and replaced by a relevant boundary condition for the transient analysis.

Table 4-1 Steady State Results

Parameters Measured(1) Calculated

Primary system

Hot leg pressure (MPa) 15.571±0.006 15.571

Inlet plenum temperature (K) 599.3±2.4 598.7

Outlet plenum temperature (K) 566.4±2.4 564.3

Cold leg flow rate (kg/s)(2) 1.8975±0.001 1.8975

Secondary system

Thermal power (MW) 0.749 0.749

Steam dome pressure (MPa) 7.812±0.003 8.013 

Steam dome temperature (K) 568.3±2.4 568.3

Steam flow rate (kg/s) 0.382±0.0002 0.42822

Economizer feedwater temperature (K) 507.3±2.4 507.3

Downcomer feedwater temperature (K) 503.5±2.4 503.5

Economizer feedwater flow rate (kg/s) 0.392±0.0002 0.3922

Downcomer feedwater flow rate (kg/s) 0.036±0.00002 0.036

Notes) 

(1) Average value during 0 to 600 sec

(2) Average of cold legs A and B 
16

4.2 Transient Calculation

In the SLB experiment (SLB-DS-01), after an initial steady-state condition was reached, this condition 

was maintained for about 600 sec. The steam line break test was initiated by closing the main steam 

line valve at 607.5 sec and opening the break line valve at 613.1 sec. 

The RELAP5/MOD3.3 calculation was also conducted for null transient analysis during 600 sec. The 

break was initiated by closing the 697 valve at 607.5 sec and opening the 931 valve at 613.1 sec as 

in the SLB experiment (SLB-DS-01). Table 4-2 shows the sequence of events in both the calculation 

and the ATLAS experiment.

Table 4-2 Sequence of Events

Events

ATLAS SLB Test

[SLB-DS-01] 

(sec)

RELAP5/MOD3.3

Calculation

(sec)

Remark

Steady state condition 0.0 to 600.0 0.0 to 600.0 null transient

Blocking of main steam line 607.5 607.5 697 valve close

SG-1 steam line break 100% 

open

613.1 613.1 931 valve open

Test end 630.0 630.0

In the transient calculation, the Henry-Fauske critical flow model was used and the maximum time 

step size was 0.000001 sec during the period from 613.0 to 613.2 sec. The RELAP5/MOD3.3 

calculation data were compared with experimental data in Figure 4-1 through Figure 4-6. The 

experimental data are labeled “Exp.”, and the RELAP5/MOD3.3 calculation data as “Cal.”.
17

4.2.1 Accumulated Mass of Break Flow 

The RELAP5/MOD3.3 calculation was in good agreement with the experiment data for accumulated 

mass of break flow, as presented in Figure 4-1. The Henry-Fauske critical flow model was used for 

the critical flow model. To obtain better agreement between calculated and experimental results, 0.41 

was used for the discharge coefficient and 0.14 (default) was used for the thermal non-equilibrium at

the throat (junction 921).

Figure 4-1 Accumulated Mass of Break Flow
18

4.2.2 Flow Rate

The main steam line flow behaved differently in the RELAP5/MOD3.3 calculation and in the 

experiment. The main steam line valve (valve 697) closed quickly in the RELAP5/MOD3.3 calculation,

but the main steam line valve was closed slowly (manually) in the experiment. After the break valve 

(rupture disk) opening, leakage appeared in the main steam line, because the main steam line 

isolation valve did not provide perfect isolation. Because, the meaningful thermal hydraulic parameters 

were acquired from the experiment during a short period of time (approximately 0.1 to 0.2 sec), this 

difference is immaterial. 

Figure 4-2 Flow Rate at the Main Steam Line
19

4.2.3 Temperature

The temperature calculated by the RELAP5/MOD3.3 was in good agreement with experimental data, 

as presented in Figure 4-3 and Figure 4-4, which show the temperature at the hot side downcomer 

and cold side downcomer, respectively. The temperatures at hot/cold side downcomers were slightly 

over-predicted after 623 sec. 

Figure 4-3 Temperature at the Hot Side Downcomer 
20

Figure 4-4 Temperature at the Cold Side Downcomer 

4.2.4 Pressure

The RELAP5/MOD3.3 calculation was in good agreement with experimental data for pressure at the 

steam dome region, as presented in Figure 4-5. After 622 sec, the steam dome pressure began to 

show deviations between the RELA5/MOD3.3 calculation and the experimental results. The 

experimental steam dome pressure decreased more rapidly because of the steam leak through the 

main steam line, which was not perfectly isolated, as presented in Figure 4-2. 

Figure 4-5 Pressure at the Steam Dome
21

4.2.5 Dynamic Pressure

The RELAP5/MOD3.3 calculation conservatively predicted the dynamic pressure for the steam line 

break of the steam generator, as presented in Figure 4-6. Locations of the dynamic pressure 

measurements are shown in Figure 2-6. The maximum dynamic pressure was observed at dynamic 

pressure - 03 which is 0.134 MPa in experiment and 0.145 MPa in the RELPA5/MOD3.3 calculation.

The dynamic pressures - 01 and 02 are not shown, because they were too small to compare.

Figure 4-6 Dynamic Pressure - 03 
22

4.3 Sensitivity Study

Sensitivity studies were performed to investigate the important factors affecting the dynamic pressure. 

The important factors are the critical flow model, the break valve opening time, the time step size, and 

nodalization. Detailed results are presented in the following sub-sections. 

4.3.1 Critical Flow Model

The critical flow model is an important factor in the steam line break analysis. Thus, a sensitivity 

analysis of the critical flow model was performed. The critical flow models used for this sensitivity 

analysis were the Henry-Fauske critical flow model, the Modified Henry-Moody critical flow model and 

the Original RELAP choked flow model.

Figure 4-7 shows the behavior of the dynamic pressure depending on the critical flow model. The 

Henry-Fauske model predicted the dynamic pressure more conservatively than the Modified HenryMoody and the Original RELAP models. To obtain conservative results, the Henry-Fauske critical flow 

model is recommended for the steam line break of the steam generator. 

Figure 4-7 Dynamic Pressure - 03 for the Critical Flow Model
23

4.3.2 Break Valve Opening Time

The break valve opening time of 1.0 milli-sec is required for the steam line break analysis by the Korea 

Institute of Nuclear Safety. In the test, the maximum break valve opening time was 1.4 milli-sec. Thus, 

a sensitivity analysis was performed to find out the effect of changes in the break valve opening time. 

The analysis ranged from 1.0 to 2.5 milli-sec. 

Figure 4-8 shows the behavior of the dynamic pressure depending on the break valve opening time.

The maximum dynamic pressure increased as the break valve opening time decreased to 1.0 millisec. When the break valve opening time was 1.0 and 2.5 milli-sec, the maximum dynamic pressure 

was 0.145 and 0.144 MPa, respectively. 
24

Figure 4-8 Dynamic Pressure - 03 for the Break Valve Opening Time

4.3.3 Time Step Size

After opening of the steam line break valve, the maximum dynamic pressure was reached within 0.007 

sec. Thus, the time step size is an important factor in steam line break analysis and a sensitivity study 

for time step size was performed. The analysis for the maximum time step size ranged from 10-7 to 

10-3 sec.

Figure 4-9 shows that the maximum dynamic pressure increased as the maximum time step size 

decreased to 10-6 sec, and that the dynamic pressure was almost the same below that value. A 
25

maximum time step size less than or equal to 10-4 sec is recommended to obtain a most conservative 

maximum dynamic pressure. When the maximum time step size was 10-4 sec, the calculated 

maximum dynamic pressure was 0.136 MPa (0.134 MPa in the experiment).

Figure 4-9 Dynamic Pressure - 03 for the Time Step Size

4.3.4 Nodalization 

In the steam line break analysis, the nodalization in the components around the steam line throat

could influence the analysis results. Thus, a sensitivity analysis on the nodalization was performed by 

varying the number of volumes in the PIPE component (C920). The numbers of C920 volumes 

chosen for the sensitivity analysis were 1, 3, and 6. 
26

The results of the sensitivity analysis on the nodalization are shown in Figure 4-10. The maximum 

dynamic pressure decreased as the number of volumes increased. However, because the difference 

was not significant, the effect of the nodalization in the components around the steam line throat is

negligible.

Figure 4-10 Dynamic Pressure - 03 for the Nodalization
27

5 CONCLUSIONS

The steam line break accident was simulated using RELAP5/MOD3.3. The calculation results were 

compared to the experimental data. As a result, the pressure, temperature, and accumulated mass 

behaviors were found to be well predicted, and the maximum dynamic pressure was predicted 

conservatively. Thus, it is concluded that use of the RELAP5/MOD3.3 should be acceptable for 

calculation of the steam generator blowdown load from steam line break accident.

29

6 REFERENCES

1. Nuclear Regulatory Commission, RELAP5/MOD3.3 Code Manual, Volume II: USER’S

GUIDE AND INPUT REQUIREMENTS, Appendix A Input Requirements, October 2010.

2. K. H. Kang, et al., “Detailed Description Report of ATLAS Facility and Instrumentation”,

KAERI/TR-4316/2011, Korea Atomic Energy Research Institute, 2011.

3. K. H. Kang, et al., “Analysis Report on the steam generator SLB test for the SPACE code

verification”, ATLAS-DSSG-TR-14-01, Rev.2, Korea Atomic Energy Research Institute, 2015.

4. American Nuclear Society, “Design Basis for Protection of Light Water Nuclear Power Plant

Against the Effects of Postulated Pipe Rupture,” ANSI/ANS-58.2-1988, American Nuclear

Society, La Grange Park, IL., 1988.

2017

Steam Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator May 

Blowdown Load Assessment 

Han-Sik JUNG, Du-Ho HONG, Jong-In KIM, Ae-Ju CHEONG*, Kyung-Won LEE*

Doosan Heavy Industries & Construction *Korea Institute of Nuclear Safety

22, DoosanVolvo-ro, Seongsan-gu, 62 Gwahak-ro, Yuseong-gu

Changwon, Gyeongnam, 642-792, Korea Daejeon, 34142, Korea

Division of Systems Analysis

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

K. Tien, NRC Project Manager

The RELAP5/MOD3.3 is generally used for best-estimate transient simulation of light water reactor coolant systems during 

postulated accidents in the Light Water Reactor (LWR). The RELAP5/MOD3.3 code is based on a non-homogeneous, and 

non-equilibrium model for a two phase system that is solved by a fast, partially implicit numerical scheme to permit 

economical calculation of system transients. This code is suitable for the analysis of transients and postulated accidents in

LWR systems, including both large- and small-break loss of coolant accidents, as well as for the full range of operational 

transients. 

For the evaluation of structural integrity for the steam generator in the Pressurized Water Reactor (PWR), the postulated 

accidents, such as the Steam Line Break (SLB) in the Advanced Power Reactor (APR1400) at the Korean domestic plants, 

are considered Design Basis Events (DBE). In order to evaluate the structural integrity of a steam generator during the SLB, 

the data for the thermo-hydraulic velocity, density and pressure are needed. 

This study was performed to calculate thermal hydraulic parameters, such as thermo-hydraulic velocity, density and 

pressure, using the RELAP5/MOD3.3 code for the structural evaluation of the steam generator internals during the 

postulated SLB accidents.

The calculation results were verified by comparing with experimental data generated from the experimental facility ATLAS 

(Advanced Thermal-Hydraulic Test Loop for Accident Simulation)

Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS)

Steam Line Break (SLB)

RELAP5/MOD3.3

Design Basis Events (DBE)

Advanced Power Reactor (APR1400)

NUREG/IA-0474

Technical



NUREG/IA-0474 Steam Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator

Blowdown Load Assessment

 May 2017