NUREG/IA-0473

Feedwater 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 Industry and Construction

22, Doosan Volvo-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: November 2016

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-0473

Feedwater 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 Industry and Construction

22, Doosan Volvo-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: November 2016

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 Feedwater Line Break (FLB) 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 FLB, 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 FLB 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 FLB EXPERIMENT.............................................................3

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

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

2.3 Experimental Procedure .......................................................................................................10

2.4 Experimental Conditions.......................................................................................................11

3 RELAP5 INPUT MODEL...............................................................................................................13

3.1 ATLAS Facility Steam Generator RELAP5 Input Model......................................................13

4 RELAP5 ANALYSIS RESULTS ...................................................................................................17

4.1 Steady State Calculation ......................................................................................................17

4.2 Transient Calculation ............................................................................................................19

4.2.1 Flow Rate ..................................................................................................................20

4.2.2 Temperature..............................................................................................................21

4.2.3 Pressure....................................................................................................................22

4.2.4 Dynamic Pressure.....................................................................................................23

4.3 Sensitivity Study....................................................................................................................24

4.3.1 Critical Flow Model....................................................................................................24

4.3.2 Break Valve Opening Time.......................................................................................26

4.3.3 Time Step Size..........................................................................................................27

4.3.4 Break Valve Discharge Coefficient ...........................................................................28

5 CONCLUSIONS.............................................................................................................................29

6 REFERENCES...............................................................................................................................31

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 System in the ATLAS .............................................................6

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

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

Figure 2-6 Detailed Locations of Dynamic Pressure Sensors Installed in the Steam Generator ......9

Figure 2-7 Measurement Location of Dynamic Pressure in the Break System .................................9

Figure 3-1 MARS-KS Code ATLAS Nodalization .............................................................................13

Figure 3-2 The ATLAS Steam Generator RELAP5/MOD3.3 Nodalization ......................................15

Figure 4-1 Flow Rate at the Economizer Feedwater Line ................................................................20

Figure 4-2 Flow Rate at the Downcomer Feedwater Line................................................................20

Figure 4-3 Temperature at the Steam Dome....................................................................................21

Figure 4-4 Temperature at the Top of the Economizer.....................................................................21

Figure 4-5 Temperature at the Top of the Hot Side Tubesheet........................................................22

Figure 4-6 Pressure at the Steam Dome ..........................................................................................22

Figure 4-7 Dynamic Pressure - 01 (the Economizer Box)................................................................23

Figure 4-8 Dynamic Pressure - 02 (Top of the Economizer)............................................................23

Figure 4-9 Dynamic Pressure - 01 for the Critical Flow Model .........................................................25

Figure 4-10 Dynamic Pressure - 01 for the Break Valve Opening Time............................................26

Figure 4-11 Dynamic Pressure - 01 for the Time Step Size...............................................................27

Figure 4-12 Dynamic Pressure - 01 for the Break Valve Discharge Coefficient ................................28

LIST OF TABLES

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

Table 2-2 Uncertainty Level of Instruments .......................................................................................8

Table 2-3 Sequence of Major Events...............................................................................................10

Table 2-4 Actual Initial Parameters of the ATLAS Steady-State Condition.....................................12

Table 4-1 Steady State Results........................................................................................................18

Table 4-2 Sequence of Events.........................................................................................................19

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 FLB 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 FLB 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 FLB experimental 

test.

The dynamic pressure estimated with the RELAP5/MOD3.3 is conservative compared to that of the 

ATLAS FLB experimental test data.

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

- The Henry-Fauske and the Modified Henry-Moody 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 10-4 sec is recommended.

- The break valve discharge coefficient of 1.0 is recommended.

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

FLB Feedwater 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 FLB 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 FLB 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 feedwater 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 feedwater line break in the steam generator. 

ATLAS has the same two-loop features as the APR1400 (Advanced Power Reactor 1400 MWe); but

is a half-height 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 feedwater line break was simulated by installing a break spool piece on the economizer feedwater 

line in SG-1. Figure 2-3 shows the configuration of a break simulation system for the feedwater line 

break of the ATLAS steam generator. The break opening time is the most crucial factor influencing a 

blowdown load during a feedwater 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 disk assembly was used 

in the test. Figure 2-4 shows the configuration of the double rupture disk assembly, which consists of 

two rupture disks that have different cut off pressure for actuation.

The double rupture disk assembly works as follows: rupture disk-2 will be opened first, when the 

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

Subsequently, rupture disk-1 will be opened within a very short period of time (approximately 1 millisec) by the driving force resulting from opening rupture disk-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 System in the ATLAS

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

2.2 Instrumentation

In the ATLAS test facility, 1,236 instrument were installed for the 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 2000 Hz. In order to precisely estimate the rupture time of the disks, 

additional dynamic pressure transducers (Dynamic-P-06 and Dynamic-P-07) were installed in the 

pipe line of the break simulation system as shown in Figure 2-3. The detailed location of dynamic 

pressure sensors installed in the steam generator is shown in Figure 2-6 and Figure 2-7.

Table 2-2 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

Flowrate

1 Outside of main steam line QV-MS1-01

2 Top of steam dome PT-SGSD1-01

3 Outlet region of steam separator TF-SGSD1-03

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

5 Top of economizer TF-SGRS1-03b

6 Bottom of downcomer region at 

cold side Dynamic-P-04

7 Economizer (wall side) Dynamic-P-03

8 Top of tubesheet on hot side Dynamic-P-05 TF-SGRS1-01a

9 Top of flow distribution plate

(wall side) Dynamic-P-02

10 Inside of feedwater box Dynamic-P-01

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

12 Inside of lower plenum from hot leg TF-SGP1-01

13 Inside of lower plenum to cold leg TF-SGP1-02

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

15 Cold leg side QV-CL1A-01B

QV-CL1B-01B
8 

Table 2-2 Uncertainty Level of Instruments

Figure 2-5 Locations of Measurement ID in the SG of the ATLAS
9 

Figure 2-6 Detailed Locations of Dynamic Pressure Sensors Installed in the Steam Generator

Figure 2-7 Measurement Location of Dynamic Pressure in the Break System
10

2.3 Experimental Procedure

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

steam generator feedwater line was simulated according to the following procedures:

- The ATLAS main data is logged.

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

acquire the dynamic pressure data in the steam generator.

- After several sec from start of the sub-DAS data logging, OV-MF1-03 and OV-MF1-04 valves 

are closed to prevent break flow rate distortion from the feedwater line during the break 

simulation.

- After several sec from closing of the OV-MF1-03 and OV-MF1-04 valves, nitrogen gas is 

supplied into the “Intermediate region” to rupture disk-2. Subsequently, disk-1 is opened by 

the driving force resulting from rupture of disk-2.

After acquiring experimental data over approximately 30 sec, Sub-DAS and the ATLAS logging are 

stopped.

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

Table 2-3 Sequence of Major Events

Events Time (sec) Description

Heat Up about 18,000 Heating up process

Test Standby about 600 Steady state condition

Test Start

0.0 ATLAS main DAS data logging start

600.0 Sub-DAS data logging start (Dynamic Pressure)

603.0 OV-MF1-03, 04 closing

Break 605.0 SG-1 feed line break 100% open

Test End 630.0 Data recording stop
11

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 feedwater line break of the 

steam generator. Three tests, named FLB-DS-01, FLB-DS-02 and FLB-DS-03, were performed using 

the double rupture disk assembly in order to simulate a sudden break of the steam generator 

feedwater line. The reason for performing FLB-DS-02 and FLB-DS-03 was to confirm the 

reproducibility of the FLB tests, and the reproducibility of the tests was confirmed by comparing the 

FLB test results. Therefore, only the FLB-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 three 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.
12

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

Design Parameter Target 

Value

Measured Value

Remark

(Sensor ID)

FLB-DS-01

(Value(1)

/STDEV(2))

FLB-DS-02

(Value(1)

/STDEV(2))

FLB-DS-03

(Value(1)

/STDEV(2))

Reactor Pressure Vessel

Normal Power (MW) 1.56 1.631 / 0.0006 1.628 / 0.0005 1.631 / 0.0005 Heat loss: 85 kW

Pressurizer Pressure 

(MPa) 15.5 15.51 / 0.034 15.56 / 0.014 15.55 / 0.002 (PT-PZR-01)

Core Inlet 

Temperature (°C)

290.7 289.01 / 0.23 290.25 / 0.19 290.15 / 0.23 (TF-LP-02G18)

Core Outlet 

Temperature (°C) 324.2 324.22 / 0.43 325.61 / 0.37 325.51 / 0.38 (TF-CO07G14,18,21,25)

Steam Generator

Net Thermal Power 

(MWt)

0.78 0.743 0.743 0.736 Heat loss: 30 kW

Steam Flow Rate 

(kg/s) 0.444 0.400 / 0.0010.430 / 0.001

0.400 / 0.001

0.430 / 0.001

0.400 / 0.004

0.427 / 0.003

SG-1 (QV-MS1-01)

SG-2 (QV-MS2-01)

Feed Water Flow Rate 

(kg/s) 0.444 0.421 / 0.0070.422 / 0.004

0.419 / 0.003

0.424 / 0.003

0.415 / 0.006

0.419 / 0.007

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

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

Feed Water 

Temperature (°C) 232.2 234.55 / 0.16233.91 / 0.16

234.75 / 0.12

233.95 / 0.12

234.40 / 0.15

233.57 / 0.15

SG-1 (TF-MF1-01)

SG-2 (TF-MF2-01)

Steam Pressure 

(MPa) 7.83 7.69 / 0.0037.69 / 0.003

7.83 / 0.003

7.83 / 0.003

7.83 / 0.015

7.82 / 0.009

SG-1 (PT-SGSD1-

01)

SG-2 (PT-SGSD2-

01)

Steam Temperature 

(°C) 293.5 294.24 / 0.09294.32 / 0.08

295.42 / 0.09

295.67 / 0.07

295.35 / 0.16

295.56 / 0.10

SG-1 (TF-SGSD1-

03)

SG-2 (TF-SGSD2-

03)

Primary Piping

Cold Leg Flow (kg/s) 2.0 2.19 / 0.08 2.21 / 0.09 2.20 / 0.09

Cold leg average

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

-01B)

Hot Leg Temperature 

(°C) 323.8 324.24 / 0.61 325.32 / 0.66 325.27 / 0.70 Hot leg average (TF-HL1,2-03A)

Cold Leg 

Temperature (°C) 289.9 290.92 / 0.68 292.13 / 0.61 292.09 / 0.65 Cold leg average 

(TF-CL1A,B-04A, 

TF-CL2A,B-04A)

Notes)

(1) Average Value at -10 to 600 sec

(2) STDEV: Standard Deviation at -10 to 600 sec
13

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 ATLAS

nodalization.

Figure 3-1 MARS-KS Code ATLAS Nodalization
14

To simulate the feedwater 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 feedwater line break is shown in Figure 3-2. The model 

is based on 95 volumes connected by 36 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 feedwater 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 32 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 feedwater break line (C720/C730) is modeled as a PIPE component, and a TIME 

DEPENDENT VOLUME component and has 14 volumes. A TIME DEPENDENT VOLUME 

component provides the boundary condition as the atmosphere and is connected by a VALVE 

component (J725).
15

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

17

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 RCS flow 

rate was a little different from that of experimental value. If we do a thermal power calculation using 

the measured inlet and outlet plenum temperatures, the RCS flow rate assuming 0.744 MW thermal 

power is 1.9 kg/sec, which is the same as the value predicted by the code. The calculated steam 

dome pressure was a little different from that of the 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 also a little different from that of the 

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, economizer feedwater and steam dome 

pressure controllers were de-activated and replaced by a relevant boundary condition for the transient 

analysis.
18

Table 4-1 Steady State Results

Parameters Measured(1) Calculated

Primary system

Hot leg pressure (MPa) 15.582±0.006 15.582

Inlet plenum temperature (K) 598.3±2.4 598.1

Outlet plenum temperature (K) 564.1±2.4 563.5

Cold leg flow rate (kg/s)(2) 2.1365±0.001 1.9

Secondary system

Thermal power (MW) 0.744 0.745

Steam dome pressure (MPa) 7.697±0.003 7.915

Steam dome temperature (K) 567.4±2.4 567.4

Steam flow rate (kg/s) 0.401±0.0002 0.42622

Economizer feedwater 

temperature (K) 507.7±2.4 507.7

Downcomer feedwater 

temperature (K) 506.8±2.4 506.8

Economizer feedwater flow rate 

(kg/s) 0.389±0.0002 0.3892

Downcomer feedwater flow 

rate (kg/s) 0.037±0.00002 0.037

Notes)

(1) Average value during 300 to 600 sec

(2) Average of cold legs A and B
19

4.2 Transient Calculation

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

was maintain for about 600 sec. The feedwater line break test was initiated by isolation of the main 

feedwater line at 602.6 sec and opening the break line valve at 604.04 sec. 

The RELPA5/MOD3.3 calculation was also conducted for null transient analysis during 600 sec. The 

break was initiated by isolation of the 702 and 703 TIME DEPENDENT JUNCTION components at 

602.6 sec and opening the 725 valve at 604.04 sec as in the FLB experiment (FLB-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 FLB 

Test

[FLB-DS01]

 (sec)

 RELAP5/MOD3.3 

Calculation

 (sec)

Remark

Steady state condition 0.0 to 600.0 0.0 to 600.0 null transient

Isolation of main feedwater line 602.6 602.6

702,703 

junction 

isolation

SG-1 feedwater line break open 604.04 604.04 725 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 604.0 to 604.2 sec. The RELAP5/MOD3.3 

calculation data were compared with experimental data as shown in Figure 4-1 through Figure 4-8. 

The experiment data are labeled “Exp.”, and the RELAP5/MOD3.3 calculation data as “Cal.”.
20

4.2.1 Flow Rate

The economizer feedwater lines were isolated in the RELAP5/MOD3.3 calculation and the experiment 

at 602.6 sec as presented in Figure 4-1. The downcomer feedwater flow rate in the experiment 

increased after isolation of the economizer feedwater lines as presented in Figure 4-2. Because, the 

meaningful thermal hydraulic parameters were acquired experimentally during a short period of time 

(approximately 0.1 to 0.2 sec), this difference is immaterial. 

Figure 4-1 Flow Rate at the Economizer Feedwater Line

Figure 4-2 Flow Rate at the Downcomer Feedwater Line
21

4.2.2 Temperature

The temperature at the steam dome calculated by the RELAP5/MOD3.3 was in good agreement with 

experimental data, as presented in Figure 4-3. The temperature at the top of the economizer and the 

top of the tubesheet at the hot side section was slightly over-predicted, as presented in Figure 4-4 and 

Figure 4-5. However, these differences were within the uncertainty band shown in Table 2-2. 

Figure 4-3 Temperature at the Steam Dome

Figure 4-4 Temperature at the Top of the Economizer
22

Figure 4-5 Temperature at the Top of the Hot Side Tubesheet 

4.2.3 Pressure

The steam dome pressure in the RELAP5/MOD3.3 calculation was determined from the table of TIME 

DEPENDENT JUNCTION. As shown in Figure 4-6, there is a difference between the 

RELAP5/MOD3.3 calculation and the experimental results. This is because the saturation pressure 

at the steam dome temperature was used as the steam dome pressure, as described in section 4.1.

Figure 4-6 Pressure at the Steam Dome
23

4.2.4 Dynamic Pressure

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

line break of the steam generator. Locations for the dynamic pressure measurements are shown in 

Figure 2-6. The maximum dynamic pressure was observed at dynamic pressure - 01 (Economizer 

Box) which was 0.157 MPa in experiment and 0.335 MPa in the RELAP5/MOD3.3 calculation, as 

presented in Figure 4-7. The dynamic pressure - 02 (Top of Economizer) was 0.085 MPa in 

experiment and 0.164 MPa in the REALP5/MOD3.3 calculation, as presented in Figure 4-8. 

Figure 4-7 Dynamic Pressure - 01 (the Economizer Box)

Figure 4-8 Dynamic Pressure - 02 (Top of the Economizer)
24

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 

the break valve discharge coefficient. 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 feedwater 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 4-9 shows the behavior of the dynamic pressure depending on the critical flow model. The 

dynamic pressure used the Henry-Fauske and the Modified Henry-Moody critical flow model were 

shown to be about the same and were more conservative than the Original RELAP critical flow model. 

Therefore, use of the Henry-Fauske and the Modified Henry-Moody critical flow model is 

recommended to obtain the most conservative maximum dynamic pressure for the feedwater line 

break of the steam generator. 
25

Figure 4-9 Dynamic Pressure - 01 for the Critical Flow Model
26

4.3.2 Break Valve Opening Time

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

Korea Institute of Nuclear Safety. In the test, the maximum break valve opening time was less than 

1.5 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 4-10 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 0.335 and 0.325 MPa, respectively.

Figure 4-10 Dynamic Pressure - 01 for the Break Valve Opening Time
27

4.3.3 Time Step Size

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

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

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

8 to 10-3 sec.

Figure 4-11 shows that the maximum dynamic pressure increased as the maximum time step size 

decreased to 10-7 sec, and that the dynamic pressure was almost the same, below that value. A

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 0.192 MPa (0.157 MPa in the experiment).

Figure 4-11 Dynamic Pressure - 01 for the Time Step Size
28

4.3.4 Break Valve Discharge Coefficient

The break valve discharge coefficient is an important factor in the feedwater line break analysis. Thus, 

a sensitivity analysis of the break valve discharge coefficient was performed. The analysis ranged 

from 0.4 to 1.0.

Figure 4-12 shows the behavior of the dynamic pressure depending on the break valve discharge 

coefficient. The maximum dynamic pressure increased as break valve discharge coefficient increased. 

The break valve discharge coefficient of 1.0 is recommended to obtain the most conservative 

maximum dynamic pressure. 

Figure 4-12 Dynamic Pressure - 01 for the Break Valve Discharge Coefficient
29

5 CONCLUSIONS

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

were compared to the experimental data. As a result, the pressure and temperature 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 feedwater line break accident.

31

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., “ATLAS Facility and Instrumentation Description Report,” KAERI/TR-3779/2009, 

Korea Atomic Energy Research Institute, 2009.

3. K. H. Kang, et al., “Final Report for F.W.L.B Test of the Steam Generator,” ATLAS-DS-SGT-09-

008, Rev.1, Korea Atomic Energy Research Institute, 2009.

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.



November 2016 Feedwater Line Break Analysis Using RELAP5/MOD3.3 for Steam Generator 

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-0473

Technical



NUREG/IA-0473 Feedwater Line Break Anaylsis Using RELAP5/MOD3.3 for Steam Generator 

Blowdown Load Assessment

November 2016