NUREG/IA-0447 

International 

 Agreement Report

RELAP5/MOD3.3 Assessment by 

Comparison with PKL III G3.1 Experiment 

(small break in the main steam line)

 

Prepared by:

A.Zaruba, K. Krezeminski, A. Benkert

AREVA-NP

Germany.

 

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: April 2014

Date Published: July 2014

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission
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NUREG/IA-0447 

International 

 Agreement Report

RELAP5/MOD3.3 Assessment by 

Comparison with PKL III G3.1 Experiment 

(small break in the main steam line)

 

Prepared by:

A.Zaruba, K. Krezeminski, A. Benkert

AREVA-NP

Germany.

 

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: April 2014

Date Published: July 2014

Prepared as part of 

The Agreement on Research Participation and Technical Exchange

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

Published by

U.S. Nuclear Regulatory Commission

iii

ABSTRACT

The behavior of the PKL facility during a small main-steam-line break was investigated by 

using the RELAP model of the main facility components, including the reactor pressure 

vessel, steam generators, pressurizer, primary piping, and main steam lines. The 

calculations were performed using the computer code RELAP5/MOD3.3. 

The main objectives of the study are to compare the simulation results with experimental 

data and to assess the accuracy of the calculated data. The major parameters and 

phenomena for simulating a main steam line break, e.g. primary and secondary pressures 

and temperatures, break flow rate and water level in SG and PZR, were analyzed. The 

results of the calculations were found to be in good agreement with experimental results. 

All the analytical activities were performed as part of the Organization for Economic 

Cooperation and Development/Nuclear Energy Agency (OECD/NEA) projects of Test G3.1.

v 

CONTENTS

Page

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

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

TABLES..............................................................................................................................vii

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

ABBREVIATIONS ...............................................................................................................xi

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

2. FACILITY AND TEST DESCRIPTION .......................................................................... 3

2.1. PKL III test facility .................................................................................................. 3

2.2. Boundary conditions: status of the facility before test start..................................... 3

2.3. G3.1 Test: Accident description and sequence of events....................................... 5

3. RELAP5/MOD3.3 CALCULATIONS ............................................................................. 7

3.1. Numerical code and RELAP5/MOD3.3 input deck preparation .............................. 7

3.2. RELAP5/MOD3.3 model of PKL III facility.............................................................. 7

3.3. Pre-calculation of the break flow rate and the break model..................................12

3.4. Boundary conditions and steady-state initialization ..............................................13

3.5. RELAP5/MOD3.3 results: diagrams of the most representative 

parameters..........................................................................................................14

4. DISCUSSIONS OF RESULTS ...................................................................................19

4.1. Break flow rate.....................................................................................................19

4.2. Secondary-side pressure and SG water level ......................................................19

4.3. RPV outlet fluid temperature................................................................................19

4.4. Pressurizer pressure and PZR water level ...........................................................20

5. CONCLUSION ............................................................................................................21

6. REFERENCES............................................................................................................23

vii

FIGURES

Page

Figure 1: PKL facility (AREVA NP, Erlangen).................................................................... 4

Figure 2: Status of PKL test facility before test start.......................................................... 4

Figure 3: Reactor pressure vessel model........................................................................10

Figure 4: Steam generator and primary circuit (Loop 10) model......................................11

Figure 5: Nodalization of PKL III facility...........................................................................12

Figure 6: Secondary side pressure and mass flow rate at the leak compared with 

experimental data ...........................................................................................13

Figure 7: Break mass FLOW RATE/ Integral mass released via break ...........................15

Figure 8: Pressure in SG 10 / SG 10 water level.............................................................15

Figure 9: PZR pressure (inlet region) / PZR water level ..................................................16

Figure 10: RPV outlet temperature (Loop10) / RPV inlet temperature 

(Loop 10 and Loop 20)....................................................................................16

Figure 11: Primary circuit flow rate (Loop 10) / Primary circuit flow rate (Loop 20)............17

TABLES

Page

Table 1: Initial conditions at test start............................................................................... 5

Table 2: G3.1 test information (sequence of major events).............................................. 6

Table 3: RELAP5/MOD3.3 input deck: primary piping and SG components..................... 8

Table 4: Input deck resources and nodalization features ................................................. 9

Table 5: Agreement of Boundary conditions (PKL / RELAP5/mod3.3) ...........................14

ix

EXECUTIVE SUMMARY

The main purpose of the report is to assess the capability of RELAP5/MOD3.3 computer 

code to predict the response of PKL III facility under simulated accidental conditions (PKL III 

Test G3.1: small break in the main steam line).For this purpose, a RELAP5/MOD3.3 

analysis of the experiment is performed and the results of the calculation are compared with 

the measured data. 

The PKL III test facility (AREVA GmbH, Erlangen, Germany) was built to investigate the 

behavior of German pressurized-water reactors under accident and transient conditions. 

A RELAP5 input deck that had been used for earlier PKL analyses was used as the basis 

for the RELAP5/MOD3.3 model of the PKL III facility. 

All calculations were performed in two steps: a steady-state calculation (in order to stabilize 

the computational process), and a transient calculation. 

The calculated values of the most important parameters and phenomena for simulating a 

main steam line break (e.g., primary and secondary pressure, fluid temperature, break flow 

rate and the mass flow rate in the primary circuit) were found to be in good agreement with 

the experimental results. 

A deviation between the calculation and the experimental results was found for the water 

level in the affected steam generator and for the collapsed level in the pressurizer.

xi

ABBREVIATIONS

ECCS Emergency Core Cooling System

MSIV Main Steam Isolation Valve

MSL Main Steam Line

MSLB Main Steam Line Break

MSRV Main Steam Relief Valve

NPP Nuclear Power Plant

PZR Pressurizer

RCP Reactor Coolant Pump

RCS Reactor Coolant System

RPV Reactor Pressure Vessel

PWR Pressurized-Water Reactor

SG Steam Generator

1 

1. INTRODUCTION

The behavior of the PKL facility during a small main-steam-line break was investigated 

using the RELAP model of the main facility components, including the reactor pressure 

vessel (RPV), steam generators (SG), pressurizer (PRZ), primary piping, and main 

steam lines. The calculations were performed using the computer code 

RELAP5/MOD3.3. 

The main objectives of the calculation are to compare the results with experimental 

data obtained during the PKL G3.1 test and to assess of the accuracy of the calculated 

data. The most important parameters and phenomena for simulating a main steam line 

break, e.g., primary and secondary pressures and temperatures, break flow rate and 

water level in SG and PZR, were analyzed.

3 

2. FACILITY AND TEST DESCRIPTION

2.1. PKL III test facility

The PKL III test facility (AREVA NP GmbH, Erlangen) was built to investigate the 

behavior of German pressurized-water reactors (PWRs) under accident and transient 

conditions [2]. The layout of the PKL III facility (Figure 1) is based on the “Vorkonvoi“ 

type (4-Loop, 1300 MWe) of a KWU pressurized-water reactor, with the Philippsburg 2 

nuclear power plant (NPP) serving as the reference plant. The entire reactor coolant 

system (RCS) primary side and the most significant components of the secondary side 

(excluding turbines and condenser), including the appropriate control system, are 

represented.

Following the scaling concept, all geodetic heights are represented in a 1:1 ratio. The 

entire volume of the primary circuit and the proportions of the individual volumes are 

scaled by a factor of 1:145. The maximum primary pressure is limited to 4.5 MPa; the 

maximum core power of 2.5 MW is equivalent to 10 % of the nominal rate. 

The core is simulated with a bundle of 314 electrically heated rods. The core geometry 

is, like the SG geometry, constructed as an “actual section;” that is, the individual 

heated rods and U-tubes have the actual geometry, but the number of heated rods in 

the core and the number of U-tubes in the SG are reduced by the scaling factor 1:145 

(volume and power scaling), as compared to the reference plant.

The representation of the primary side by four identical and symmetric loops arranged 

around the reactor pressure vessel allows the realistic investigation of accidents, even 

with asymmetric boundary conditions in the individual loops. Through the 

representation of all significant interfaces and auxiliary system functions on the primary 

and secondary side, the overall system behavior as well as the interaction between the 

individual systems can be investigated under a wide range of accident conditions and 

the effectiveness of the automatic or manually-performed counter actions can be 

ascertained [1]. 

With approximately 1500 measurement points, the PKL facility is comprehensively 

instrumented. This allows detailed analysis and interpretation of the phenomena that 

develop in the course of the tests. 

2.2. Boundary conditions: status of the facility before test start

The status of the facility before the accident is summarized in Table 1 and graphically 

presented in Figure 2. The primary circuit is filled with water and the pressurizer 

pressure is maintained at 42 bar. The loop flow rate in the primary circuit is set to the 

33.7 kg/s and the core power is 260 kW. The fluid temperature in the core inlet / exit is 

243°C / 245°C respectively. All steam generators are isolated from the feedwater lines 

during the last stage of the preparation phase. 

The core power of 260 kW includes compensation for appropriate 150kW heat losses 

of the primary and secondary side components. The remaining power describes the 

decay heat power and leads to a slow and continuous temperature increase in the 

primary circuit (~ 12°C/h) before starting the test.
4 

Figure 1: PKL facility (AREVA NP, Erlangen)

Level 12.2 m Level 9.2 m Level 12.2 m Level 12.2 m

Level 7.4 m

RCP flow rate: 33.7 kg/s RCP flow rate: 33.7 kg/s

SG pressure: 35 bar SG pressure: 35 bar

PZR 

pressure:

42 bar

Fluid temperature - 246 °C 

(core exit region)

Core power:

260 kWe

Figure 2: Status of PKL test facility before test start 
5 

Table 1: Initial conditions at test start

Primary circuit Condition / Value

Heater rod bundle power (decay 

heat) 260 kW

RCPs in operation Mass flow rate: 33.7 kg/s

PZR pressure 42 bar

Collapsed level in PZR 7.4 m

Fluid temperature (core exit) 245°C

Flow pattern Symmetrical heat removal via 4 SGs with 

subcooled forced circulation in all 4 loops

Coolant inventory Primary circuit completely filled with water

Secondary side

Main steam pressure SG 1-4: 35 bar

Water temperature (Riser) SG 1-4: 240°C

Collapsed level (Riser) SG1: 9.2 m*), SG 2 - 4: 12.2 m

*) Lower collapsed level in affected SG1 to avoid of entrained droplets 

2.3. G3.1 Test: Accident description and sequence of events

The G3.1 test can be divided into 2 major phases:

Phase1: starts with opening of the break and ends when the secondary side of 

the affected SG 10 becomes empty 

Phase 2: starts with the initiation of the emergency core cooling system 

(ECCS) injection into 2 cold legs and includes the pressure limitation (control) 

by the PZR safety valves.

The following discussion applies only to the first phase of the experiment. 

The investigated case is a small break in the main steam line connected to SG 10. The 

break (break size is 0.1A) is initiated by opening the break valve. The feedwater lines 

in all SG have been isolated before starting the test. Immediately after the break 

opens, the coast down of all four reactor coolant pumps is initiated and the nonaffected steam generators are automatically isolated by closing the isolation valves.

The break leads to the fast depressurization of the affected steam generator. The 

temperature difference between primary and secondary side increases and 

considerably improves the heat transfer in the SG. As a result, the pressure and the 

temperature in the primary circuit start to decrease. The water level in the affected SG 

decreases due to the intensive evaporation. Concurrently, the water level in the 

pressurizer also decreases due to the volume contraction. 

After the coast down of the main coolant pumps, the heat removal from the core is 

provided using natural circulation. The natural circulation flow in loop 10 is greater due 

to the increased heat transfer from the primary to the secondary side in the affected 

SG. 

The secondary water level in the affected SG further decreases and eventually the 

affected SG becomes empty. The fluid temperature at the RPV inlet and the PZR 

water level reach their minimum values and start increasing again. 
6 

The information about the major events, measurements and essential system 

response during the test is given in Table 2. 

Table 2: G3.1 test information (sequence of major events)

Time (s) Event Remarks

0 

Isolation of SG 20, 30, and 40

Start of the test:

Opening of the break-valve in the

main steam line of SG10

Coast down signal for all RCPs

PZR heaters are switched off

Evaporation in SG 10 → decrease in

secondary side pressure and water level

Decrease of the RCP speed → decreased

mass flow rate

210 Blockage of RCPs Natural circulation in each loop

600 Minimum temperature (153 °C) at RPV 

inlet of loop 10

1000 End of phase 1

Complete evaporation of secondary 

inventory in the affected SG

Further decrease of the natural circulation

Increase of the primary and secondary 

side temperature and PZR water level 
7 

3. RELAP5/MOD3.3 CALCULATIONS

3.1. Numerical code and RELAP5/MOD3.3 input deck preparation

The RELAP5/MOD3.3 computer code is used for the calculations. This advanced 

computer code is designed for use in realistic studies of accident thermal-hydraulics in 

pressurized-water reactors. The basis for the RELAP5/MOD3.3 model of PKL III facility 

was a RELAP5 input deck which had already been used for earlier PKL analyses [3], 

[4] and was adapted for the main steam line break case [1]. 

All calculations were performed in two steps: a steady-state calculation in order to 

stabilize the computational process and the transient calculation after the opening of 

the break-valve. The details are presented in the following chapters.

3.2. RELAP5/MOD3.3 model of PKL III facility

The analysis uses a full four-loop model of the facility developed by AREVA NP GmbH 

[4]. 

The reactor pressure vessel model contains the vessel with its major components: the 

reactor core, the downcomer, the lower and the upper plenum and the RPV head 

(Figure 3). 

The heat production in the core is realized by a heat structure connected to six pipe 

volumes. Additional heat structure models are used to simulate the behavior of the 

vessel wall and the main internals (see Figure 3 for details).

The following components, numbered from 20 to 256, represent the reactor pressure 

vessel.

B20 and B22 ** Lower plenum

P42 Reactor core (core simulator)

A40 Core bypass

B102 and B112 Core exit region

P130 and P220 Upper plenum and upper head

B242, B244, P246, and P256 Downcomer

B232 and B234 Inlet nozzles

B122 and B124 Outlet nozzles

B230 and B222 Upper head bypass

**  P - PIPE component, B - BRANCH component, A - ANNULUS component

The steam generator and primary piping model is given in Figure 4. The components 

which represent the steam generator and the primary piping are presented in Table 3.
8 

Table 3: RELAP5/MOD3.3 input deck: primary piping and SG components

Hot leg Crossover leg RCP Cold leg

Loop 10 P300 and P310 P324, B326, and P328 PP332 and SV330 P338

Loop 20 P400 and P410 P424, B426, and P428 PP432 and SV430 P438

Loop 30 P500 and P510 P524, B526, and P528 PP532 and SV530 P638

Loop 40 P600 and P610 P624, B626, and P628 PP632 and SV630 P638

SG Inlet U-tubes SG Outlet

SG10 

(primary) 

B314 P316, P318, and P320 B322

SG20 

(primary) 

B414 P416, P418, and P420 B422

SG30 

(primary) 

B514 P516, P518, and P520 B522

SG40 

(primary) 

B614 P616, P618, and P620 B622

SG riser SG downcomer Steam dome

SG10

(secondary)

P350 and B372 B362, B364, B366, P368, and B370 P352 and P354

SG20 

(secondary)

P450 and B472 B462, B464, B466, P468, and B470 P452 and P454

SG30 

(secondary)

P550 and B572 B562, B564, B566, P568, and B570 P552 and P554

SG40 

(secondary)

P650 and B672 B662, B664, B666, P668, and B670 P652 and P654

The main reactor coolant pumps (RCPs) are modeled by a special RELAP5/MOD3.3 

PUMP component. The primary loops are symmetrical except for loop 20, where the 

pressurizer surge line is connected. The pressurizer and PZR surge line are 

represented by P800, B805, and P810 components. The SG main steam lines are 

modeled by B710, B720, B730, and B740. 

Figure 5 presents the nodalization scheme for the whole facility. The PKL III model 

consists of 146 hydrodynamic components connected by 55 junctions. The facility 

structures are represented by 145 heat structures with 655 mesh points. The detailed 

information about the adopted code resources and the nodalization features are given 

in Table 4.
9 

Table 4: Input deck resources and nodalization features

ADOPTED CODE RESOURCES

Total number of hydraulic components (whole facility) 

PIPE / BRANCH / ANNULUS / SNGLVOL / TMDPVOL

146

58 / 65 / 1 / 4 / 18

Total number of hydraulic components in RPV 18

Total number of hydraulic components in SG 

(per SG; primary and secondary side) 56

Total number of heat structures 145

Total number of mesh points in the heat structures 655

Total number of 

SNGLJUN / TMDPJUN / VALVE-connections 18 / 21 / 16

NODALIZATION FEATURES

Number of modeled loops 4 

Number of components modeling the DC in RPV 2 Pipes, 

6 Branches

Number of U-tubes (per SG ) 3 

Number of axial meshes of SG U-tubes (per SG) 60

Core model (1-D or 3-D component) 1-D 

Number of hydraulic channels in the core region 1 

Cross-flow junctions between parallel channels in the core 

(YES / NO) NO

CODE OPTIONS

Break flow model Henry-Fauske

critical flow model **

SEPARATOR or DRYER models in SG dome (YES / NO) NO

Specific models activated in PZR (YES / NO) NO

** - for more details see § 3.3 
10

B-222

SJ-43

Loop20-K

Loop30-K

Loop20-H

Loop10-H Loop40-H

Loop30-H Loop10-K Loop40-K

B-234

DC-Pipe1 DC-Pipe2 Core Bypass

B-20 Lower Plenum

Head Bypass

P-42

1

2

3

4

5

6

7

8

A-40

1

2

3

P-246

3

1

2

4

5

P-256

3

1

2

4

5

Upper Plenum Lower Part

B-22 Lower Plenum

B-242

B-230

SJ-235: Cross Fl. B-232 to B-234 SJ-235: Cross Fl. B-232 to B-234

B-244 Upper Plenum Lower Part

P-130

1

2

3

Upper Plenum

B-232

B-220

Upper Head

B-222

Head Bypass

B-230

B-242

B-244

B-242

B-244

B-242

B-244

1

1

3 4

2

1

2

1

2

1

1

1

2 3

1

2 3

1

1

1

2

B-112

1

2 2

B-102

B-122 B-124

SJ-125

Figure 3: Reactor pressure vessel model 
11

10

Steam Generator 10

 

B

HPIS

MSIV

MSRV 

P

PP

SJ

SV

TJ

TV

VLV 

Branch

High Pressure Injection System

Main Steam Isolation Valve

Main Steam Relief Valve

Pipe

Pump

Single Junction

Single Volume

Time Dependent Junction

Time Dependent Volume

Valve

P-310 P-324

SJ-305

1

2

3

4

5

6

P-328

5

1

2

3

4 6

7

8

SJ-329

P-338 Pump

Loop 10

B-364

B-366

B-362

P-316

P-318

P-320

1 1 1 222018

9

B-314 B-322

P-350

B-352

B-354

1

2

3

4

5

6

7

8

9

11

SV-710

SV-750

TV-754 TV-762

VLV-752

B-760

VLV-716

MSRV

VLV-712

MSIV

Containment

Main Steam

Bypass Station

Header

5

2

3

4B-326

SV-330

PP-332

P-368

1

2

3

4

5

6

7

9

8

P-300

1 2 3 4 5 6 7

Turbine

B-372

B-372

B-370

11

10

Feedwater

TV-390

TJ-391

5 4 3 2 1

B122 J-2

1 2

1

1

4 3 2

4

1 2 3

VLV-726

VLV-736

VLV-746

1

2

1

1

B232 J-1

VLV-902

TV-956

TV-950

HPIS

TJ-957

TJ-951

P-952

1

B122 B-954

TJ-301

1

1

2 3

VLV-371

Figure 4: Steam generator and primary circuit (Loop 10) model
12

B-222

SJ-125

SJ-43

SJ-235

B-234

Core Bypass

B-20

Lower Plenum

Head Bypass

P-42

1

2

3

4

5

6

7

8

A-40

1

2

3

B-220

Head

P-246

3

1

2

4

5

P-256

3

1

2

4

5

B-102

Upper Plenum Lower Part

B-22

Lower Plenum

B-242

B-230

SJ-235: Cross Fl. B-232 to B-234 SJ-235: Cross Fl. B-232 to B-234

B-244

B-112

Upper Plenum Lower Part

P-130

1

2

3

Upper Plenum

B-124 B-122 B-232

10

Steam Generator 10

P-310 P-324

SJ-305

1

2

3

4

5

6

P-328

5

1

2

3

4 6

7

8

SJ-329

P-338 Pump

Loop 10

B-364

B-366

B-362

P-316

P-318

P-320 1 1 1 222018

9

B-314 B-322

P-350

B-352

B-354

1

2

3

4

5

6

7

8

9

11

SV-710

SV-750

TV-754

TV-762

VLV-752

B-760

VLV-716

MSRV

VLV-712

MSIV

Containment

Main Steam

Bypass Station

Header

TV-392

TJ-393

1

5

2

3

4

B-326

SV-330

PP-332

P-368

1

2

3

4

5

6

7

9

8

P-300

1 2 3 4 5 6 7

Turbine

B-372 B-372

B-370

11

10

TV-950 TV-956

HPIS

TJ-951 TJ-957

B-952

5 4 3 2 1

Containment

TV-908

VLV-904

Break Location

10

Steam Generator 40

P-610 P-624

SJ-605

1

2

3

4

5

6

P-628

5

1

2

3

4 6

7

8

SJ-329

P-638 Pump

Loop 40

B-664

B-666

B-662

P-616

P-618

P-620 1 1 1 222018

9

B-614 B-622

P-650

B-652

B-654

1

2

3

4

5

6

7

8

9

11

SV-740

SV-750

TV-754

TV-762

VLV-752

B-760

VLV-746

MSRV

VLV-742

MSIV

Containment

Main Steam

Bypass Station

Header

TV-692

TJ-693

1

5

2

3

4

B-626

SV-630

PP-632

P-668

1

2

3

4

5

6

7

9

8

P-600

1 2 3 4 5 6 7

Turbine

B-672 B-672

B-670

11

10

5 4 3 2 1

10

Steam Generator 20

P-424 P-410

SJ-405

1

2

3

4

5

6

P-428

5

1

2

3

4 6

7

8

SJ-429

Pump P-438

Loop 20

B-464

B-466

B-462

P-416

P-418

P-420

22 20 18 111

9

B-422 B-414

P-450

B-452

B-454

1

2

3

4

5

6

7

8

9

11

SV-720

SV-750

TV-754

VLV-752

B-760

VLV-726

MSRV

VLV-722

MSIV

Main Steam

Bypass Station

Header

TV-492

TJ-493

1

5

2

3

4B-426

SV-430

PP-432

P-468

1

2

3

4

5

6

7

9

8

P-400

45 3 2 1 6 7

Turbine

B-472 B-472

B-470

11

10

1 32 4 5

TV-966 TV-960

HPIS

TJ-967 TJ-961

B-962

TV-762

Containment

P-810

B-805

1

2

3

5

6

8

7

TV-820

VLV-815

TV-892

TJ-891

TJ-893

TV-896

VLV-895

3 2 1

6 45

78

P-800

SJ-803

VLV-801

TV-890

10

Steam Generator 30

P-524 P-510

SJ-505

1

2

3

4

5

6

P-528

5

1

2

3

4 6

7

8

SJ-529

Pump P-538

Loop 30

B-564

B-566

B-562

P-516

P-518

P-520

22 20 18 111

9

B-522 B-514

P-550

B-552

B-554

1

2

3

4

5

6

7

8

9

11

SV-730

SV-750

TV-754

TV-762

VLV-752

B-760

VLV-736

MSRV

VLV-732

MSIV

Containment

Main Steam

Bypass Station

Header

TV-592

TJ-593

1

5

2

3

4B-526

SV-530

PP-532

P-568

1

2

3

4

5

6

7

9

8

P-500

45 3 2 1 6 7

Turbine

B-572 B-572

B-570

11

10

1 32 4 5

Feedwater

TJ-391

(STST)

TV-390

Feedwater

TJ-691

(STST)

TV-690

Feedwater

TJ-491

(STST)

TV-490

Feedwater

TJ-591

(STST)

TV-590

Figure 5: Nodalization of PKL III facility

3.3. Pre-calculation of the break flow rate and the break model

An additional study was performed to validate the break flow rate and the behavior of 

the steam generator during the depressurization phase after the opening of the break 

valve without an influence of the heat transfer from the primary to the secondary side. 

The calculation results were compared with the experimental data obtained during the 

depressurization test performed at the PKL III facility. The main purpose of this 

analysis is an assessment of the break model, which can then be used in the 

calculation of the small MSLB accident. 

The results of the RELAP5/MOD 3.3 calculation and the experimental data obtained 

during the PKL depressurization test are presented in Figure 6. By varying the 

discharge coefficient of the break valve and the valve-opening time, the 

computationally determined pressure in the SG and mass flow at the leak were 

adjusted to meet the experimental results as closely as possible. 

The default Henry-Fauske model for two-phase critical flow is used in the calculation. 

The adjusted discharged coefficient is set to 0.8.
13

 

 

 

0. 200. 400. 600. 800. 1000.

0.

10.

20.

30.

40.

bar

Relap

PKL

0. 200. 400. 600. 800. 1000.

Time(s)

0.

1.

2.

3.

kg/s

Relap5

PKL

Figure 6: Secondary side pressure and mass flow rate at the leak compared with 

experimental data 

3.4. Boundary conditions and steady-state initialization

The initial conditions of the facility does not represent a “pure” steady-state situation 

since the heat losses in PKL III facility are smaller than the power supplied to the core 

simulator (§ 2.2). As a result, the temperature in the primary circuit constantly 

increases. For this reason, the initial conditions for the calculation were also adjusted 

to produce a “quasi-steady-state”, which would approximate the behavior of the PKL III 

facility. The computation of the main steam line break was started when primary-side 

temperatures of 243.6 °C and pressure of 42.3 bar were reached.

The major parameters representing the initial conditions used in RELAP5 calculations 

are given in Table 5.
14

Table 5: Agreement of Boundary conditions (PKL / RELAP5/mod3.3)

PKL RELAP5/MOD3.3 Value

Primary circuit

Heater rod bundle 

power

(decay heat)

260 kW 260 kW

RCPs in operation Mass flow rate: 33.7 kg/s Mass flow rate: 34.3 kg/s

PZR pressure 42 bar 42.3 bar

Collapsed level in PZR 7.4 m 7.4 m

Fluid temperature (core 

exit)

245°C 243.6°C

Secondary side

Main steam pressure SG 1-4: 35 bar SG 1-4: 35 bar

Temperature (Riser) SG 1-4: 240°C SG 1-4: 242.5°C

Collapsed level (Riser)

SG1: 9.2 m**, SG 2 - 4: 

12.2 m

SG 1: 9.2 m, SG 2 - 4: 12.2 

m 

** - lower collapsed level in affected SG1 to avoid of entrained droplets 

3.5. RELAP5/MOD3.3 results: diagrams of the most representative 

parameters

The main steam line break computation began with opening of the break valve in the 

main steam line of the SG 10, simultaneously isolating the non-affected SGs and 

switching off the main coolant pumps. On the secondary side, the feedwater pumps 

were shut off. Core heaters continuously supplied a power of 260 kW inclusive heat 

losses. The pressurizer remained connected during the test. 

The diagrams of the most representative parameters of the RELAP5/MOD3.3 

calculation together with experimental data are given in Figures 7-11.
15

 

 

 

0. 200. 400. 600. 800. 1000.

0.

1.

2.

3.

4.

KG/S

RELAP5

PKL

0. 200. 400. 600. 800. 1000.

TIME (S)

0.

100.

200.

300.

400.

500.

KG

RELAP5

PKL

Figure 7: Break mass FLOW RATE/ Integral mass released via break 

 

 

0. 200. 400. 600. 800. 1000.

0.

10.

20.

30.

40.

BAR

RELAP5

PKL

0. 200. 400. 600. 800. 1000.

TIME (S)

0.

2.

4.

6.

8.

10.

M

RELAP5

PKL

Figure 8: Pressure in SG 10 / SG 10 water level
16

 

 

 

0. 200. 400. 600. 800. 1000.

36.

38.

40.

42.

44.

BAR

RELAP5

PKL

0. 200. 400. 600. 800. 1000.

TIME (S)

0.

2.

4.

6.

8.

10.

M

RELAP5

PKL

Figure 9: PZR pressure (inlet region) / PZR water level 

 

 

0. 200. 400. 600. 800. 1000.

220.

230.

240.

250.

°C

RELAP5

PKL

0. 200. 400. 600. 800. 1000.

TIME (S)

140.

160.

180.

200.

220.

240.

260.

°C

RELAP5_10

PKL_10

PKL_20

RELAP5_20

 Figure 10: RPV outlet temperature (Loop10) / RPV inlet temperature 

(Loop 10 and Loop 20)
17

 

 

 

0. 200. 400. 600. 800. 1000.

0.

10.

20.

30.

40.

KG/S

RELAP5

PKL

0. 200. 400. 600. 800. 1000.

TIME (S)

0.

10.

20.

30.

40.

KG/S

RELAP5

PKL

Figure 11: Primary circuit flow rate (Loop 10) / Primary circuit flow rate (Loop 20)

19

4. DISCUSSION OF RESULTS 

4.1. Break flow rate

The predicted break flow rate is found to be in good agreement with the experimental 

results during the depressurization of the SG10 only and with all facility components, 

as presented in Figure 6 and Figure 7. A small discrepancy between the experiment 

and the calculation is observed between 180 and 230 seconds, where the predicted 

break flow rate is rather higher than that from the experiment (see Figure 7). The 

difference between the curves becomes significant when the SG 10 in the RELAP5 

calculation empties (after approx. 760 seconds) and the calculated break flow rate 

reduces to zero.

4.2. Secondary-side pressure and SG water level

The secondary side pressure and the water level in the affected SG are presented in 

Figure 8. The opening of the break valve causes a fast pressure drop and decrease of 

the collapsed level in the affected SG 10. 

The calculated secondary side pressure is close to that measured experimentally 

during first 80 seconds of the transient. RELAP5 slightly over-predicts the secondary 

side pressure in the time between 80 and 650 seconds. After approximately 720 

seconds, the secondary side pressure from the RELAP5 calculation drops below the 

experimental curve after the SG 10 becomes empty.

The water level in the affected SG decreases rapidly after break opening due to the 

intensive evaporation. The results are found to be in a good agreement with the 

measured data during 200 seconds after the beginning of the transient. After that, the 

water level in the RELAP5 calculation becomes lower than that in the experiment. This 

can be explained by the fact that the RELAP5 slightly over-predicts the evaporation in 

the affected SG. Correspondingly, the break flow rate in the calculation slightly 

exceeds the measured data after approx. 200 seconds, as presented in Figure 7.

4.3. RPV outlet fluid temperature

The RPV inlet / outlet fluid temperatures and their comparison with experimental data 

are presented in Figure 10.

The calculated RPV outlet temperature is in good agreement with the experimental 

results during approx. 250 seconds after the beginning of the transient. During this 

time, the initial difference between the measured and the predicted curves remains 

virtually the same. After this, the cooldown on the primary side in the RELAP5 

calculation occurs slowly. This can be explained by the fact that the water level in the 

SG in the RELAP5 calculation is smaller, and therefore, the decay heat removal from 

the primary circuit is less effective.

After approximately 850 seconds, the RPV outlet temperature starts increasing again 

as the heat removal by the affected SG terminates. This is consistent with the 

experimental observations.
20

4.4. Pressurizer pressure and PZR water level

The fast depressurization of the affected SG causes evaporation and associated 

turbulent mixing on the secondary side. The latter improves the heat transfer from the 

primary to the secondary side. This leads to cooling of the water on the primary circuit 

(cooldown transient). The water density on the primary side increases with decreasing 

temperature, and the primary pressure decreases (Figure 9). 

The RELAP5 PZR pressure is in a good agreement with the experimental results. A 

small difference between two curves is observed after approx. 760 seconds. 

The predicted water level in the PZR matches the experimental data most closely in 

the initial phase of the transient (<100 sec). Later on, the RELAP5 over-predicts the 

measured water level in the pressurizer. The discrepancy between the results 

becomes larger after approx. 760 seconds as SG 10 empties. This is associated with 

discrepancy of calculated primary temperatures.
21

5. CONCLUSION

The main purpose of the report is to assess the capability of RELAP5/MOD3.3 

computer code to predict the response of PKL III facility under simulated accidental 

conditions (PKL III test G3.1: small break in the main steam line). For this purpose, a 

RELAP5/MOD3.3 analysis of the experiment is performed and the results of the 

calculation are compared with the measured data. 

Comparative analysis between the RELAP5 calculations and the measured results 

shows that the RELAP5/MOD3.3 code is able to predict the thermal-hydraulic 

parameters of the PKL facility with a sufficient level of precision. The calculated values 

of the most important parameters and phenomena for simulating a main steam line 

break, primary and secondary pressure, fluid temperature, break flow rate and the 

mass flow rate in the primary circuit, are found to be in good agreement with the 

experimental results. 

Deviation between the calculation and the experimental results was noted for the water 

level in the affected SG10 and for the collapsed level in the pressurizer. RELAP5 

slightly under-predicts the water level in the SG and over-predicts the collapsed level 

in PZR during the transient. This was associated with discrepancy of calculated 

primary temperatures. 

23

6. REFERENCES

1. Dennhardt, L., and Umminger, K. “PKL Versuch zum Frischdampf-Leckleitung.”

Jahrestagung Kerntechnik, Technical Center, Berlin, 2010 (in German). 

2. Kremin, H., Limprecht, H., Güneysu, R., and Umminger, K. “Description of the 

PKL III Test Facility.” Technical Center of Framatome ANP, 2001. 

3. Depisch, F. “S-RELAP5 Code Manual: Code Description, System Models and 

Solution Methods.” Siemens Technical Report KWU NDS1/95/E1510 

(Restricted) 1995. 

4. Wegner, R., “PKL III: RELAP5/Mod3 Input-Model.” Framatome-ANP Summary 

of Data NGES1/2002/en/0059 (Restricted) 2002. 




UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

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

OFFICIAL BUSINESS
NUREG/IA-0447 RELAP5/MOD3.3 Assessment by Comparison with PKL III G3.1 Experiment 

(small break in the main steam line)