NUREG/IA-0200

International 

Agreement Report 

Assessment Study on the 

PMK-2 Total Loss of 

Feedwater Experiment 

Using RELAP5 Code 

Prepared by 

J. Bdndti 

Lappeenranta University of Technology 

Department of Energy Technology 

P.O. Box 20 

FIN-53851 Lappeenranta, Finland 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

March 2001 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
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NUREG/IA-0200

International 

Agreement Report 

Assessment Study on the 

PMK-2 Total Loss of 

Feedwater Experiment 

Using RELAP5 Code 

Prepared by 

J. Biniti 

Lappeenranta University of Technology 

Department of Energy Technology 

P.O. Box 20 

FIN-53851 Lappeenranta, Finland 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

March 2001 

Prepared as part of 

The Agreement on Research Participation and Technical Exchange 

under the International Code Application and Maintenance Program (CAMP) 

Published by 

U.S. Nuclear Regulatory Commission
ABSTRACT

The main objective of the present report is to evaluate the predictability of RELAP5/MOD3.2.2 

Beta computer code for the thermal-hydraulic system behavior during a total loss of feed water 

(LOFW) transient. The relevant experiment was conducted in the PMK-2 facility at the KFKI 

Atomic Energy Research Institute in Budapest, Hungary. The test simulated a beyond de

sign-basis accident scenario with unavailability of the hydro-accumulators. For prevention of 

core damage, accident management strategies were applied, including a primary side 

bleed-and-feed procedure with intentional depressurization of the secondary side. After a brief 

description of the facility, the profile of the experiment is presented. Modeling aspects are dis

cussed in the RELAP analysis of the test. Emphasis is placed on the ability of the code to repro

duce the system response to opening of the pressurizer safety valve and feeding of coolant by the 

high pressure injection system. The calculations revealed that the code was capable of predicting 

the parameters with a sufficiently good overall agreement.

iii
CONTENTS

A bstract ................................ 

C ontents ............................... 

List of Tables and Figures ................. 

Acknowledgments ....................... 

Nom enclature ............................ 

1. Introduction ........................... 

2. Modeling of the VVER-440 Reactor ........ 

2.1 The Reference Reactor ............... 

2.2 Characteristics of the PMK-2 Facility ... 

3. Profile of the PMK-2 TLFW Experiment .... 

4. Analysis of the RELAP Calculations ........ 

4.1 Features of the RELAP5 input ......... 

4.2 The Base Case Calculation ............ 

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

4.4 Run Statistics ...................... 

5. Conclusions ........................... 

References ..............................

S.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 

. . •. . . . . . .. *. . . . . . . . . . .. . °. . . . . . . . . .. . . V 

.. . . . . . . .. . . . . . . .. . .. . . . . . . .. .. . I . . . . v i 

. . . .. . .. . . . . . . . . .. .. . .. . . . . .. . . . . . . v ii 

... ..... . . .... . . ... . ... . . . . ... .. . . . ix .

.. 12 

.. 12 

.. 13 

.. 18 

.. 19 

.. 20 

.. 21

V
List of Tables 

Table 1. Main parameters of the PMK-2 facility and the VVER-440 reactor .............. 2 

Table 2. Initial conditions ...................................................... 6 

Table 3. Occurrences in the test ................................................. 6 

List of Figures 

Fig. 1. Axonometric view of the PMK-2 facility .................................... 4 

Fig. 2. Cross section of the core ................................................. 5 

Fig. 3. The SG model of the PMK-2 facility ....................................... 5 

Fig. 4. Initial temperatures and the heating power ................................... 7 

Fig. 5. Initial pressures and differential pressures ................................... 8 

Fig. 6. Initial collapsed levels ................................................... 9 

Fig. 7. Primary pressure and the collapsed level in the PRZ .......................... 10 

Fig. 8. Effects of the bleed on the secondary side pressure ........................... 10 

Fig. 9. Nodalization scheme of the PMK-2 ....................................... 11 

Fig. 10. Pressures in the SG secondary side ....................................... 14 

Fig. 11. Pressures in the upper plenum ........................................... 14 

Fig. 12. Collapsed level in the PRZ ............................................. 15 

Fig. 13. Mass flow rate at PRZ safety valve ....................................... 15 

Fig. 14. Integrated mass flow at the PRZ safety valve ............................... 16 

Fig. 15. The total steam mass dumped from the SG ................................. 16 

Fig. 16. Rod surface temperatures .............................................. 17 

Fig. 17. Fluid temperature at the core inlet ........................................ 17 

Fig. 18. Prim ary pressures .................................................... 18 

Fig. 19. Collapsed levels in the PRZ ............................................ 18 

Fig. 20. Mass flow rate in PRZ safety valve ....................................... 19

vi
ACKNOWLEDGMENTS

The author is grateful to the personnel of the PMK-2 team at the KFKI Atomic Energy Research 

Institute for conducting the tests and providing the experimental database to us. Special thanks 

are due to Dr. Gy6rgy ltzs6l and Dr. Ldszl6 Perneczky for their valuable suggestions and instruc

tions.

vii
NOMENCLATURE 

BRU-A Steam generator atmospheric relief valve 

CL Cold Leg 

DC Downcomer 

ECC Emergency Core Cooling 

HA Hydro Accumulator 

HL Hot Leg 

HF Henry-Fauske model 

HPIS High Pressure Injection System 

IAEA International Atomic Energy Agency 

LOCA Loss Of Coolant Accident 

LP Lower Plenum 

LPIS Low Pressure Injection System 

MV Motor Valve 

NPP Nuclear Power Plant 

PMK Paks Model Circuit (in Hungarian) 

PRZ Pressurizer 

PV Pressure Valve 

PWR Pressurized Water Reactor 

SG Steam Generator 

SIT Safety Injection Tank (synonym for hydro-accumulator) 

SPE: Standard Problem Exercise 

TLFW: Total Loss of Feed Water 

UP: Upper Plenum 

VVER: Water cooled Water moderated Energetic Reactor (in Russian) 

dt: Time step

ix
1. INTRODUCTION

There are a few geometrical and structural differences between the Russian-design VVER-440 

reactors and standard Western types pressurized water reactors widely used in the world. Due to 

the specific features of these nuclear power plants, the computer codes originally developed for 

typical PWRs have to be validated for VVER transients. In order to set up experimental data

bases, the PMK-2 facility was constructed in the early 1980s and upgraded during the past few 

years. 

Performance of the Hungarian Paks NPP has been excellent since the beginning of their com

mercial operation and this PWRs are among the highest in the ranking of load factors. Looking at 

the status of system code verification process of VVER-440/213 in the thermal hydraulic area, 

there is a great need for enlargement of existing data base with new experiments. Without the 

proper benchmarking against experimental data, a computer code cannot be used with any de

gree of confidence. With this purpose in mind, the International Atomic Energy Agency orga

nized test series, such as the Standard Problem Exercises (SPE- 1 to 4) in the PMK-2 facility. 

Results of these large projects demonstrated the applicability of various computer codes for 

small break LOCA and primary-to-secondary leakage transients (IAEA 1987, 1988, 1991 and 

1996). 

The PMK-2 total loss of feed water test (TLFW) simulated a beyond design basis accident. This 

initiating event was selected, because loss of feed water accidents give a relatively high contri

bution to core melt frequency. Furthermore, this test served as an experimental basis to promote 

the implementation of accident management measures, e.g. primary and secondary side bleed 

and primary side feed, at the Paks NPP. More specifically, the principal aim of the present test 

was to reveal wether the reactor level can decrease to the elevation of the of the hot-leg and con

sequently, interrupting the natural circulation. On the other hand, the database is used in the 

evaluation of the effects of the HPIS injection on the pressure decrease in the primary circuit. 

The project was supported by the National Committee of Technological Development in 1995. 

The goal of the current study is to contribute to validation process with an analysis using 

RELAP5/Mod 3.2.2 Beta version of the code against a specific type of accidental situation, 

when the feed water is lost and the operator intervenes. In particular, the code has been chal1
lenged by a few characteristic phenomena in the test, such as the simulation of the blowdown of 

the pressurizer and the heat transfer in the horizontal tube bundle in the steam generator during 

the evaporation of the secondary side inventory. 

A thorough description on the findings of the TLFW experiment was worked out by Ezs6l et al. 

(1996). The main parameters of the PMK-2 facility and the reference reactor are compared in 

Table 1. 

Parameter PMK-2 VVER-440 

Volumetric scaling ratio 1:2070 

Vertical scaling factor 1:1 1 

Number of primary loops 1 6 

Max. heating power 0.7 MW 1375 MW 

Number of fuel rods 19 44000 

Diameter of heater rods 9.1 mm 9.1 mm 

Heated length of the core 2.50 m 2.42 m 

Axial power distribution uniform cosine 

Axial peaking factor 1 

Number of SG 1 6 

Number of SG tubes 82 5500 

Average SG tube length 3.715 m 9.26 m 

Number of SG tube rows 82 76 

Pitch of the SG tubes 18-32 mm 24 mm 

Max. operating pressure 12.3 MPa 12.3 MPa 

Nominal prim. temperature 573 K 573 K 

Nominal. sec. pressure 4.65 MPa 4.65 MPa 

Nominal sec. temperature 533 K 533 K 

Accumulator pressure 5.5 MPa 5.5 MPa 

LPIS pressure 0.7 MPa 0.7 MPa 

HPIS pressure 12.0 MPa 12.0 MPa

Table 1: Main parameters of the PMK-2 facility and the VVER-440 reactor

2
2. MODELING OF THE WER-440 REACTOR

2.1 The Reference Reactor 

The VVER-440/213 PWRs of Russian design are located in mainly Eastern and Central Europe, 

among others, at Loviisa, in Finland and at Paks, in Hungary. This type of reactor has a number 

of unique features, such as horizontal steam generators, combined safety injection, accumulator 

injection to upper plenum and to the downcomer with higher set-point pressure than the second

ary pressure, loop-seal in both hot-leg and cold-leg, and fuel bundles in hexagonal arrangement. 

Consequently, compared with Western type PWRs, the system's response to the disturbances 

can be different. The following phenomena have also been reproduced in most of the PMK-2 ex

periments: 

"* Heat transfer in the horizontal tube bundle of the steam generator: because of the relatively 

large secondary side inventory, the temperature and density distribution is different and inter

nal circulation may develop. 

"* Loop-seal effect: phase separation in the horizontal components is different, with the possible 

formation of loop-seals in the hot-legs and cold-legs. Loop-seals in the hot-leg may reduce or 

completely stop primary coolant flow to the SG during natural circulation or in boiler-con

denser mode. 

"* Different mixing and condensation effects, especially during emergency core cooling (ECC) 

injection to the upper plenum. 

2.2 Characteristics of the PMK-2 Facility 

The PMK-2 is a full-pressure, integral-type model of the four VVER-440 type pressurized water 

reactors used in the Paks NPP in Hungary. The facility (Figure 1) is located in the KFKI Atomic 

Energy Research Institute in Budapest in Hungary. The aspect ratio for power and volumes is 

1:2070. Relative elevations are preserved the same as in the reference reactor, (except for the 

pressurizer and lower plenum), in order to maintain the driving head for natural circulation. The 

primary and secondary sides of the PMK-2 correspond to six loops in the NPP. The facility is 

equipped with similar safety systems, e.g. two hydro-accumulators (safety injection tanks,

3
Pressurizer

Hydro-accumulators

Figure 1: Axonometric view of the PMK-2 facility

4

Steam 

generator

leg
SITs), high and low pressure injection 

systems (HPIS and LPIS). The SIT-I in

jects water to the top of the downcomer, 

while the SIT-2 is connected to the up

per plenum. The LPIS is attached to the 

downcomer head. The setpoint values of 

these components are identical to those 

of the reference plant. The reactor vessel 

is simulated with a U-tube construction, 

consisting of the core model and an ex

ternal downcomer. The core itself com

prises 19 directly heated fuel rod 

simulators embedded in a hexagonal ce

ramic shroud (Figure 2). The diameter 

and pitch of the heater rods are 9.1 mm 

and 12.2 mm, respectively and the ac

tive length is 2.5 m. The nominal power 

of 664 kW is distributed uniformly 

along the height of the rods. 

The steam generator model (Figure 3) is 

a full height, vertically divided section 

of the horizontal VVER-440 steam gen

erator, with 82 tubes bent in serpentine 

shape. The ratio between the volumes of 

steam and water is kept in the secondary 

side. The PV23 valve acts as a steam re

lief valve, which opens and closes at the 

setpoint secondary pressures or by oper

ator intervention. Steam output from the 

SG can be controlled by the PV22 valve, 

which is used before the transient initia

tion.

Figure 2: Cross section of the core

I ~0508 x 16 

SdFee water 

Figur m 3 -the M-- fi 

Figure 3: The SG model of the PMK-2 facility

5
3. PROFILE OF THE PMK-2 TLFW EXPERIMENT

The total loss of feed water Parameter Code Value 

transient started from the nom

inal operating parameters of Primary pressure PR21 12.23 ± 0.05 MPa 

the reference plant at time=0, Loop flow FL53 5.44 ± 0.06 kg/s 

when the feed water is lost. The Core inlet temp. TE63 530.3 ± 1 K 

initial conditions and the main Core power PWO1 649.6 ± 3 kW 

occurrences of the PMK-2 Level in PRZ LE71 9.308 ± 0.02 m 

TLFW test are listed in Tables 

SG sec. pressure PR81 4.22 ± 0.02 MPa 2 and 3, respectively. As a con

Level in SG secondary LE81 9.03 ± 0.05 m servative assumption, the hay

dro-accumulators were Feed water flow FL21 0.37 + 0.02 kg/s 

considered to be inoperable. Feed water temp. TE21 500.6 ± I K 

The core input power used in Table 2: Initial conditions 

this test followed the decay 

heat curve after the scram. The test included the modeling of the pump trip and coast-down. In 

order to promote the implementation of the accident management measures in the Paks NPP, 

primary bleed-and-feed (using the pressurizer safety valve and the HPIS, respectively) and bleed 

of the sec. side by opening the SG relief valve (BRU-A) were modeled. The primary feed was as

sured by only 1 operable HPIS Occurrence Time Actuated by 

and it was triggered from the 

SG isolation 0Os 

signal of"PRZ level 1 m below 

nominal". The primary and Loss of feed water 0 s 

secondary bleed valves were Scram 6 s PR21=12.0 MPa 

modeled with orifices of 1 mn Sec. bleed 127 s PR81=4.96 MPa 

and 4 mm of diameter, respec- Primary feed 513 s LE71=8.33 m 

tively. These valves remained Pump trip 1523s PR21-7.6MPa 

open after their actuation. Lo

Primary bleed 1523 s PR21=7.6 MPa cation of the transducers and 

the measured initial values are Pump stopped 1673 s Pump trip + 150 s 

shown on the simplified flow Test terminated 7886 s 

diagrams in Figures 4, 5 and 6. Table 3: Occurrences in the test

6
INITIAL TEMPERATURES AND POWER 

IN PMK-2 LOFW EXPERIMENT

Figure 4: Initial temperatures and the heating power

7

Height [m]

-11 

-10

-- 1
INITIAL PRESSURES AND DIFF. PRESSURES 

IN PMK-2 LOFW EXPERIMENT

PV71

-11 

-10 

-7 

-6 

-5 

-4 

-3 

-2 

-1 

-0 

-1

Figure 5: Initial pressures and differential pressures 

8

Height [m]

4 220 MPa

9.343

'.04 3 Wý

MV12
INITIAL LEVELS AND MASS FLOWRATES 

IN PMK-2 LOFW EXPERIMENT

Figure 6: Initial collapsed levels

9

Height [m] 

.10

1644, ý
The transient can be 

characterized by the 

variation of pressure in 

the upper plenum. As it is 

shown in Figure 7, PR21 

decreases rapidly after 

the scram and this trend 

slows down until the sec

ondary bleed (opening of 

PV23, t=-127 s). Due to 

the heat sink and the

PMK TLFW: Primary pressure and PRZ collapsed level

13 

12 

11 

10 

L9 

wu8 

ix 6 

5 

4 

3 

2

0 1000 2000 3000

10.5 

10 

SPRZ level (LE71) 

9.5 -J 

W 

9 

0. 

Primary pressure (PR21) 

8.5 

8 

4000 5000 6000 7000 8000 

TIME [s]

lower secondary pres- Figure 7: Primary pressure and the collapsed level in the PRZ 

sure, further decrease 

was observed in PR2 1. The first characteristic point can be seen at the actuation of high pressure 

injection system (primary feed, t=513 s). Initially, the PRZ level decreases due to the lower cool

ant temperature. After the start-up of the HPIS, the PRZ level starts to rise again. Reaching the 

nominal level in the PRZ, the safety valve PV71 opens and the primary bleed begins at 1523 s. 

The fast depressurization continues until approximately 2 100 s, when two-phase flow starts to 

be discharged. The maximal level in the PRZ is approached at 4000 s, followed by a nearly stabi

lized state. In this phase, the inventory loss through the safety valve is roughly compensated by 

mass of the primary feed. PMK TLFW: Pressure in SG secondary

6

The pressure in the steam 

generator secondary side 

is depicted in Figure 8. 

The test results show that 

the SG is effective for the 

whole duration of the 

transient. The core re

mains covered, since 

changing of the level in 

the primary side takes 

place only in the PRZ.

5 

0a 

ci) 

U) 

1 

0

0 1000 2000 3000 4000 5000 

TIME [s]

6000 7000 8000

Figure 8: Effects of the bleed on the secondary side pressure

10
_11 Height [m] 

.10

-9 

-8 

ý9 -7 

z 0 

-6 

0 ýo 

.5 

Ct 

C)2 

CD 4 

0 

-3 

-2 

-0

--1
4. ANALYSIS OF THE RELAP CALCULATIONS

4.1 Features of the RELAP5 input 

The basic structure of the nodalization scheme (Figure 9) was adopted from the former IAEA 

Standard Problem Exercise calculations, with a few modifications according to the needs of the 

TLFW test. (E.g. the steam line was removed and the pressurizer surge line was connected to the 

bottom of the hot leg). The model includes 96 volumes and 105 junctions. 

The main features of the input can be summarized as follows: most of the hydrodynamic vol

umes were modeled with branches. Crossflow junctions are applied at connecting the following 

parts: downcomer to the lower plenum, upper plenum U-tubes, upper plenum to the hot leg, SIT 

connections, cold leg connection to the downcomer top, valve connections to the steam dome, 

the pressurizer safety valve and finally the steam generator hot and cold collector connections to 

the heat exchange tubes. 

The core is axially subdivided into 10 parts. Rod bundle interphase friction model is used in the 

volumes of the heated section. Counter-current flow limitation (CCFL) model is turned on only 

in the core and the downcomer. Choking option is applied in the vicinity of the pressurizer safety 

valve (PV71) and the steam generator relief valve (PV23), in order to simulate the primary and 

secondary bleed processes. 

In fact, the pressurizer safety valve behaves like a break in the primary side during the bleed 

phase. Consequently, it is necessary to assure the adequate value for the valve discharge coeffi

cient to match the correct amount of released primary coolant. The same requirement is valid for 

the SG depressurization. 

The steam generator model includes basically three primary side horizontal layers for the heat 

exchanger tubes and three vertical volumes in the secondary side. The steam dome consists of a 

separator component near the initial secondary water level. The original PMK-2 steam generator 

does not contain any steam separator. However, this arrangement helped to intensify the internal 

circulation and resulted in a better heat transfer. Based on the experiences of earlier calculations,

12
void fraction limitation quantities of 0.35 and 0.55 yielded to the best results for the vapor exit 

junction and the liquid fall-back junction, respectively. 

The current RELAP5 model allows the code user to have full control over appropriate simula

tion of the pump. The steady-state loop flow rate is achieved by regulation of the pump speed. In 

the transient, the by-pass line is closed by the MV1 1 valve and the pump is running at constant 

speed until the pump trip. When the pump is tripped, the PV1 1 valve starts to close while the 

MV1 1 valve is gradually opening in a prescribed manner. Simulation of the pump coast-down is 

terminated by closing of the MV12 valve. The nominal length of this process is 150 s. 

Heat structures are used in the heater rod simulators, in the PRZ heaters, in the steam generator 

tubes and in all the pipe walls along the facility, describing the heat exchange between the 

environment. For heat losses, convective boundary conditions were considered, with a constant 

heat transfer coefficient of 5 W/mn2K. 

The present model of the PMK-2 includes the safety injection tanks and the pressurizer spray 

line, however, these components had no specific role during this particular transient. 

The initial data set and set-point values were derived from the nominal operating parameters of 

the VVER-440 plant. The strategies used to achieve steady-state with the code were similar to 

those of applied in earlier analyses. Auxiliary components were connected to primary and sec

ondary sides to stabilize the initial parameters nearly at the same values as it was observed in the 

test. 

4.2 The Base Case Calculation 

The base case calculation focused on the system behavior during the steam dump at the 

pressurizer and SG relief valve. For this purpose, homogeneous (h=2) choking flow (c=O) option 

was selected at the bleed valve junctions, with default discharge coefficients (1.0, 1.0, 1.0). 

The search run took only a few seconds with using the "steady-state" option of the code. The re

quested time step was 0.1 s for the whole duration of the transient calculation. The transient

13
started by the isolation of PMK LOFW: Pressure in SG secondary (h=9.300 m) 

the SG secondary side. Relap 

4.5 .PR31 

Following the scram at 6 

4 . 

s of transient time, the 3. 

depressurization of the 3 

primary system started r 2.5 

rapidly. Figure 10 illus- OW 2 

0- 1.5 

trates the time variation 1 

of measured and calcu- 0.5 -- ----- 

lated values of the pres- 0 

0 1000 2000 3000 4000 5000 6000 7000 8000 

sure in the SG secondary TIME [s] 

side. The set-point of the Figure 10: Pressures in the SG secondary side 

SG relief valve was 

reached at 127 s. At this time the valve PV23 opened and remained in open position for the rest 

of the transient. Bleeding of the secondary side resulted in a fast decreasing of pressure. 

While RELAP5 predicted the SG pressure at a good accuracy, the pressure in the upper plenum 

was well simulated only in the first part, until about 4000 s (Figure 11). The first characteristic 

point can be observed at 513 s, when the HPIS (feed) was initiated. Beginning of the primary 

side bleed is well matched at 1523 s. However, the slight increase of the pressure after 4000 s can 

be explained by the wa

ter level reaching the ele- 14 PMK LOFW: Pressure in upper plenum (h=3.754 m) 

Relap - I 

vation of PRZ safety P2 ........ 12 

valve. 

W 10 

By examination of Fig- " 8 

ure 12, the following go 6 

events have particular in- 0 4" 

terest. There was a con 2•

tinuous dropping of the 

collapsed level in the 0 1000 2000 3000 4000 5000 6000 7000 8000 

TIME [s] 

pressurizer until 513 s. 

At this time the primary Figure 11: Pressures in the upper plenum

14
side feed was actuated. PMK LOFW: Collapsed level in pressurizer 

Injection of cold water 10.4 

10.2 

by HPIS started to fill up 10 

the pressurizer. The col- . 9.8 

lapsed level reached the > 9.6 .. 9.4 

nominal value at 1523 s. tw 9.2 

At this moment the pn- 9 

mary side bleed process 8 

8.6 / 

was initiated by opening 8.4 6. Relap 

the pressurizer safety 8.2 

0 1000 2000 3000 4000 5000 6000 7000 8000 

valve PV7 1. The fill-up TIME [s] 

process was terminated Figure 12: Collapsed level in the PRZ 

at approximately 2100 s, 

when the collapsed level reached the elevation of valve PV7 1. It can be concluded that RELAP5 

predicted the variation of the level in the PRZ fairly well. In the period, while there is one phase 

flow through the safety valve, is well simulated. The nearly-stabilized level is only about 0.1 m 

higher in the calculation than in the experiment. (Rather complex thermal-hydraulic process 

took place in the vicinity of the PRZ safety valve during the primary bleed. Arrangement of the 

valve and the ducts of the DP and level measurements should be considered in this respect). 

The mass flow rate fell 

PMK LOFW: Mass flow at PRZ safety valve back to zero twice soon 0.016 

after triggering the bleed. 0.014 .. 

As Figure 13 shows, 0.012 • 

RELAP could not repro- 4 0.01 ... .. t.r 

duce the short o 0.008 0 

stagnations of the steam 0.006 

(0.0 flow between 1500 and < 0.004 . 

2000 s. However, the 0.0023 .i 

general trend is 0 R--e--.... l 

-0.002 1 

sufficiently well calcu- 0 1000 2000 3000 4000 5000 6000 7000 8000 

lated with using default TIME [s] 

values for the valve. Figure 13: Mass flow rate at PRZ safety valve

15
The total inventory loss 

through the PRZ safety 

valve is depicted in Fig

ure 14. It is visible that 

this parameter is under

estimated. The final cal

culated value is approxi

mately 60 kg, instead of 

the measured amount of 

70 kg. (A minor initial 

discharge of the safety 

valve was assumed be

fore the valve opening in 

the calculation).

PMK LOFW: Integrated mass flow at PRZ safety valve

U)

80 

70 

60 

50 

40 

30 

20 

10 

0

0 1000 2000 3000 4000 5000 

TIME [s]

Bleeding of the secondary side can be seen in Figure 15. Opening of the SG safety valve oc

curred at 127 s of simulation time. The BRU-A valve was intentionally left open for the rest of 

the transient. This resulted in a continuous dumping of the steam through an orifice with a diam

eter of 4 mm. This value was selected considering the scaling factor of the PMK-2 and the fact 

that one steam generator simulates 6 ones in the reference reactor.

The bleeding process is 

very well matched by the 

code calculation. Using 

the default values for the 

orifice assured an excel

lent reproduction of the 

total mass in the first half 

of the transient. The dis

crepancy increased just a 

few percent near the end 

of the simulation.

PMK LOFW: Integrated mass flow at BRU-A

0) _J 

Ci) Ci)

180 

160 

140 

120 

100 

80 

60 

40 

20 

0

0 1000 2000 3000 4000 5000 6000 7000 8000 

TIME [s] 

Figure 15: The total steam mass dumped from the SG

16

6000 7000 8000

Figure 14: Integrated mass flow at the PRZ safety valve
The history of the tem- PMK LOFW: Rod surface temperature (h=3.464 m) 

perature of the heater 600 .e..p 

Rel.3 

rods at the elevation of 580 T , 

560 3.464 m, near the exit of ua 540.. ... .. .... .. 

the core, is shown in Fig- cr 

5520 

ure 16. Due to the scram, o ir 500 

the temperature started 02 480 

to drop very rapidly. A • 460 

stagnation can be ob- 440 

served at nearly 540 K 420 

0 1000 2000 3000 4000 5000 6000 7000 8000 

for a short duration, fol- TIME [s] 

lowed by a decreasing. Thisd parametrerasialso Figure 16: Rod surface temperatures This parameter was also 

simulated with a good degree of accuracy until approx. 500 s. However, it started to be slightly 

underestimated before a minor heat-excursion at 1800 s. The timing and the magnitude of the 

minor were well predicted by RELAP. 

The primary fluid temperature has similar characteristic points as the secondary pressure curve. 

Figure 17 shows the coolant temperature at the inlet of the core. Reaching the maximum at the 

initiation of the SG bleed, the process is driven by the energy release in the secondary. Decrease 

is accelerated after open

PMK LOFW: Fluid temperature at core inlet (h=0.1 90 m) ing of the PRZ valve. 54 540 r 

Relap This is the break point of : YI.. 

520 .. 

the curve at - 1800 s. 

,, 500 

The calculation has con- 480 

firmed that the reactor w 460 

level remained stable. 

Voiding did not take 420 

420 

place in the upper ple 400 ± L t_ 

num and consequently, 0 1000 2000 3000 4000 5000 6000 7000 8000 

the core was constantly TIME [s] 

covered. Figure 17: Fluid temperature at the core inlet

17
4.3 Sensitivity Study

It was found in the PMK LOFW: Pressure in upper plenum (h=3.754 m) 

base-case calculations 14 HF model off 

HF on. c=tD.85. 

that the process was sen- 12 HF on, c=1.20 

Experiment sitive for both the pri- 10 

mary and secondary • • uW 8 

bleed. Using default dis- c 

cO 6 

charge coefficients un- 'wO 

derestimated the PRZ 4 

bleed flow. Increasing 2 

the value resulted in o 

0 1000 2000 3000 4000 5000 6000 7000 8000 

wrongly simulated pri- TIME [s] 

mary pressure. There is 

an option in the currently Figure 18: Primary pressures 

applied version of the code to use the Henry-Fauske critical flow model. In order to evaluate the 

effects of the new model, sensitivity calculations were performed with varied parameters. 

Values of 0.85 and 1.2 were used for the discharge coefficient, while the thermal non-equilib

rium parameter was left constant (0.14, as recommended in the code manual). Figures 18, 19 and 

20 show the results of comparison to the previous calculation, where the H-F model was not ap

plied. It can be seen in 

Figure 18 that the small- PMK LOFW: Collapsed level in pressurizer 

10.4. ... 

est discrepancy is 10.2 

achieved without the HF 10 

model, until approx&. T 9.8 

w 9.6 

4000 s, but c=0.85 shows w -j 9.4 

better result after that. A OW 9.2 

0) 

similar conclusion can 25 9 

O 8.8 

be drawn for the PRZ 0 HF model off 

8.6 HF on. -=O,8 5 

level but all the 3 param- 8.4 HF. on, c=1.20 

Experiment 

eters resulted in identical 8.2 

0 1000 2000 3000 4000 5000 6000 7000 8000 

calculated levels after TIME [s] 

4600 s. Figure 19: Collapsed levels in the PRZ

18
It is interesting to note PMK LOFW: Mass flow at PRZ safety valve 

that the highest dis- 0.016 

charge coefficient gives 0.014 

0.012 the best correlation in the • 0.01 

first half of the transient A 

S0.008 0 but the flow becomes I 0.006 

more underestimated un- u 0.004 

til the end of the calcula- 0Hm l 0.002 .. . .. .. HF model off 

tion (Figure 20). 0 HF on, c=-0.20 S.. .. .......... HF on, c=1.20 

Looking at the plots of -0.002 -xemnt 

0 1000 2000 3000 4000 5000 6000 7000 8000 the sensitivity study, it TIME [s] 

cannot be stated that the 

results of Henry-Fauske Figure 20: Mass flow rate in PRZ safety valve 

model would be in clearly better agreement with the experimental data. The ordinary critical 

flow model shows closer matching for some of the parameters. 

4.4 Run Statistics 

The calculations were performed on a Silicon Graphics Octane workstation with 175 MHz 

R10000 processor and 256 MB central memory. 

With requested time step of 0.1 s, the 8000 s long transient consumed 1371.07 s CPU time. It 

means the CPU time / transient time ratio was 1371.07 s / 8000 s = 0.17138375. 

The mass error ratio (emass / tmass) was reasonably low: -0.354311 kg / 361.207 kg = 

-9.809086E-04

19
5. CONCLUSIONS

The analyses with RELAP5/MOD3.2 computer code showed that code is capable of predicting 

the key events in the experiment. It can be stated generally, that main trends have been repeated 

in the calculation and the qualitative agreement is good. 

Matching of the calculated and measured conditions was excellent in the pressurizer. From ther

mal hydraulic aspects, very complex processes took place in this relatively small volume: feed

ing with cold ECC water and bleeding by the safety valve, furthermore, phase separation near 

the elevation of the valve. The inventory loss through the bleed valve challenged the code be

cause it behaved as a small break in the top of PRZ. RELAP5 calculated this process with a good 

accuracy. 

In spite of the fact that the code was originally developed for simulation of reactors with vertical 

steam generators, the secondary side model has been proven to be adequate. The horizontal heat 

exchanger tubes were lumped into 3 layers. Even with this simplified arrangement, the results 

produced by the code were in very good agreement with the experimental data. The bleed phase 

was better simulated in the secondary side. 

The sensitivity study revealed that the application of the Henry-Fauske critical flow model did 

not necessarily improved the prediction of the two-phase flow at the orifices. 

Dependence of the results on the requested time step value was not experienced. The results ob

tained by this simulation has confirmed that the built-in automatic time step selection method of 

RELAP5/MOD3.2.2beta could be used with confidence.

20
REFERENCES

J. Bdinditi: "Assessment Study on the PMK-2 Total Loss of Feedwater Experiment Using 

RELAP5 Code", Research Report, Lappeenranta University of Technology, Finland, Vol. A-5 1, 

ISBN 951-764-518-X, ISSN 0785-823X, Lappeenranta, 2001. 

J. Bdnkti, Gy. Jtzs~l and J. Kouhia: "RELAP5/MOD3.2 Assessment Studies Based on the 

PACTEL and PMK-2 Loss of Feed Water Tests", 3rd ASME / JSME Joint Fluids Engineering 

Conference / Symposium on System Transient Analysis Codes, San Francisco, CA, USA, July 

18-22, 1999. ISBN 0-7918-1961-2, Paper no.: FEDSM99-7009. (Published on CD-ROM) 

Binkti, J., 1995, "Assessment of RELAP5/MOD3.1 Code for the IAEA SPE-4 Experiment", 

Proceedings of the International Symposium on Validation of System Transient Analysis Codes, 

ASME & JSME Fluids Engineering Conference, Hilton Head, SC, USA. FED-Vol. 223, pp. 

17-24 

Bdinditi, J., 1995, "Simulation of a Beyond Design-Basis-Accident with RELAP5/MOD3.1", 

Proceedings of the 7th International Topical Meeting on Nuclear Thermal Hydraulics 

(NURETH-7), Saratoga Springs, NY, USA, pp. 1116-1126 

Bdndti, J. and tzsSl, Gy., 1997, "Simulation of Some Emergency Operating Procedures for 

VVER-440 Reactors", Proceedings of the ASME International Mechanical Engineering Con

gress and Exposition, (IMECE-97), Dallas, TX, USA, NE-Vol. 21, pp. 57-64. 

B~inti, J. and tzs5l, Gy., 1997, "Modeling of Accident Management Procedures for 

VVER-440-Type Reactors", Proceedings the 8th International Topical Meeting on Nuclear 

Thermal Hydraulics (NURETH-8), Kyoto, Japan, pp. 1134-1141 

lzs6l, Gy., Perneczky, L. and Szabados, L., 1996, "Bleed and Feed Studies to Assess the Devel

opment of Emergency Operating Procedures for VVER-Type Reactors", Proceedings of the 4th 

International Conference on Nuclear Engineering (ICONE-4), New Orleans, USA, Vol. 3, pp. 

511-516.

21
Hyvirinen, J., 1996, "On the Fundamentals of Nuclear Reactor Safety Assessment, Inherent 

Threats and Their Implications", Doctor of Technology Thesis, Lappeenranta, Finland, 

STUK-A135 

IAEA, 1987, "Simulation of a Loss of Coolant Accident", (SPE-1) IAEA TECDOC-425, Vi

enna, Austria 

IAEA, 1988, "Simulation of a Loss of Coolant Accident with Hydroaccumulator Injection", 

(SPE-2) IAEA TECDOC-477, Vienna, Austria 

IAEA, 1991, "Simulation of a Loss of Coolant Accident with Rupture in the Steam Generator 

Hot Collector", (SPE-3) IAEA TECDOC-586, Vienna, Austria 

IAEA,1995, "Simulation of a Loss of Coolant Accident without High Pressure Injection but with 

Secondary Side Bleed and Feed. Results of the 4th Standard Problem Exercise", 

IAEA-TECDOC-848, Vienna, Austria 

Tuomisto, H., et al. (editors), 1995, Proceedings of the Third International Seminar on Horizon

tal Steam Generators, Lappeenranta University of Technology, Finland, Research Papers, Vol. 

43. 

Tuunanen, J., Nakada, K., 1993, "Analysis of PACTEL Loss of Secondary Side Feed Water with 

RELAP5/MOD3 Code", ASME Conference on "Transient Phenomena in Nuclear Reactor Sys

tems", Atlanta, Georgia, USA, HTD-Vol. 245, NE-Vol. 11, pp. 81-91.

22
Time (sec) 

stl O.spb

10 20 30 40 50

Time (sec) 

- stl10.S:p~b]

14 

12 

a. 6 

E 

ci 

•.4 

2 

0

0 60
Time (sec) 

stlO.spb

0 10 20 30 40 50

Time (sec) 

-- stlO.spb

16000 

14000

4)

12000 

10000 

8000

E 

"a) 

E 0 

LL

6000 

4000

60
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add Vol., Supp., Rev., 

NRCM 1102, and Addendum Numbers, if any.) 3201, 3202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) NUREG/IA-0200 

2. TITLE AND SUBTITLE 

Assessment Study on the PMK-2 Total Loss of Feedwater Experiment Using 

RELAP5 Code 3. DATE REPORT PUBLISHED 

MONTH YEAR 

March 2001 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

J. Banati Technical 

7. PERIOD COVERED (Inclusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, 

provide name and mailing address.) 

Lappeenranta University of Technology 

Department of Energy Technology 

P.O. Box 20 

FIN-53851 Lappeenranta, Finland 

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type 'Same as above"; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

and mailing address.) 

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less) 

The main objective of the present report is to evaluate the predictability of RELAP5/MOD3.2.2 Beta computer code for the 

thermal-hydraulic system behavior during a total loss of feed water (LOFW) Transient. The relevant experiment was conducted 

in the PMK-2 facility at the KFKI Atomic Energy Research Institute in Budapest, Hungary. The test simulated a beyond 

design-basis accident scenario with unavailability of the hydro-accumulators. For prevention of core damage, accident 

management strategies were applied, including a primary side bleed-and-feed procedure with intentional depressurization of the 

secondary side. After a brief description of the facility, the profile of the experiment is presented. Modeling aspects are 

discussed in the RELAP analysis of the test. Emphasis is placed on the ability of the code to reproduce the system response to 

opening of the pressurizer safety valve and feeding of coolant by the high pressure injection system. The calculations revealed 

that the code was capable of predicting the parameters with a sufficiently good overall agreement.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT 

PMK unlimited 

RELAP5 14. SECURITY CLASSIFICATION 

(This Page) 

unclassified 

(This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

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NUREG/IA-0200 ASSESSMENT STUDY ON THE PMK-2 TOTAL LOSS OF FEEDWATER EXPERIMENT 

USING RELAP5 CODE

MARCH 2001

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

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