NUREG/IA-009 1

ECN-89-9 1

International

Agreement- Report

Assessment of RELAP5/MOD2

Against a Natural Circulation

Experiment in Nuclear Power

Plant Borssele

Prepared by

L. Winters

Ministry of Social Affairs

Nuclear Inspectorate (KFD)

Netherlands Energy Research Foundation ECN

P.O. Box 1

1755 ZG Petten

The Netherlands

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

July 1993

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report was prepared under an international cooperative

agreement for the exchange of technical information. Neither

the United States Government nor any agency thereof, or any of

their employees, makes any warranty, expressed or implied, or

assumes any legal liability or responsibility for any third party's

use, or the results of such use, of any information, apparatus product or process disclosed in this report, or represents that its use

by such third party Would not infringe privately owned rights.

Available from

Superintendent of Documents

U.S. Government Printing Office

P.O. Box 37082

Washington, D.C. 20013-7082

and

National Technical Information Service

Springfield, VA 22161
NUREG/IA-0091

ECN-89-9 1

International

Agreement Report

Assessment of RELAP5/MOD2

Against a Natural Circulation

Experiment in Nuclear Power

Plant Borssele

Prepared by

L Winters

Ministry of Social Affairs

Nuclear Inspectorate (KFD)

Netherlands Energy Research Foundation ECN

P.O. Box 1

1755 ZG Petten

The Netherlands

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

July 1993

Prepared as part of

Tle Agreement on Research Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report is based on work performed under the sponsorship of the

Netherlands Ministry for Social Affairs and Employment. The information in

this report has been provided to the USNRC under the terms of an information

exchange agreement between the United States and The Netherlands (Technical

Exchange and Cooperation Agreement between the United States Nuclear

Regulatory Commission and the Netherlands Energy Research Foundation ECN).

The Netherlands has consented to the publication of this report as a USNRC

document in order that it may receive the widest possible circulation among

the reactor safety community. Neither the United States Government nor The

Netherlands or any agency thereof, or any of their employees, makes any

warranty, expressed or implied, or assumes any legal liability of

responsibility for any third party's use, of the results of such use, or any

information, apparatus, product or process disclosed in this report or

represents that its use by such third party would not infringe privately owned

rights.
iii

ABSTRACT

As part of the ICAP (International Code Assessment and Applications

Program) agreement between ECN (Netherlands Energy Research Foundation) and USNRC, ECN has performed a number of assessment calculations for the thermohydraulic system analysis code RELAP5/MOD2/36.05.

This document describes the assessment of this computer program versus a natural circulation experiment as conducted at the Borssele

Nuclear Power Plant. The results of this comparison show that the

code RELAP5/MOD2 predicts well the natural circulation behaviour of

Nuclear Power Plant Borssele.

The work has been sponsored by the Dutch Licensing Authority and ECN.

v

CONTENTS Pave

1. INTRODUCTION 1

2. NATURAL CIRCULATION EXPERIMENT AT NPP BORSSELE 2

2.1. Description of the Nuclear Power Plant 2

2.2. Natural circulation experiment 2

2.3. Experimental data 3

3. RELAP5/MOD2 INPUT MODEL 9

3.1. Thermal hydraulic and control system model 9

3.2. Initial conditions 10

3.3. Assumptions for the simulation 11

4. RELAP5/MOD2 ANALYSIS OF THE NATURAL CIRCULATION EXPERIMENT 15

5. COMPARISON OF RELAP5 CALCULATIONS TO EXPERIMENTAL DATA 28

6. CONCLUSIONS 39

7. REFERENCES 40

Appendices:

Appendix A. Description of RELAP5 model for KCB A-1

Appendix B. Analytical model for natural circulation B-i

Appendix C. Run statistics C-1
vi

LIST OF TABLES Page

2.1. KCB system parameters 4

2.2. Sequence of events of natural circulation test 4

2.3. Water temperature of VCS 5

2.4. Recorded signals /6

3.1. Experimental and calculated initial conditions 13

4.1. Sequence of events for simulation of the 23

natural circulation experiment

5.1. Designation of experimental and calculated data 32

B-i. Steady state natural circulation mass flow for different B-4

nodalization. schemes

B-2. Steady state natural circulation mass flow for different B-4

power distributions
vii

LIST OF FIGURES

2.1. N'uclear Power Plant Borssele

2.2. Measures secondary side pressure transient during

natural circulation experiment

2.3. Measured temperature difference between hot and cold

legs during natural circulation experiment

3.1. RELAP5 nodalization of KCB

4.1. RELAP5 mass flow rate inside loop 01 and 02

4.2. RELAP5 steam generator secondary side pressures

41.3. RELAP5 temperature gradients in primary system loops

and core

4.4. RELAP5 hot and cold leg fluid temperatures in loop 01

14.5. RELAP5 hot and cold leg fluid temperatures in loop 02

4.6. RELAP5 energy balance for primary system

4.7. RELAP5 driving forces for natural circulation in steam

Page

7

8

8

14

24

24

25

25

26

26

27

generators, loop seals and core

5.1. Comparison of calculated and measured

5.2. Comparison of calculated and measured

levels in the pressurizer

5.3. Comparison of calculated and measured

levels in steam generator 01

5.4. Comparison of calculated and measured

levels in steam generator 02

5.5. Comparison of calculated and measured

secondary side of steam generator 01

5.6. Comparison of calculated and measured

secondary side of steam generator 02

5.7. Comparison of calculated and measured

differences across loop 01

5.8. Comparison of calculated and measured

differences across loop 02

system pressures

collapsed liquid

collapsed liquid

collapsed liquid

pressures at

pressures at

temperature

temperature

33

33

34

34

35

35

36

36

37

37

5.9. Comparison of calculated and measured fluid

temperatures in the hot leg of loop 01

5.10. Comparison of calculated and measured fluid

temperatures in the cold leg of loop 01
viii

5.11. Comparison of calculated and measured fluid 38

temperatures in the hot leg of loop 02

5.12. Comparison of claculated and measured fluid 38

temperatures in the cold leg of loop 02

A-i Representation of loop 1 A-4

A-2 Representation of the pressurizer A-5

A-3 Representation of the reactor vessel A-6

A-4I Representation of the steam generator A-7

A-5 Representation of ECC system A-8

B-i Schematic of a loop of a PWR B-5

B-2 RELAP5 model of a ioop of a PWR B-5

C-i CPU-time as a function of real time C-2

C-2 Time step size as a function of real time C-2
1

1. INTRODUCTION

Following a reactor scram and trip of the reactor coolant pumps in a

pressurized water nuclear power plant (PWR), decay heat removal may

be established by means of natural circulation. The natural circulation phenomena have been investigated experimentally by many international research organisations. Well known natural circulation experiments have been conducted in the LOFT, Semiscale, PKL and LOBI facilities. In general these facilities are scaled down versions of

commercial plants, however, all have their specific limitations with

respect to natural circulation. The LOFT facility has been equiped

with one steam generator while the elevations of the components are

a-typical. Semiscale, PKL and LOBI have component elevations comnparable with a commercial plant while volumes have been scaled down

drastically. The resulting small pipe diameters may strongly

influence the natural circulation phenomena. For this reason natural

circulation experiments in full size installations are usefull to

confirm the data from the experimental facilities.

Theoretical considerations suggest that the natural circulation flow

in a loop will stagnate under certain conditions. A stagnating flow

in a loop of a nuclear power plant is not desirable due to possible

deboration in the core and consequential reactivity effects after

stagnation is discontinued. Another point of concern is the formation

of a steam bubble in the stagnant loop during depressurization due to

the relative hot water in the loop.

In Borssele Nuclear Power Plant (KCB) a natural circulation experiment has been performed with the objective to investigate the natural

circulation phenomena and the possible occurance of flow stagnation

in the event of an isolated steam generator. EON has performed

assessment calculations based on the available experimental data

using the thermohydraulic system analysis code RELAP5/MOD2/36.05,

ref. Li].

The outline of this document has been chosen in accordance with the

recommendations as given in NUREG-1271, ref. [2]. A description of

the experiment is given in chapter 2 while the RELAP5 input model has

been described in chapter 3. Chapter 4j and 5 respectively present the

results of the analyses and a comparison with the experimental data.
2

2. NATURAL CIRCULATION EXPERIMENT AT NPP BORSSELE

2.1. Description of the Nuclear Power Plant

Nuclear Power Plant Borssele (KGB) is a two-loop, 1365 MWt Pressurized Water Reactor (PWR) designed by Kraftwerk Union (KWU) and owned

by the Provinciale Zeeuwse Energie Maatschappij (PZEM). The plant is

located near the town of Vlissingen in the South West part of the

Netherlands, and operating since 1973. A schematic picture of the

plant including relative elevations is given in figure 2.1. The

inside diameter of the primary loop piping is 0.8 mn. The nuclear core

has 121 fuel elements with a heated length of 2.65 mn. The two steam

generators are U-tube steam generators with a total production at

full power of 7314 kg/s dry steam at a pressure of 5.7 MPa. Each steam

generator has 142314 Incoloy U-tubes with a total surface of 3600 in2 .

The internal diameter of a U-tube is 0.0196 mn and the wall thickness

is equal to 0.0012 m. Table 2.1 gives some important nominal system

parameters of KCB.

2.2. Natural circulation experiment

A description of the natural circulation experiment is given in ref.

[3] and [4I]. Before execution of the natural circulation test the KCB

plant had been in operation almost continuously for a period of 10

months. On February 5, 1983 at 01h. 0Dm., shutdown of the plant was

started in preparation for refueling. At 06h. 00m. the reactor became

subcritical. The natural circulation experiment started at 13h. 145m.

the same day by tripping the two reactor coolant pumps. The test

lasted about 4i hrs during which different actions were taken in order

to establish the right conditions for the test.

The sequence of events of the experiment is given in table 2.2 (the

step numbers are shown also in figure 2.2 and 2.3.). In this table two

times are given viz, the actual time (hrs, in, s) and the relative

time with respect to the start of the experiment (s). In figure 2.2

and 2.3 the measured secondary side pressures and the temperature

differences across the loops are given.
3

During the experiment the Volume Controle System (VCS) was in operation. The temperature of the VCS injection water as a function of

time is given in table 2.3. There is no measurement of the VCS mass

flow available.

2.3. Experimental data

Because there was no opportunity to install additional measuring

instruments only the signals of selected plant process instruments

were recorded. The signals recorded by the process instrumentation

and the location of the measuring devices are given in table 2.4. The

location in the table gives the loop identification (Cold Leg (CL)

and Hot Leg (HL) loops 01 and 02), the location of the measurement

device on the main coolant pipe (0* = top of the pipe) and the distance of the device to the centerline of the reactor vessel (RV) or

bottom of the steam generator pipe plate (main coolant pipe vertical).

All temperature measurements have been performed using Ni-Cr thermocouples, the system pressure by a so-called Bourdon cell and the liquid level in steam generators and pressurizer by means of a pressure

difference measurement. No possibility exists to measure mass flows

in the main loops of the primary system. In the experiment the occurence of natural circulation can be demonstrated by evaluation of the

different temperature measurements. All the signals have been recorded on paper charts only. In order to make a comparison with calculated data easy the measured data of the strip charts was digitized by

hand and stored on magnetic tape.

There is no qualification and uncertainty analysis of measured and

recorded data available. Also some important information to understand discrepancies in the test is missing or not well documented.
4

Table 2.1. KCB system parameters

Nuclear power 1365.6 MW th

System pressure 15.5 MPa

Total core mass flow 10000 kg/s

Vessel outlet temperature 588 K

Number of fuel elements 121

Core average heat flux 615000 W/m'

Steam generator outlet pressure 5.70 MPa

Total steam mass flow 73)4 kg/s

Table 2.2. Sequence of events of natural circulation test

1. Time 13.45.00 (t =0.0 s).

Both reactor coolant pumps are tripped. A few minutes later a

stable natural circulation flow is established in both loops.

2. Time 13.59.00 (t = 8)40 s).

Steam generator 01 is isolated by closing the feedwater and the

main steam valves.

3. Time 14.50.00 (t = 3900 s).

During 360 s the primary system is cooled down with a gradient of

100 K/hr (in accordance with the procedures). The cooling down is

created by a depressurization of about 0.02 MPa/min of the secondary side of steam generator 02.

4i. Time 14.5)4.co (t = 414o1 s).

Injection of auxiliary feedwater into steam generator 02 during

5)40 s.

5. Time 15.15.00 (t = 5400 s).

Isolation of steam generator 01 discontinued by reconnecting the

steam generator to the feedwater and steam lines. Due to the connection line between the main steam lines the pressure on the

secondary side of the steam generators is balanced again.
5

6. Time 15.37.00 (t =6720 s).

Isolation of steam generator 02 by closing the feedwater and main

steam valves.

7. Time 15.56.00 (t = 7860 s). The primary system is cooled down

during 2100 s at a gradient of' 50 K/hr. The cooling down is created by a controlled depressurization of the secondary side of

steam generator 01.

8. Time 16.52.00 (t =11220 s).

Isolation of' steam generator 01 by closing the feedwater and main

steam valves. At this point both steam generators are isolated.

9. Time 17.22.00 (t =13020 s).

Isolation of both steam generators is discontinued by reconnecting each to the main steam and feedwater lines.

10. Time 17.3'4.00 (t = 137140 s).

Isolation of steam generator 01 by closing the feedwater and main

steam valves.

11. Time 17-54.00 (t = 1149140 s).

Discontinuation of isolation of steam generator 01 by reconnecting it to the main steam and feedwater line.

12. Time 18.00.00 (t = 15300 s).

End of the test.

Table 2.3. Water temperature of' the VCS

Time interval Temperature (K)

start -14.142.00 513

14.142.00 -14.50-00 503

114.50.00 -15.15.00 502

15.15.00 -15.37.00 14914

15.37.00 -16.56.00 1490

16.56.00 -17.22.00 1469

17.22.00 - end 473
6

Table 2.4. Recorded signals

Sensor Measurement Location

* YAOIT090

* YA02T090

* YA01T098

* YA02T098

YA01TO11/01

YA01TO11/02

(YA0 iTO 11

YA02TO 11 /01

YA02TO11/02

(YA02T011

YA01T054I/01

YA01T054/02

* (YA01T054

YA02T054/01

YA02TO54/02

* (YA02T0514

* YA01P002

* RA01POOl

* RA02POOl

* YBOlLO51

* YB02L051

* YPO1L1O1

temperature

temperature

temperature

temperature

temperature

temperature

avg. temp.

temperature

temperature

avg. temp.

temperature

temperature

temp. diff.

temperature

temperature

temp. diff.

pressure

pressure

pressure

liquid level

liquid level

liquid level

CLOl, -3D', 10.'470 m. to ERV

CL02, -30', 101170 m to ~RV

HLO1, 30', 8.4214 m to RV

HL02, 30*, 8.4214 m to ~RV

CLOl, NA 3.0140 m to SG pipe plate

HLO1, - 75', 8.424 m to ý RV

avg. of YA01TO11/01 and YAO1TO11/02)

CLO2, NA 3.0140 m to SG pipe plate

HLO2, - 75', 8.424 m to 0 RV

avg. of YA02TO11/01 and YAO2TO11/02)

CLO1, - 60', 10.970 m to P RV

HLO1, - 25', 8.424 m to g RV

YA01T0514/02-YA01T0514/ol)

CLO2, - 60-, 10.970 m to 0 RV

HLO2, - 25%, 8.424 m to 0 RV

YA02T054/02-YAO2TO5'4/01)

HLO1, 120', 9.335 m to 0 RV

main steam line loop

main steam line loop

steam generator 01

steam generator 02

pressurizer

01

02
7

4t 24

E 20

z

C0

U> 16 -J

LU

12

8

4

0

Fig. 2. 1.

r,

kSG- STEAM GENERATOR

RCP: REACTOR COOLANT PUMP

M. MOTOR RCP

P ;PRESSURIZER

R :REACTOR VESSEL

1-

LOOP 1

L

Nuclear Power Plant Borssele:

(a) schematic

(b) elevation of components
8

70

60

m

L 50

U)

(I)

LL 40

t--0

,30

1 4 5 78 9 10 11

-eSGo2 (RAo2Pool57 -E+- SCOl (RA0iPO0l)J

13.5 14.0 14.5 15.0 15.5 16.0

TIME (HOURS)

16.5 17.0 17.5 18.0

Fig. 2.2.

30 -

20 -2

F10

10-

13.5

Measured secondary side pressure transient during

natural circulation experiment. Ref. [3].

18.0

TIME (HOURS)

Fig. 2.3. Measured temperature difference between hot and cold

legs during natural circulation experiment. Ref. [3].
9

3. RELAP5/MOD2 INPUT MODEL

3.1. Thermal hydraulic and control system model

As part of a cooperative program between ECN and the Dutch Licensing

Authority (KFD) the Netherlands Energy Research Foundation has developed a RELAP5/MOD2 input model for the Borssele NPP.

The model has extensively been used for the purpose of transient

analyses sponsored by the KFD and the operator of the plant (PZEM).

The same input model has been used for analysis of the natural circulation experiment.

The RELAP5 input model consists of 1412 volumes, 160 junctions and 85

heat structures. The corresponding nodalization diagram is shown in

figure 3.1. The two primary loops have been modelled separately and

are designated as loop 01 and 02. The pressurizer is connected to

loop 01 and the pressurizer spray lines are attached to the outlet of

the two reactor coolant pumps. The reactor coolant pumps are modelled

with the pump component of RELAP5. The homologous curves to describe

the pump behaviour under two phase conditions have been obtained from

manufacturers data as well as scaled experiments using similar pumps.

In the reactor vessel two leakage paths are represented: bypass of

the fuel elements in the core and leakage from downcomer inlet to

upperhead.

The secondary sides of the two steam generators are modelled identically. The separators and the dryers in the steam generators are represented within one single volume, at the separator location. The

steam lines from the steam generator to the turbine have been modelled separately. The system boundaries for this part of the model are

the steam bypass valves, the relief valves and the turbine valves.

The feedwater piping has been modelled starting off with the feedwater pumps.

The Emergency Core Cooling (ECC) *system of the Borssele NPP makes use

of combined injection into the primary loops (ECC water enters both

the hot and cold legs). The simplified RELAP5 ECC model injects half

of the total ECC flow in each loop and the flow is than equally divided across both injection points.
10

The Volume Control System (VCS) has been modelled and is connected to

the reactor coolant pump inlet and outlet of both ioops.

Heat structures are included to simulate the different heat sources,

heat transfer areas and stored energy. The structures connected to

the vessel model represent the fuel rods as well as the metal mass of

the reactor vessel internals and vessel wall. The heat slabs connected to the pressurizer simulate the pressurizer wall and the heater

elements. For the secondary side of the steam generators the heat

slabs represent the internals structures (separator, steam dryer,

inner shell), the steam generator outer shell as well as the U-tube

heat transfer area between primary and secondary side. The wall of

the piping of the primary circuit has not been included in the model.

The RELAP5/MOD2 input model includes control systems for the steam

generator water level, the primary pressure as well as the pressurizer level. In the analysis of the natural circulation experiment however the steam generator control system for the liquid level has not

been activated. A detailed description of the RELAP5 input model for

KCB is given in ref. [5]. A brief description of the input model is

presented in Appendix A.

3.2. Initial conditions

Just prior to the start of the natural circulation experiment in NPP

Borssele the decay heat in the core was estimated at 13 MW and both

reactor coolant pumps were running. Because of a known deficiency in

the code concerning the dissipation of pump energy within the primary

system the heat generation in the core is slightly modified for the

calculation of steady initial conditions. The RELAP5/MOD2 code only

accounts for wall friction dissipation while the form loss energy is

neglected. In the FIELAP5 model for NPP Borssele about 54i% of the pump

energy is lost in this way. In order to get a correct energy balance

514% of the power of the main coolant pumps is added to the power production in the core and the steam bypass valve is controlled to an

opening sufficient to remove decay heat and total pump energy. For

the steady state calculation is the steam generator feedwater flow
11

balanced with the steam flow so the collapsed level is kept constant

at the desired value. The steady state plant conditions as calculated

by RELAP5/MOD2 have been compared against plant data as shown in

table 3.1. The experimental values for the initial condition are

obtained from the recorded plant data just prior to the start of the

natural circulation experiment. The calculated initial conditions are

in good agreement with the measured initial conditions.

3.3. Assumptions for the simulation

Based on the calculated steady state conditions the next assumptions

have been made in order to simulated the experiment adequately.

- At time zero when the reactor coolant pumps are tripped, the core

power is reset to the desired value of 13 MW. At the same time the

steam bypass valve has been reset to a value sufficient to remove

core power and the feedwater is balanced with the steam flow. There

is no direct measurement to confirm the reset position of the steam

bypass valve. Because after the main coolant pump trip the pressure

measurement of the secondary side pressure shows an almost constant

pressure it is assumed that the opening of the steam bypass valve

is just sufficient to remove the decay heat.

- In the simulation model the feedwater control is strongly simplified. At time zero the feedwater mass flow has been reset to the

steam mass flow in order to keep the steam generator liquid level

constant. During the analysis the feedwater flow is kept constant

when feedwater is available and the feedwater flow is set to zero

when the steam generator has been isolated.

- There is no information available on the behaviour of the primary

system pressure controller during the experiment. Because relatively small pressure changes do have only minor influence on the

natural circulation behaviour a simplified pressure control is modelled. For this purpose the pressure controller in the RELAP5 model is replaced by a simple time-dependent volume which keeps the

system pressure during the test at a constant level.
12

- Due to lack of information about the mass flows of the Volume

Control System it is assumed that the primary system liquid level

controller is not in operation. The mass flows of the Volume Control System are constant during the simulation, charge and discharge flows are balanced at 4.4~ kg/s per loop.
13

Table 3.1. Experimental and calculated initial conditions

Parameter Experiment RELAP5

Core thermal power 13.00 MW 13.00 MW

Hot leg pressure 15.35 MPa 15.32 MPa

Cold leg temperature 549.0 K 548.4 K

Core temperature difference -- 0.3 K

Core mass flow -- 9623 kg/s

Pressurizer water level 4.414 m 4.4I5 m

Steam generator outlet pressure 5.86 MPa 5.8~4 MPa

Feedwater mass flow -- 5.8 kg/s

Feedwater temperature -1489.2 K

Steam generator water inventory -- 4f4300 kg

Steam generator liquid level 7.85 m 7.96 m
Figure 3.1. RELAPS nodalization of KCB
15

4.* RELAP5/MOD2 ANALYSIS OF THE NATURAL CIRCULATION EXPERIMENT

In this chapter the RELAP5/MOD2 results are presented as obtained

from the analysis of the natural circulation experiment. Furthermore

an explanation of the occuring phenomena is given.

The analysis starts with the steady state condition for KCB at nominal flow conditions and a core thermal power of 13 MW. The sequence

of events of the analysis is given in table 4.1 while the main parameters of analysis are presented in figure 4.1 through 4.7.

The main driving forces for natural circulation are the differences

in hydrostatic fluid pressure between downcomer and core within the

reactor pressure vessel and between the upward and downward legs of

the steam generator U-tubes and in the loop seal. The hydrostatic

pressure difference terms are given by summation of the "pgh" terms

for each of the three vertical components. In Appendix B a more comprehensive explanation of the driving forces is given.

An explanation of the calculated system behaviour will be given next

for the different periods as defined in table 4.1 . In the analysis

of the results the figures 4.1 through 4.7 are extensively used but

no reference is given in the text.

0-8140 s.

After trip of the reactor coolant pumps at time zero, the pumps coast

down and a natural circulation mass flow of about 165 kg/s is established in both loops. The temperature difference across the core

increases from 0.3 K towards 7.7 K. The driving force for the natural

circulation is generated inside the core and in both steam generators. Heat transfer into and out of the primary system are balanced;

almost 95% of the total energy generated in the core is removed

equally by the steam generators while the remaining 5% is lost in the

Volume Control System. At the end of this period almost steady state

conditions exist inside the primary system.
16

840- 39 00 s.

At time = 8410 s steam generator 01 is being isolated (closure of' the

main steam isolation valve and interruption of feedwater). The secondary pressure in steam generator 02 is kept constant by the pressure

control system in the model. After isolation decreases the heat

transfer to steam generator 01 to zero and all the heat generated in

the core is removed from the system through steam generator 02. After

redistribution of the heat transfer is established the system converges to a steady state with equal hot leg temperatures in the loops

and no driving force in steam generator 01. The pressure at the secondary side of the isolated steam generator levels out close to the

saturation pressure of the hot leg temperature. The natural circulation in loop 01 is caused only by the driving force from the reactor

vessel while in loop 02 a larger natural circulation is established

due to the existence of a driving force from reactor vessel core and

steam generator 02. The mass flows in loop 01 and 02 resp. level out

at 96 kg/s and 202 kg/s.

3900-14260 s.

During this period the primary system is cooled down at a rate of 100

K/hr by depressurizing steam generator 02. The cooling down in steam

generator 02 causes an almost immediate increase of the driving force

in steam generator 02 (outlet of steam generator 02 decreases in temperature while the inlet has a constant temperature). The colder

fluid moving through loop 02 causes delayed also an increase in driving force in the core. Again somewhat later in time the colder fluid

arrives at the inlet of steam generator 01. Because the outlet of the

isolated steam generator is constant and the inlet temperature decreases, a decrease to a negative driving force in steam generator 01

is created. This negative driving force in steam generator 01 compensates the net driving force of the core resulting in a stagnation in

loop 01. The 360 s cooling down time is not sufficient to reach a

steady state condition. Due to the relative long "refresh" time of

the loops and the core (due to the small mass flows it lasts long to

replace the water in a component) the effect goes on after discontinuation of the cooling down..
17

L126o-Sl4oo s

At time = 14260 s the cooling down of the primary system is discontinued and the secondary side pressure is controlled constant at its

last value. The maximum negative driving force in the steam generator

01 is set by the temperature difference between the secondary sides

of the two steam generators. The primary fluid entering steam generator 01 has a temperature equal to the secondary side fluid of steam

generator 02 while the fluid leaving steam generator 01 has a temperature equal to the secondary side fluid of steam generator 01. The

maximum (negative driving force) will not be reached because a smaller temperature difference already causes a negative driving force

resulting in an almost stagnant mass flow in loop 01. At the end of

the period an almost steady state condition is reached with a mass

flow in loop 01 of 5 kg/s and in loop 02 of 222 kg/s.

51100-6720 s.

At time = 54100 s the isolation of steam generator 01 is discontinued

and a return to the situation described in time frame 0-8410 s is

observed. Discontinuation of the isolation causes a rapid decrease of

the primary fluid temperature in the U-tubes of steam generator 01.

The effect of this temperature decrease is a shrinkage of the fluid

followed by a change of the temperature gradient in the U-tubes. Due

to the shrinkage there is an immediate increase of the mass flow

through steam generator 01. The resulting change of the temperature

gradient in steam generator 01 causes an increase of the driving

force and this again gives another increase of mass flow in loop 01.

After some oscillations in driving forces in core and steam generator

region, due to "refresh" time of the components of the primary

system, a steady state condition is established. The steady state

mass flow in both loops is 163.5 kg/s and the core temperature difference is 7.8 K. This condition is very close to the mass flows in the

loops and the core temperature difference observed at the end of time

frame 0-8410 s.

6720-786o s.

At time = 6720 s steam generator 02 is being isolated (closure of the

main steam isolation valve and interruption of feedwater). The obser-
18

ved behaviour is symmetric compared to the behaviour in the period

8410-3900 s. The total time in this period (11140 s) is to short to

reach a steady state condition but the system behaviour is a reflection of the observed behaviour in period 8410-3900 s. The loop mass

flows at the end of the period are 188 kg/s and 117 kg/s while the

core temperature difference 8.3 K is.

7860-9960 s.

At time = 7860 s the primary system is cooled down with a gradient of

50 K/hr during 2100 s by means of depressurization of steam generator

01. The behaviour during this cooling down phase is similar to the

system behaviour during the time interval 3900-41260 s. Cooling down

of steam generator 01 causes a temperature decrease of the fluid in

the hot leg of loop 02. The decreasing fluid temperature at the inlet

of steam generator 02 and the constant fluid temperature at the outlet causes a decreasing driving force in steam generator 02 and so a

decreasing mass flow in loop 02. Due to the small mass flow in loop

02 and the injection of relatively cold water from the Volume Control

System into the cold leg, a strong increase of the temperature difference across loop 02 is observed.

Just prior to the end of the period the driving force in steam generator 02 starts to increase while the driving force of loop seal 02

starts to decrease. This effect is caused by the onset of a negative

mass flow in loop 02. An additional analysis in which the boundary

conditions are maintained beyond t =9960 s shows the following behaviour. The slow decreasing fluid temperature at the inlet of steam

generator 02 causes a decreasing mass flow. When the mass flow gets

negative it sudden decreases to about -35 kg/s and then increases

back to zero and stagnates. This all happens in the time between 9750

s and 10150 s. The mass flow behaviour can be explained by the driving forces in loop 02. The reversed flow causes, due to cold water

of the Volume Control System, a negative driving force in the loop

seal of loop 02 and an increasing driving force in steam generator

02. For the positive driving force in steam generator 02 and the core

it is impossible to overcome the large negative driving force in loop

seal 02 resulting in a stagnant mass flow.
19

9960-11220 s.

At time =9960 s the cooling down of the primary system by means of

steam generator 01 is interrupted and the pressure on the secondary side of the steam generators is kept constant. In this period

the system behaviour is more or less a reflection of the behaviour

during time period 4260-54l00 s and moreover a continuation of the

behaviour observed at the end of the previous period. A small reversed mass flow in loop 02 pushes cold water from the Volume Control

System inlet into the steam generator leg of the loop seal causing a

decreasing driving force in the loop seal. Moreover, relatively cold

water is pushed into the downward leg of steam generator 02 causing

an increase in driving force in this region. The result of the driving forces in core, steam generator 02 and loop seal 02 region is a

very small (1-2 kg/s) negative mass flow in loop 02. The time period

is to short to reach a steady state condition in loop 01, but the

system moves towardt a zero mass flow in loop 02. Due to the decrease

in driving force in loop 01 (less heat transfer in steam generator

01) the mass flow in loop 01 decreases from 250 kg/s to 223.5 kg/s.

11220-13020 s.

At time = 11220 s steam generator 01 is also isolated. During this

part of the transient both steam generators have been isolated and

the heat generated in the core is added to the primary system. After

both steam generators have been isolated the fluid temperatures in

loop 01 and the secondary side of steam generator 01 increase. Due to

the heating up of the secondary side of steam generator 01 decreases

the driving force in steam generator 01 drastically. This driving

force added to the driving force of the core maintains a smaller natural circulation in loop 01 (180 kg/s). The stagnant loop 02 is not

affected by the heating up of the primary and secondary side of loop

01.

13020-13740 s.

At time = 13020 s isolation of steam generators 01 and 02 is discontinued by opening of both main steam isolation valves and restoring the feedwater flow. After the main steam isolation valves have

been opened the pressure on the secondary side of the steam genera-
20

tors is kept constant. By coincidence are the secondary side pressures of both isolated steam generators equal just prior to discontinuation of' the isolation (t = 13020 s). So the reconnection of the

steam generators to the main steam line does not give a large discontinuity in heat transfer in one of the steam generators. Because

steam generator 02 has a stagnant flow all the heat produced in the

core must be removed by steam generator 01. The increasing heat

transfer in steam generator 01 causes an increase in driving force

and so in mass flow in loop 01. The time period is to short to reach

a steady state condition but the natural circulation mass flow in

loop 01 is-of the magnitude of 215 kg/s. In contrast with the system

behaviour at time =51400 s is the mass flow in the stagnant loop not

restored. The main reason for the difference in system behaviour is

the pressure level at the secondary side of the steam generators at

which discontinuation of the isolation takes places. The main steam

isolation valve is being closed when the two secondary side pressures

are almost identical and so a discontinuity in the primary system

fluid temperature is avoided.

13740-149110 s.

After discontinuation of the isolation of the steam generators in the

previous time period remains the mass flow in loop 02 stagnated. In

order to restore the mass flow steam generator 01 is being isolated.

The behaviour of the system is a copy of the system behaviour after

time = 8410 s. However due to stagnation in loop 02 all the heat has

to be removed by steam generator 02. This. causes a more rapid

increase of the primary system temperature in loop 01 and secondary

pressure of steam generator 01.

141940-15900 s.

At time = 141940 s the isolation of steam generator 01 is discontinued. The pressure at the secondary side of steam generator 01 drops

to the setpoint value of the pressure controller and a strong cooling

down by steam generator 01 is introduced.

The cooling down of primary system fluid in steam generator 01 causes

shrinkage and a strong increase of the driving force in the steam

generator resulting in an increasing mass flow in loop 01. The

increasing mass flow causes an increasing driving force in the core
21

resulting in a positive mass flow in loop 02. This mass flow ends the

negative driving force of the loop seal in loop 02 resulting in an

increasing mass flow. After some oscillations the natural circulation

in both loops is re-established. At the end of the period the mass

flow is 158 kg/s in both loops.

The general behaviour of the primary coolant system during the natural circulation experiment can be characterized as follows:

After the trip of the reactor coolant pumps natural circulation is

established in the loops. After a steady state condition has been

established in the primary system a simple relation exists between

the natural circulation mass flow and the heat production in the core.

M= C Q/

m = natural circulation mass flow in the core

Q = heat production in the core

C = constant

The relation exists when all the heat produced in the core is removed

equally by the steam generators. In Appendix B a derivation of the

relation is presented.

Isolation of a steam generator while keeping the average primary

system temperature constant (all the heat produced in the core is

removed by the non-isolated steam generator) maintains a natural circulation mass flow in the loops of the primary system. The analysis

shows, at an almost constant core mass flow, a redistribution of the

mass flows over the loops. The redistribution of the mass flows over

the loops strongly depends on the distribution of the hydraulic resistances over the system.

Cooling down of the primary system by depressurizing one of the steam

generators results in the RELAP5 analysis in stagnation of the mass

flow in the other loop. The cooling down rate sets the time to get to

flow stagnation.
22

Possibilities to restore the mass flow in a stagnant loop depend

strongly on the temperature distribution in the loop. In the analysis

this sensitivity is shown by the influence of the location of the

Volume Control System injection line. Cold water is added close to

the loop seal and the temperature distribution in this vertical

oriented part of the system is important for the driving force. A

negative driving force in this part of the primary system has to be

overcome by *the rest of the driving forces in order to restore mass

flow in the loop. The analysis shows that a sudden increase of heat

transfer in one of the steam generators (depressurization of the

steam generator) is able to restore the natural circulation in the

stagnant loop. Probably a sudden decrease of heat transfer to a steam

generator will show the same effect. However, a sudden pressure

increase on the secondary side of the steam generator is hard to realize.
23

Table 4.1. Sequence of' events for simulation of the natural circulation experiment

Time(s) Event

0.0 Reactor coolant pumps tripped.

8110.0 Main steam valve on steam generator 01 closed and the

feedwater of steam generator 01 turned off (isolation of

steam generator 01).

3900.0 Cooling down of steam generator 02 with 100 K/hr. A

controlled pressure reduction with a constant feedwater

mass flow in the affected steam generator provides cooling down of the primary system.

4140.0 Injection of auxiliary feedwater. This step is neglected

in the analysis due to lack of information about the

auxiliary feedwater mass flow.

4260.0 Discontinuation of the cooling down of the primary

system.

5400.0 Isolation of steam generator 01 discontinued by opening

of the main steam valve and re-establishing the feedwater

flow.

6720.0 Main steam valve of steam generator 02 closed and

feedwater off (isolation of steam generator 02).

7860.0 Cooling down of steam generator 01 with 50 K/hr

A controlled reduction of the pressure in the steam generator provides the cooling down of the primary system.

9960.0 Discontinuation of the cooling down of steam generator

01.

11220.0 Main steam valve steam generator 01 closed and feedwater

discontinued (isolation of steam generator 01). At this

point both steam generators are isolated.

13020.0 Isolation of steam generator 01 and 02 discontinued by

opening of the main steam valves and restoring the feedwater flow.

13740.0 Isolation of steam generator 01 (compare event at t = 8410

S).

1494o0.0 Isolation of steam generator 01 discontinued (compare

event at t = 5'400 s).

15300.0 End of test.

15900.0 End of the calculation.
24

0

o E-ELOOPO01 __ &†LOOPQ02 CO0

0;

U)

<9~~ P____ 4= ApA

0Q

0.0 4.0 8.0 12.0 16.0

TIME (SEC)

Figure 4.1. RELAP5 mass flow rate inside loop 01 and 02

(6.......................................................

U)o

0' ST. GEN. 01

ST. -...GEN. 02

0.0 4.0 8.0 12.0 16.0

TIME (SEC) *1O,

Figure 4.2. RELAP5 steam generator secondary side pressures
25

00

Z%

09

zo

U I I

CORE .. .. LOOP 01 ---LOOP 02

II

p , S

fit

.1

....... .. .. . .. .. . ..

0.0 4.0 8.0

TIME (SEC)

12.0 16.0

*10,

Figure 4.3. RELAP5 temperature gradients in primary system loops and

core

I

TIME (SEC)

16.0

*1. d

Figure 4.4. RELAP5 hot and cold leg fluid temperatures in loop 01
26

0

0

0

E-' -1 -

L0

.... LG COL L .

,)0

----ý STGE.0

0

Co ~ ~ ___ ~ __ ~ __ ~ ~ ~ _ _ . ........ _ [..__ __ _ _

0.0 4.0 8.0 12.0 16.0

TIME (SEC) *d

Figure 4 .6. RELAP5 energy balance for primary system
27

0

0

0--ST. GEN. 01

A---bST. GEN. 02

W 0+--+-LOOP SEAL 02

0. 4. 8.0- 12.016. TIE(SC

FiuwCQ*RLP riigfre o aualcruaini ta

generators, loop seals and core
28

5. COMPARISON OF RELAP5 CALCULATIONS TO EXPERIMENTAL DATA

A quantitative comparison of the measured data and the calculated

results is presented in this section. The comparison can not be very

detailed due to the limited number of measurements and the construction, location and measurement range of the devices. Moreover there

is no data qualification of the measured data available. The data

channels which are available from the strip charts are given in table

2.4. From the available data channels presented in table 2.4 the marked (*) ones did not show major problems during the test and will be

used for comparison with the RELAP5 calculations.

In table 5.1 the designation of the calculated and experimental data

is given together with the figure numbers where the comparison is

presented. Fig. 5.1 through 5.12 show overlays of the measured and

calculated data. The parameter labels in the figures are described in

tables 2.4I and 5.1. For completeness are the measured and calculated

liquid levels in the steam generators and the system pressure presented. However, lack of information made it necessary to simplify the

controllers and for this reason the comparisons are not meaningful

for assessment purposes.

A short description of the comparisons for the different time periods

will be given next.

Time 0-840 s.

The temperature behaviour during this first period in the calculation

and in the experiment is very similar. There is an increase in hot

leg temperature and an almost constant cold leg temperature.

Time 840-3900 s.

The calculated temperature behaviour during this period is in reasonable agreement with the measured values. An exception is the deviation in the temperatures in the hot leg of loop 01. An explanation

for this deviation is an excessive spray into the pressurizer during

part of this period. Because of the spray hot water from pressurizer
29

is pushed into the hot leg of the primary system. At the end of the

period the spray is turned off, but there is no measured confirmation

of the spray flowrate and the timing of the discontinuation of the

flow. In general the calculated secondary side pressure in the isolated steam generator 01 increases to fast. A possible reason for this

deviation is the heat loss from the wall of the steam generator to

the environment. This effect is not modelled in the RELAP5 calculation.

Time 3900-4260 s.

In the calculation the temperature decrease during this time frame is

controlled by the depressurization rate of steam generator 02. At the

secondary side of steam generator 02 the calculated temperature is

very close to the measured value. Due to the small mass flow in the

loop of the isolated steam generator the cooling down effect on the

fluid temperature in this loop is delayed into the next time frame.

The calculated liquid level in the pressurizer decreases faster (about double) than the measured level. A possible cause of this deviation is an uncertainty in the control function for the mass flow of

the Volume Control System.

Time 4260-5 4100 s.

During this time frame the cooling down effect of the previous period

is shown in the fluid temperatures in loop 01. A strong temperature

decrease is observed in the calculated cold leg temperature of loop

01. This decrease is caused by the injection of relatively cold water

from the Volume Control System into the cold leg. The calculated temperatures in loop 02 are close to the experimental values.

Time 54~00-6 720 s.

During this part of the transient (natural circulation in both loops)

the experimental and calculated values are very close.

Time 6720-7860 s.

The pressure and temperature behaviour in steam generator 02 and in

the loops is identical to the behaviour for steam generator 01 and

the opposite loops during the period 8140-3900 s. The calculated pres-
30

sure increase in steam generator 02 is during this period also fast

compared to the experimental value and is being caused by the same

effect as in period 8140-3900 s.

Time 7860-9960 s.

The calculated and measured temperatures in loop 01 are in good

agreement. However, a significant deviation is observed in the fluid

temperatures in loop 02. The calculated temperature decrease in the

hot leg of loop 02 is relatively small while the cold leg temperature

decrease is relatively large. A reason for this different temperature

behaviour is the very small and decreasing mass flow in loop 02 and

the injection of cold water from the Volume Control System into the

cold leg. The liquid level in the pressurizer again decreases too

fast compared to the measured one.

Time 9960-11220 s.

The calculated temperatures in loop 01 are close to the measured

temperatures. In loop 02 however deviations in the fluid temperatures

still exist due to the same reasons as described in previous period.

Time 11220-end.

The temperature and pressure comparisons after about 11000 s are not

realistic. In the experiment the nature and timing of the different

events are not well described and hence a comparison of calculated

and measured data is not meaningful.

In Appendix B an analytical solution of natural circulation in a

simplified model is presented together with a comparison of the

analytical and with RELAP5 calculated solution. The Appendix shows

that RELAP5 calculates the natural circulation mass flow well. With

the assumption that 8W. of the heat is transferred in the upward leg

of the steam generators the maximum error in the calculated mass flow

due to nodalization is about 3%. Also presented in Appendix B is the

importance of the initial temperature distribution and nodalization

on the mass flow behaviour.
31

Although there is no direct measurement of the mass flow in the core

or in the loops a thorough investigation of all the measured temperatures confirms stagnation of the mass flow during some parts of the

transient, ref. [3]. The mass flow stagnation is observed at the same

location and time as in the calculation.

A general conclusion of the comparison between calculated and measured data is the good agreement in the temperatures in the non-s tagnant loop. In the loop with a small or stagnant mass flow deviations

between the calculated and measured fluid temperatures are observed.

Reasons for the deviations are the injection of relatively cold water

of the Volume Control System and a probable stratification of the

mass flow close to the measuring device.

Isolation of a steam generator results for the RELAP5 analysis in a

too fast increase of steam generator secondary side pressure and primary system temperature. Heat loss from the steam generator to the

environment and uncertainties in the mass inventories of the steam

generators are reasons for the deviations. Another uncertainty is the

decay heat in the core. The power in the calculation is fixed to

13 MW but there is no confirmation by a measurement of the power.

The behaviour of the calculated liquid level in the pressurizer is

close to the measured value. However the calculation shows a more

pronounced control action than is observed in the test. A reason for

the deviation is uncertainties in the controller and pump characteristics of the Volume Control System.
32

Table 5.1. Designation of' experimental and calculated date

Experiment RELAP5 Figure no.

YA01T090 TEMPF 185010000 5.10

YA02TO90 TEMPF 285010000 5.12

YA01T098 TEMPF 110010000 5.9

YA02T098 TEMPF 210010000 5.11

YAO1T054 CNTRLVAR 4090 5.7

(YA01TO54I/01 - YA01TO54/02) (TEMPE 110010000 - TEMPE 185010000)

YA02T054 ONTRLVAR 4091 5.8

(YA02T054/01 - YAO2TO5LI/02) (TEMPF 210010000 - TEMPF 285010000)

YA01P002 P 110010000 5.1

RA01POOl P 596010000 5.5

RA02POOl P 696010000 5.6

YBO1IO51 CNTRLVAR 109 5.3

YB02LO51 CNTRLVAR 110 5.4

YPO1L11 CNTRLVAR 108 5.2
33

IR V-4 a---EEXP. YA01P%02

to'

0D

0.0 4.0 8.0 12.0 16.0

TIME (SEC) *10J

Figure 5.1. Comparison of calculated and measured system pressures

0d

0-

o: a -...EJEXP. YB02LO51

6o CNTRLVAR 110

0 1

0.0.0 4.0 8.0 12.0 16.0

TIME (SEC) *d

Figure 5.2. Comparison of calculated and measured collapsed liquid

levels in the pressurizer
34

CO,

> C,2

TIME (SEC)

Figure 5.3. Comparison of' calculated and measured collapsed liquid

levels in steam generator 01

I

8.0

TIME (SEC) *1d'

Figure 5 .4.* Comparison of' calculated and measured collapsed liquid

levels in steam generator 02
35

,Co) __ _ _ _ _ _ _ _ _ _ _ _ _ _ _

r100

U)

a 0---EJEXP. RAO1POOl

-P 596010000

0.....

0

C'ý _ _ _ _ _ _ _ _

0.0 4.0 8.0 12.0 16

TIME (SEC) *1

Figure 5.5. Comparison of calculated and measured pressures at secondary side of steam generator 01

00

Figure 5.6. Comparison of calculated and measured pressures at secondary side of steam generator 02
36

0

F-1'

0--- BEXP. YA01T054 - CNTRLVAR 409C

0.0 4.0 8.0 12.0 1

TIME (SEC)

Figure 5.7. Comparison of calculated and measured temperature differences across loop 01

0

0.

0

C0.

-4 .

I

0D

6;

8.0

TIME (SEC) *10,

Figure 5.8. Comparison of calculated and measured temperature differences across loop 02
37

a

0

E-1 'CQ

0

0;

0

LO

0.00

Figure 5.9.

4.0 8.0 12.0 16.0

TIME (SEC) *iOP

Comparison of calculated and measured fluid temperatures

in the hot leg of loop 01

0

0;

00

2ýg

LO~

0.0 4.0 8.0 12.0

TIME (SEC)

16.0

*1 03

Figure 5.10. Comparison of calculated and measured fluid temperatures

in the cold leg of loop 01
38

0

C;

124LO

E-1 6

I

0.0 4.0 8.0 12.0

TIME (SEC)

16.0

*10

Figure 5.11. Comparison of calculated and measured fluid temperatures

in the hot leg of loop 02

0

0

CD

LO

0

E-O

0

a

8.0

TIME (SEC)

16.0

*10,

Figure 5.12. Comparison of calculated and measured fluid temperatures

in the cold leg of loop 02
39

6. CONCLUSIONS

Generalily the RELAP5 analyses of the natural circulation experiment

in Nuclear Power Plant Borssele showed qualitatively good agreement

with the test data. However, lack of qualified and relevant data from

the test (typical for plant transients) prevented a quantitative

assessment.

A deficiency in the energy equation in RELAP5 causes a deviation in

energy balance and steady-state. Adding part of the "pump power" to

the "core power" solved the problem.

RELAP5/MOD2 simulates the primary system behaviour under natural circulation conditions in Nuclear Power Plant Borssele adequate. The

calculated mass flow and temperature difference across the loops

agree in general well. Also the observed mass flow stagnation in the

experiment is well predicted by the code. In general mass flow stagnation in a loop with an isolated steam generator occurs when the

system is cooled down by the intact steam generator. Restoration of

the mass flow in a loop requires a large and sudden heat transfer

increase in one of the steam generators.

The comparison of measured and calculated temperatures shows a good

agreement as long as a mass flow exists, during stagnation the temperatures deviate. The observed temperature differences are probably

caused by uncertainties in the Volume Control System behaviour. Also

stratification at the location of the measuring devices can play a

role.

The secondary side pressure behaviour during isolation is identical

in calculation and experiment. The slower increase of the pressure at

the secondary side of the steam generator during isolation is probably caused by uncertainties in feedwater and steam mass flows, heat

losses from the steam generators to the environment and mass inventory uncertainties of the steam generators. A comparison between an

analytical solution and a RELAP5 calculation for a simplified reactor

model shows that the used RELAP5 nodalization in core and steam generators generates a nodalization error less than 3%.
40

7. REFERENCES

[1] Ransom, Victor H. et al.: RELAP5/MOD2 Code manual. Volume 1: Code

Structure, Systems Models and Solution Methods. Volume 2: Users

Guide and Input Requirements.

NUREG/CR-13 12.

[2] Guidelines and Procedures for the International Code Assessment

and Applications Program

U.S. Nuclear Regulatory Commission.

NUREG-1271.

[3] Kalverboer, C. (PZEM): Natural Circulation and Heat Transfer

Tests in the Borssele Nuclear Power Plant;

KVB/IB/NCIRC, July 11, 1983. (Internal report)

[4i] Technical Note: Natural Circulation Tests in the Borssele Nuclear

Power Plant;

Nuclear Safety, Vol. 25, No. 3, May-June 19841.

[5] L. Winters: RELAP5 model KCB (in progress).
A-1

Appendix A. Description of the RELAP5 model for KCB

Introduction

This Appendix gives a brief description of the RELAP5 model for Nuclear Power Plant Borssele (KCB). A comprehensive description of' the

model with assessment calculations is in progress (ref. [5]). The

model was developed at ECN for the Dutch Licensing Authority (KFD).

The model has been used succesfully for different transient calculations with the simulation code RELAP5.

Reactor coolant systern

The overall RELAP5 nodalization scheme of KCB is presented in figure 3.1 of the report. A more detailed nodalization of the different

components is given in figure. A-i through A-5. In the figures the

correspondence between the physical and the mathematical components

is shown by means of a schematic drawing of the component and the

corresponding RELAP5 nodalization scheme. The loops are designated as

loop 01 and 02 with the pressurizer connected to loop 01. Figure A-i

gives the nodalization scheme for loop 01. Loop 02 is identical to

loop 01 except that loop 01 has a connection with the pressurizer

through the surge line. The pressurizer is represented in figure A-2.

The heater elements in the pressurizer provide a maximum power of

1.660 MW. The Pilot Operated Relief Valves (PORV) and the safety valves on the pressurizer are represented in the model by a servo valve.

The pressurizer spray valves are modelled as a servo valve with six

different opening positions, which are controlled by the primary

system pressure. The maximum spray capacity is about 214 kg/s.

Figure A-3 shows the RELAP5 vessel configuration and nodalization.

Two minor flow paths are modelled viz, the bypass of the core and the

leak path from inlet annulus of the downcomer to the upper head. The

flow through the two leak paths is resp. 5.9% and 1.5%'0 of the total

mass flow through the reactor vessel.

The steam generator primary and secondary sides are represented in

figure A-14. The separators and the dryers in the steam generator are

modelled in one separator model at the seperator location. The steam
A-2

line with its components (safety valves, relief valves, steam bypass

station and turbine valves) are all modelled. The feed water pumps

and the feed water control valves are also included in the model. The

auxiliary feed water system is simplified and modelled as a single

fill junction. In the model the turbine valves and the feedwater

tanks represent the boundaries of the steam generator.

Added to the model of the primary system are the Volume Control

System (VCS) and the Emergency Core Cooling System (ECCS) (figure A-5). The VCS is represented in 4{ time dependent junctions

(charge flow in bottom of the loop seal). The ECCS includes the High

Pressure Injection System (HPIS), accumulators and the Low Pressure

Injection System (LPIS). The ECCS is connected to both hot legs (connection between reactor vessel and steam generator) and the cold legs

(lowest part of the loop seal).

Heat structures are added to the model to represent the fuel rods,

the wall of reactor vessel and steam generators and internals in the

steam generators and the reactor vessel.

Control systTem

Added to the RELAP5 model are the most important reactor control

systems viz, primary system pressure control system, level control

system in the pressurizer and the level control system in the steam

generators.

The pressure in the primary system is controlled by the heater elements and the spray in the pressurizer. The adjustment of these components is set by control elements which have the primary system

pressure as input signal.

The liquid level in the pressurizer is controlled by the mass flows

of the Volume Control System. Signals representing the Volume Control

System charge- and discharge flow and the actual level in the pressurizer are used as input for a proportional integral controller. This

P1-element controls the position of the discharge flow valve of the

Volume Control System. The charge flow of the Volume Control System

is constant.
A-3

The level control in the steam generators is similar to the pressurizer level control. The feed water valves are controlled by a P1-controller which has as input signals the feed water mass flow, the

steam flow and the deviation of the steam generator actual level and

the level setpoint.
A-4

Figure A-1. Representation of loop 01

a) Configuration

b) RELAP5 nodalization diagram
4~3700

S

40500

42500

LOOP 01

-LOOP 02

LOOP 01

Figuire A-21. Rlepresentation of the pressurizer

a)~ Cofgration

b) BI3EAP5 nodalization diagram
I

HL LOOP 02

CL LOOP 02

H-L LOOP 01

CL LOOP 01

Figure A-3. Representation of' the reactor vessel

a) Configuration

b) RELAP5 nodalization diagram
SG 02 -

69300 -

59500 -

58500 -

54001 -

54003--

52200 -

52100 -

51003 -

51002 -

51001 -

HOT LEG

PCs

02

Figure A-11. Bepresentation of the steam generator

a) Configuration

b) RELAP5 nodalization dingram
ACCUMUL ATOR

H-0.7 864tPI

885HP1 860

I885(LPII 880

I ~881.

Figure A-5. Representation of ECO system

a) Configuration

b) BELAP5 nodalization diagram
B-1

Appendix B. Analytical model for natural circulation

In a single loop a simple relation exists between the steady state

natural circulation mass flow and the heat production. Figure B-i is

a schematic simplified representation of one loop of a reactor

system. In the core the generated heat (Q) is added to the fluid and

in the steam generators Q 1 (upward leg) and Q 2 (downward leg) is

transfered from the primary system fluid to the secondary side of the

steam generator. During steady state conditions is the generated

heat equal to the removed heat (Q I + Q2= Q). The driving force for

the natural circulation is given by

AP' = fpgdz (B.1)

where:

AP = driving pressure (N/in2)

p = density of the fluid (kg/in3)

g = acceleration due to gravity (9.81 m/s2 )

z = vertical ordinate (in)

In evaluating the integral (B.1) the next relations are important

Ap = - p AT (B.2)

AT = Q /ii C p(B-3)

AP' = j (k + k f) jh / A A2 (B.4)

where:

Ap = change in density due to temperature 1change (kg/in3)

P= coefficient of thermal expansion (K

AT = temperature change (K)

ih = mass flow (kg/s)

Q = heat transfer (J/s)

C p= specific heat capacity of the fluid at constant pressure

~'(J/kg K)

k = total form loss coefficient of the circuit

k f= total loss factor due to friction

A = flow area in relation to loss factor (in2 )
B-2

For a RELAP5 calculation the schematic loop of' fig. B-i has to be

divided into subvolumes (fig. B-2). Assume the "core"' is subdivided

in N cvolumes and both the upward and downward leg of the "steam generator" in N volumes.

For the analysis the RELAP5 assumption is important that the fluid

density in a volume is defined by the outlet condition of the volume.

Evaluating the above mentioned integral (B.1) according to the RELAP5

code, with the assumption of a clockwise directed mass flow, gives

the next relation for the driving force.

-- h h hhQ

AP - ( + h + -- )Q + -EQ Q + (B-5

mC 222 N 2Nc

p

P, p and C pare mean physical constants for the fluid in the system

and h, h cand h sare resp. height of hot leg, core and steam generator (fig. B-2). Starting from eq. (B.5) and using the relations

(B.2), (B.3) en (B.4) the natural circulation mass flow is given by:

;3 2 ý -2 A2 h h h h Q

2- c( + h + 2 )Q + 2 !Q N Q2+2 (B.6)

(k + k f)C pc

The first term in this relation represents the analytical solution

and the last two terms represent an error introduced by the RELAP5

code. The error is caused by the fact that the system is subdivided

in discrete volumes in connection with the RELAP5 assumption with

respect to the volume density. The last two terms disappear and so

the error goes to zero when N and N cget infinite.

With the simple model of figure B-2 two sets of RELAP5 calculations

are performed.

1. One set of calculations with different nodalization in core and

steam generators.

2. One set of calculations with different power distributions for Q,

and Q2 in the steam generators.
B-3

Ad 1. Nodalization sensitivity

Table B-i gives the results of a RELAP5 calculation and the analytical solution according to eq. (B.6) for the mass flow for different

nodalizations. The power distribution for all the cases is set fixed

viz. Q = 13.0 MW, Q, = 6.5 MW and Q2 = 6.5 MW and the different

heights are h c= 2.6 rn, h = 3.6 rn and hs = 7.6 rn. The system fluid

temperature is 530 K. The analytical solution of eq. (B.6) is based

on the next average physical properties:

= 1.7 4~ lo K_1

(k f + k)/A 2= 36.9 mrn

g 9.8 rn/s 2

C p= 4~795 J/kg

p = 795 gm

The solution of eq. (B.6) and the RELAP5 mass flow calculation are

identical for different nodalizations (small deviations are due to

the averaging proces of the physical properties). The analytical solution for the simple model is 239.6 kg/s (eq. (B.6) with N, N c4 .

The nodalization of the steam generators and the core in the standard

RELAP5 model for Nuclear Power Plant Borssele (N = 4j, Nc 6) is based a.o. on efficiency considerations. The applied nodalization mesh

size is a good choice, a maximum error of 1.5% can be expected due to

nodalization.

Ad 2. Variation in power distribution

Table B-2 gives the results of a RELAP5 calculation and the analytical solution according to eq. (B.6) for the mass flow for different

power distributions. The nodalization is the standard one N c= 6

(core divided in 6 subvolumes) and N = 4~ (up and downward side of the

steam generator divided in 4i volumes). The calculation started with

relatively cold water at the steam generator outlet and hot water at

the core outlet resulting in an average fluid temperature of about

530 K.

Table B-2 shows that the analytical and RELAP5 solutions are

identical. The analytical solution according to eq. (B.6) and with

N and N c 4 for this problem is 239.6 kg/s. based on average fluid
B-4

physical properties. Assuming (0,25 < Q1/Q < 0,75) a maximum deviation of about 15% can be seen between the RELAP5 calculated mass flow

and the analytical solution for the mass flow.

Combining the seperate effects of nodalization and power distribution

shows a maximum nodalization error of 3%.

Important in the RELAP5 analysis of a single loop is the possibility

to calculate a reversed natural circulation flow. A reversed flow in

the simple model will be calculated by starting with an isothermal

system and Q 1»>Q2

Table B-1. Steady state natural circulation

nodalization schemes

mass flow for different

number of volumes number of volumes mass flow mass flow

in the core N c in steam generator N RELAP5 eq. (B.6)

(kg/s) (kg/s)

1 1 251.1 251.1

2 2 245.2 245.1

4 4 242.1 242.7

6 6 241.1 241.7

Table B-2. Steady state natural circulation mass flow

power distributions.

for different

heat fluxes (MW) RELAP5 mass flow

Q QQmass flow eq. (B.6)

(kg/s) (kg/s)

13 -- 13 248.9 250.0

13 3.25 9.75 245.1 245.9

13 6.5 6.5 241.1 241.7

13 9.75 3.25 237.1 237.3

13 13 -- 232.9 232.8
B- 5

U-TUBES

C ORE PM

FIG.B- 1 SCHEMA TIC OF A LOOP OF A PWR

TTh

TT

FIG. 8 -2 REL A P5 MODEL OF A LOOP OF A P WP

C-1

Appendix C. Run statistics

The version of the simulation code used was RELAP5/MOD2, CY36.05.

The code is implemented on a CDC Cyber 170-855 computer at the

Technical Computing Center (ENR) at Petten, The Netherlands.

The run statistics for the documented analysis are given by the next

numbers:

CPU =total execution time(s) =9241

C =number of volumes = 114o

DT = number of time steps = 18912

RS =millisec. per volume per time step = 3.1490

In figure C-i the CPU-time is given as a function of the transient

time. The CPU-time parameter in RELAP5 is reset to zero for each

restart job. This explains the several drops back to zero in the

CPU-time history.

In figure C-2 the time step size is given as a function of the transient time. The specified maximum time step size for the entire

transient was 10 s.

WI-88/1936yv
C-2

0

0.

0

0

00

0 0 0801.01

TIME (SEC) *0

Fig. C-I. CPU-time as a function of real time

Cf)

U1

0.

00 4.0 8 0 12.0 16.0

TIME (SEC)*1

Fig. C-2. Time step size as a function of real time
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I. REPORT NUMBER

12jC191 (Assigned by NRC. Add Vol.. Supp., Rev.,

1101.and Addendum Numbers. it any.) 3201.3202 BIBLIOGRAPHIC DATA SHEET

(See instrucvions on the aeverse) NUREG/IA-0091

2. TITLE AND SUBTITLE ECN-89-91

Assessment of RELAP5/MOD2 Against a Natural Circulation Experiment 3. DATE REPORT PUBLISHED

in Nuclear Power Plant Borssele MONTH *1EAR

July 1993

4. FIN OR GRANT NUMBER

___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ ___ __ L2245

5. AUTHORIS) 6. TYPE OF REPORT

L. Winters Technical

7. PERIOD COVERED (inclusive Direst

liame and mailing *ddrniis

Minis try of Social Affairs

N\uclear Inspectorate (KFD)

NTetherlands Energy Research Foundation ECN

P. 0. Box 1, 1755 ZG Petten, The N~etherlands

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (fifNRC. type "Samei as above", ii coritractor. provide NRC Division. Otlie or Region. U.S. Nuclear Regularorv Comnmission.

and mailing address.)

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

It. ABSTRACT 1200 worys, or less)

As part of the ICAP (International Code Assessment and Applications Program) agreement between ECN

(Netherlands Energy Research Foundation) and USNRC, ECN has performed a number of assessment calculations

for the thermohydraulic system analysis code RELAP5/MOD2/36.05. This document describes the assessment of

this computer program versus a natural circulation experiment as conducted at the Borssele Nuclear Power Plant.

The results of this comparison show that the code RELAP5/MOD2 predicts well the natural circulation behaviour of

Nuclear Power Plant B orssele.

The work has been sponsored by the Dutch Licensing Authority and ECN.

12. KEY WOROS/OESCR:PTORS (List words orphrosss thate witttauist researchers In locating the rrport., 13. AVAILABILITY STATEMENT

ICAP Program Unlimited

RELAP5/MOD2 14. SECURITY CLASSIFICATION

Borssele I This Pagel

Unclassifieid

ilhz Rspoor"l Unclassified

15. NUMBER OF PAGES

16. PRICE

1NRC FORM 335 12-891


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NUREG/IA-009l ASSESSMENT OF RELAPS/MOD2 AGAINST A NATURAL CIRCULATION

EXPERIMENT IN NUCLEAR POWER PLANT BORSSELE

JULY 1993

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

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