NUREG/IA-0 108 -- 6 ICSP-V2-R 100-R

International

Agreement Report

Assessment of RELAP5/MOD2

Against a Turbine Trip From

100% Power in the Vandellos 11

Nuclear Power Plant

Prepared by

C. Llopis/A. N. V.

J. Perez, R. Mendizabal/C. S. N.

Asociacion Nuclear Vandellos (A. N. V.)

Consejo de Seguridad Nuclear (C. S. N.)

Madrid, Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

June 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-0 108

ICSP-V2-R 100-R

, International

Agreement Report

Assessment of RELAP5/MOD2

Against a Turbine Trip From

100% Power in the Vandellos 11

Nuclear Power Plant

Prepared by

C. Llopis/A. N. V.

J. Perez, R. MendlizabalfC. S. N.

Asociacion Nuclear Vandellos (A. N. V.)

Consejo de Seguridad Nuclear (C. S. N.)

Madrid, Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

June 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 documents work performed under the sponsorship of the Consejo De

Seguridad Nuclear of Spain. The information in this report has been provided

to the USNRC under the terms of an information exchange agreement between the

United States and Spain (Technical Exchange and Cooperation Agreement Between

the United States Nuclear Regulatory Commission and the Consejo De Seguridad

Nuclear of Spain in the field of reactor safety research and development,

November 1985). Spain 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

Spain 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, or 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.
LIST OF CONTENTS

PACE

LIST OF TABLES....................................... iv

LIST OF FIGURES............................... V

EXECUTIVE SUMMARY .................................... vii

ABSTRACT ............................................ ix

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

2. PLANT AND TRANSIENT DESCRIPTION................... 1

2.1 PLANT DESCRIPTION........................... 1

2.2 PLANT SIGNAL ACQUISITION SYSTEM

DESCRIPTION................................. 2

2.3 TRANSIENT DESCRIPTION........................ 3

3. CODE INPUT MODEL DESCRIPTION...................... 5

3.1 PRIMARY SYSTEM ............................. 5

3.2 SECONDARY SYSTEM........................... 7

3.3 CONTROL SYSTEMS ............................ 8

4. STEADY STATE CALCULATION ......................... 9

5. BASE CASE RESULTS ............................... 10

5.1 BOUNDARY CONDITIONS........................ 10

5.2 TRANSIENT RESULTS.......................... 11

6. ADDITIONAL STUDIES............................... 14

6.1 STUDY 1 FULL SCALE COMPONENTS MODEL ... 14

6.2 STUDY 2 PRESSURIZER WITH HOMOGENEOUS

MODEL ..................................... 15

7. RUN STATISTICS................................... 16

.8. CONCLUSIONS ..................................... 17

9. REFERENCES ...................................... 18

iii
L IST OF TABLES

PAG E

TAB3LE I :MA IN CHARACTER IST ICS OF VANOELLOS I I PLANT.....20

TAB3LE I I :STEADY STATE VALUES..............21

TABLE III :SEQUENCE OF EVENTS..............22

TABLE IV: RUN STATISTICS................23

.iv
LIST OF FIGURES.

PAG E

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

1

2

3

4

5

6

7

8

9

RELAP5 NOD ING D IAGRAM F OR VAND ELL6s I I NPP ..

BASE CASE :REACTOR POWER............

BASE CASE : SECONDARY PRESSURE.........

BASE CASE :SG. REC IRCULAT ION RATIO......

BASE CASE :PRIMARY TEMPERATURES.......

BASE CASE PRIMARY AVERAGE TEMPERATURE...

BASE CASE FEEDWATER MASS FLOW........

BASE CASE SG. LEVEL..............

BASE CASE :PRESSURIZER PRESSURE.......

24

25

26

27

28

29

30

31

32

FIGURE 10 : BASE CASE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

11 :

12 :

BASE

BASE

CASE

CASE

:PRESSUR IZER LEVEL..........33

:PRESSURIZER VAPOUR GENERATION .... 34

:PHASic RELATIVE VELOCITY IN

PRESSURIZER....................3

13

14

15

FIGURE 16:

FIGURE

FIGURE

17:

18:

1ST. ADDITIONAL STUDY

1ST. ADDITIONAL STUDY

1ST. ADDITIONAL STUDY

AVERAGE TEMPERATURE.

1ST. ADDITIONAL STUDY

PRESSURE........

1ST. ADDITIONAL STUDY

2ND. ADDITIONAL STUDY

VAPOUR GENERATION..

2ND. ADDITIONAL STUDY

PRESSURE..............

BASE CASE : CPU TIME

BASE CASE :TIME STEP

:NOD ING DIAGRAM ....... 36

:SECONDARY PRESSURE ... 37

:PRIMARY

.~

PRESSURIZER

.~

PRESSURIZER LEVEL .... 40

PRESSURIZER

.~

PRESSURIZER

.~

.~

.~

FIGURE 19:

FIGURE

FIGURE

20 :

21 :

V

EXECUTIVE SUMMARY

An assessment of RELAP5/MOD2 cycle 36. 04 against a

turbine trip from 100 X power in the Vandell6s II nuclear power

plant is presented. The work Is inscribed In the framework of the

spanish contribution to ICAP Project.

Vandell6s II Is a plant owned by ENDESA (72 %) and

HIDROELECTRICA ESPAROL.A (28 %) located In Tarragona (Spain).

The transient under study was part of the

preoperational test program and a large number of plant signals

were recorded by the Signal Acquisition System.

The model used consisted of a single loop, a steam

generator and a steam line up to the steam header all of them

enlarged on a scale of 3:1, and full-scaled reactor vessel and

pressurizer.

The analysis followed the usual steps: modelling

of the plant; calculation of the plant steady state previous to

the test; calculation of the transient; comparison with plant

measurements; and performance of sensitivity studies.

Calculations were carried out using Cycle 36.04 of

RELAPS/MOD2 code installed in the CDC CYBER 830 computer owned by

the CSN.

The steam dump demand signals recorded in plant

showed a perturbation, inconsistent with the average temperature

behaviour. This was attributed to a malfunction in the Signal

Acquisition System. The actual response of the steam dump banks

under demand was uncertain.

Vii
The results of the calculations have been in

reasonable agreement with plant measurements. An additional study

has been performed, to check the ability of a model In which all

the plant components are full-scaled to reproduce the transient.

A second study has been performed, using the Homogeneous

Equilibrium Model in the pressurizer, trying to elucidate the

Influence of the phasic velocity slip In the primary

depressurization rate.

Viii
ABSTRACT.

An assessment of RELAP5/MOD2 cycle 36.04 against a

turbine trip from 100% power In Vandell6s II NPP (Spain) is

presented. The work is Inscribed In the framework of the spanish

contribution to ICAP Project.

The model used In the simulation consists of a

single loop, a steam generator and a steam line up to the steam

header all of them enlarged on a scale of 3:1; and full-scaled

reactor vessel and pressurizer.

The results of the calculations have been In

reasonable agreement with plant measurements. An additional study

has been performed, to check the ability of a model in which all

the plant components are full-scaled to reproduce the transient.

A second study has been performed using the Homogeneous

Equilibrium Model In the pressurizer, trying to elucidate the

Influence of the velocity slip in the primary depressurization

rate.

ix

FOREWORD

This report r-epresents one of the assessment/application

calculations submitted in f~ulfilment of the bilateralagreement for cooperation in thermalhydraulic activities

between the Consejo de Seguridad Nuclear of Spain (CSN) and

the United States Nuclear Regulatoy Commission (US-NRC) in -

the form of Spanish contribution to the International Code

Assessment and Applications Program (ICAP) of the US-NRC whose

main purpose is the validation of the TRAC and RELAP system

codes.

The Consejo de Seguridad Nuclear has promoted a coordinated -

Spatish Nuclear Industry *effort (ICAP-SPAIN) aiming to -

satisfy the requirements *of this agreement and to improve the

quality of the technical support groups at the Spanish -

Utilities, Spanish Research Establishments, Regulatory Staff

and Engineering Companies, for safety purposes.

This ICAP-SPAIN national program includes agreements between

CSN and each of the following organ~izations:

- EMPRESARIOS, AGRUPADOS, S.A.

- Unidad Elictrica (UNESA)

- Un16n Iberoamericana de Tecnologia Elictrica (UITESA)

- Empresa Nacional. del Uranio (ENUSA)

- TECNATOM

- LOFT-ESPANA

The program is executea oy 13 working groups and a generic code

review group and is coordinated by the "Comiti, de Cooz-dinaci6n".

This committee has approved the distribution of this document -

for ICAP purposes.

xi
1-
1. INTRODUCTION.

The results of an assessment of the REL.AP5/MODZ

code against a turbine trip are presented in this report. This

work Is inscribed In the Spanish contribution to the

International Code Assessment and Applications Program (ICAP).

Its main additional objective is to promote the elaboration of a

Vandell6s II plant model with RELAP5/MOD2 code.

The transient under study was one of the

preoperational tests of the Vandellds II nuclear power plant. A

Signal Acquisition System recorded a large number of plant

signals.

The analysis followed the usual steps: modelling

of the plant; calculation of the plant steady state previous to

the test; calculation of the transient; comparison with the plant

measurements; and performance of sensitivity studies.

Calculations were carried out using Cycle 36.04

of RELAP5/MOD2 code installed In the CDC CYBER, 830 computer owned

by the CSN

This same turbine trip test has been analyzed

using the TRAC-PF1/MQDl code by UITESA, in the framework of the

Spanish contribution to ICAP (8].

2. PLANT AND TRANSIENT DESCRIPTION

2.1 PLANT DESCRIPTION

Vandell6s II Is a three-loop Westinghouse PWR

nuclear power plant owned by ENDESA (72%) and HIDROELECThICA

1
ESPANOLA (28%). It Is located In Tarragona, in the North-East of

Spain, and uses the Mediterranean Sea as the final heat sink. The

plant started Its commercial operation In 1988. The nominal

power is 982 MWe (2775 MWt).

The reactor vessel is cold head type . The plant

Is equipped with three Westinghouse U-tube steam generators

(model F) without preheaters. The feedwater is fed directly to

the upper part of the downcomer via J-tubes. The circulation

ratio on the secondary side of the steam generators Is 3.27 at

rated power.

The Auxiliary Feedwater System consists of one

turbopump and two motorpumps.

In the plant there are, among others, control

systems for the reactivity (rods and boron), primary pressure,

pressurizer level, steam dump and steam generator level. The

Reactor Protection System Includes safety valves In the

pressurizer and the steam generator.

The main plant features are shown in Table I.

2.2 PLANT SIGNAL ACQUISITION SYSTEH DESCRIPTION

To record the main parameters of the plant, during

the startup period (including the transient under study), a

temporary Signal Acquisition System was installed. It consisted

of a digital system with an up to 0.05 seconds and 144 signals

trail capacity.

The recorded parameters depended on the test

carried out.

The quickness of data attainment was very

Important to Improve the time required for data interpretation.

2
For this reason, once the nuclear plant tests had finished,

Vandell6s II NPP decided to Install a permanent equipment in

order to interpret and analyze the transients.

The availability of this great number of signals

allows to check the partial performances of the control blocks,

specially those of feedwater control, rod control and steam dump.

2.3 TRANSIENT DESCRIPTION.

The transient under study is a startup turbine

t~h trip from 100%! power. It was conducted on February 27 , 1988.

Objectives of this test were to verify the ability

of the plant to accept a total load rejection, reaching stable

conditions; and to make some evaluations (response times of

RTD's, changes in control systems setpoints... ).

Previously to the test, the plant was in stable

regime, at 100% power. All control systems were correctly

performing in automatic mode.

The transient started with a manual turbine trip.

The trip signal produced the closure of the turbine control

valves, and the reactor scram. After this, the hot leg

temperature decreased, and so did pressurizer pressure and level.

The spare heaters activated when the corresponding setpoints were

reached.

The turbine trip produced a quick secondary

pressure Increase. This fact deteriorated the primary-to

-secondary side heat transfer in the steam generator, and had as

a consequence a slight Increase in the cold leg temperature

during the early seconds of the transient.

As a result of the turbine trip, the reference

3
temperature suddenly changed from full load to zero load, and

there was a significant temperature error which produced the

quick opening of the steam dump valves. The heat removed through

these valves reduced the primary average temperature below the

reference value.

The primary-to-secondary heat transfer decrease

(and, In a lower scale, the secondary pressurization) originated

a void collapse in the steam generator, resulting in a quick fall

of the downcomer liquid level. The low-low level signal was

reached, and the Auxiliary Feedwater System was activated. The

main feedwater pumps were tripped at low average temperature

signal in coincidence with reactor trip signal.

During the transient 144 plant signals were

monitored by means of a Signal Acquisition System, with a

frequency of 0.05 seconds, and stored In a computer.

The demand signals of steam dump banks 1 and 2 had

an Irregularity at 10.5 seconds, detaching for few seconds from

the evolution of the compensated average temperature. At this

time, the signals suddenly fell to zero, and rose again In 0.2

seconds. Until around 30 seconds, the signals did not follow the

average temperature evolution.

As stated in (8], the origin of these abnormal

signals could be:

a) A malfunction of the Signal Acquisition System or

b) A malfunction of the control block transmitting to

the bank positioners, which modified the closure sequence

required by the average temperature.

In addition to this, the actual response (dead

times and movement velocity) of the steam dump valves was not

certain, and the valve positions were not recorded.

4
3. CODE INPUT MODEL DESCRIPTION.

The plant model (Fig. 1)' consists of a single loo p,

a steam generator, and a steam line up to the steam header, all

of them enlarged on a scale of 3:1; and full-scaled reactor

vessel and pressurizer. It derives from the 1:1 nodalization of

each Individual component, separately elaborated and tested. The

scaling was done by triplicating the values of flow cross

sections and heat transmission areas; pump torque, flow and

Inertia were also multiplied by 3. Such a model Is app ropriate to

the transient under study, which is basically symmetric. The

nodalization includes 116 hydrodynamic volumes, 121 junctions and

78 heat structures, with 316 mesh points.

The boundaries of the model are feedwater

collector, turbine and CVCS tank, simulated by means of RELAP

Time Dependent Volumes (TMDPVOL).

Point kinetics is used to simulate the source of

power. So, the plant model will be unable to reproduce the axial

power distribution change that takes place as the control rods

are going up or down through the core arnd the effect that thid

change produces in reactivity coefficients.

This plant model was based on a RETRAN two-loop

model [3], and incorporated additional plant data. The

corresponding nodalization studies are detailed in [6].

3.1 PRIMARY SYSTEM

It includes the reactor vessel, loops, steam

generator pri mary side, pumps and pressurizer.

The loop is scaled-up 3:1, excluding the vessel

5
and pressurizer, which are full-scaled. Each component of the

model has been separately tested.

The reactor vessel Is cold head type. The dome has

been separated In three nodes, representing the upper zone, the

Inner circular one and the surrounding annulus. respectively. The

upper plenum consists of two volumes, to ensure the proper

connection of the outlet junction. The lower plenum has been also

split In two nodes: one previous to the active core and the other

one representing the hemispheric zone.

The reactor core has been simulated with six

control volumes and a heat structure with six axial nodes. Use' of

the point kinetics model of the code has been done, with a null

moderator temperature coefficient (because the test under study

was done at beginning of life). The scram reactivity has been

Input through a table in function of time.

The core bypass path Is divided in six nodes. Both

the core bypass and bypass-to-head flow rates have been tuned

through the energy loss coefficients.

Cylindric heat structures represent the heat

losses through the vessel walls.

The pressurizer nodalization Includes ten

hydrodynamic volumes. The surge line Is split In two PIPE

components, accounting for the horizontal and vertical zone,

respectively. Heat structures are used to represent the heaters

and heat losses to the environment, trying to obtain a realistic

temperature distribution. Relief and safety valves have also been

simulated.

Homologous curves for the primary coolant pumps

performance have been obtained through characteristic curves.

Only data for normal operation conditions were included In the

Input deck. The moment of Inertia, and rated flow, torque and

6
motor torque have been triplicated.

The primary side of the steam generator has been

split In 12 nodes, two of which represent the inlet and outlet

chambers. The U-tubes have 10 nodes, with Increasing length In

the flow direction, in order to reproduce in detail the

temperature profile and enhance the primary-to-secondary heat

transfer.

3.2 SECONDARY SYSTEM

The three steam generators have been unified, and

so have been the steam lines up to the collector. Mean values

have been assumed in the pipe simulation, because they are not

exactly equal In the plant.

The steam generator has been modelled In a great

detail [6]. Heat losses to the environment are represented by

RELAP heat structures. It is interesting to point the existence

of a heat structure which connects the boiler volumes and those

of the downcomer, representing the wrapper.

The moisture separators zone has been modelled by

means of an "ideal" SEPARATR component.

Relief valves are simulated by VALVE components;

and safety valves, by Time Dependent Junctions (TMDPJUN). No one

was activated during the transient under study.

Downstream the collector, the four turbine

admission valves are assimilated to one VALVE. Four VALVE

components represent the four banks in which gather the 12 steam

dump valves, and account for the modulate behaviour of this

system. Its capacity is adjusted to =30%. of the full power steam

mass flow at nominal pressure. A Time Dependent Junction accounts

7
for the steam extraction towards the MSR, ejectors, turbopumps,

etc...

3.3 CONTROL SYSTEMS

The following control systems have been Included

In the plant model:

- Control rods.

- Pressurizer level control.

- P ressurizer pressure control.

- Steam dump control.

- Steam generator level control.

The five groups have been simulated according to

the plant design [61. The plant actual control setting values

during the test have been used as setpoints.

The CVCS charge was simulated by means of a VALVE

and a THDPVOL. The discharge. was-, represented byý a iTMDPJUW

extracting a continuous mass flow of 2.6 Kg/s from the primary

system. Such a model Is judged right for the purposes of this

analysis.

The steam generator level control system did not

include the speed control of the turbine driven pumps, which were

not modelled.

The steam mass flow has been used as a measure of

the turbine power. It Is more closely related to the

Impulse chamber pressure than the valve position.

8
4. STEADY STATE CALCULATION

Before the test simulation. a null transient was

run to establish the initial conditions.

The STDY-ST code option was used. To adjust the

100% power steady state, use was made of the data measured in the

plant previously to the test, and showed in Table II. Other data

that were used are

- Design values of the core bypass mass flow rates

- Standard pressure losses In a PWR-W vessel and

loops [3].

- Design steam generator recirculation ratio.

- Design heat losses to the environment.

In this job was very useful the achievement of

steady states for Isolated components, such as reactor vessel,

.steam generator..and -pressurizer.....

The energy loss coefficients in the junctions were

assigned Handbook values [51, and then tuned to adjust pressure

losses or bypass f lows. For instance, the core bypass mass f lows

were adjusted by properly tuning the energy loss coefficients In

the reactor vessel.

To adjust the steady state use was made of the

real plant control systems. In addition, a dummy control system

was added to adjust the primary mass flow rate by tuning the pump

speed.

Known shortcomings in the RELAP5/MOD2 heat

transfer correlations (2) forced to Increase the

primary-to-secondary heat transfer area in about 10 X to achieve

the desired steady state.

9
Table II shows the comparison between the steady

state values calculated by the code and those measured In plant.

Signed with an asterisk are the parameters used to define the

steady state (91]; they were thus controlled or imposed In the

calculation. The agreement Is good. Nevertheless, It Is Important

to point that the calculated steam generator water mass is 30%

lower than the reference full power value.

5. B3ASE CASE RESULTS

5.1 BOUNDARY CONDITIONS

The simulation started with the turbine valve

closure In 0.5 seconds, according to the data recorded in the

plant. The closure rate was supposed constant.

A Time Dependent Junction was kept extracting a

mass flow of 37.06 Kg/s from the steam collector volume, trying

to represent the MSR's effect during the transient.

The reactor scram was supposed at 0. seconds. At

this time, the turbine valve started to close. The total duration

of the rod Insertion was 1.66 seconds (including an initial dead

time of 0. 16 seconds). The total Inserted reactivity was 7208 pcm

(10.22556 S). The decay heat was that calculated by the code

according to a 1000. seconds at full power history.

Header measured temperature and pressure were

Imposed as boundary conditions In the TMDPVOL representing the

main feedwater source.

Auxiliary feedwater mass flow and temperature were

not recorded in plant. Design values of 95 Kg/s and 300 K,

10
respectively, were taken, assuming the performance of two

motorpumps and the turbopump. The auxiliary feedwater' started by

low-low steam generator level signal. In the simulation, this

system was activated 18.6 seconds after the turbine trip,

according to plant data.

The abnormal demand signals for steam dump banks 1

and 2 were considered a malfunction of the Signal Acquisition

System and given no credit. Nevertheless, a run performed

Imposing these demand signals as boundary conditions showed small

Influence on the general transient evolution. This fact is also

stated as one of the conclusions in the TRAC analysis by UITESA

(8J. The response of these valves was supposed as follows:

- After the "trip open" signal, 3.9 seconds to

open.

- After modulation signal, 5 seconds to fully open

(or to fully close).

The steam dump capacity was adjusted to 30% of the

full power steam mass flow at nominal pressure.

According to recorded data, there was a partial

opening demand for one of the three relief valves during a few

seconds. The secondary pressure In that loop was slightly higher

than in the other two, and the pressure setpoint was reached. In

the simulation this relief was not Included, because the steam

mass so released was negligible In comparison with steam dump

flow.

5.2 TRANSIENT RESULTS

The simulation was initiated from the already

described steady state. The calculated sequence of events Is

compared with the measured one in Table III.

The steam dump demand signals were given no credit

11
from 10.5 seconds on. So, some signals on the Table III have been

derived from the recorded average temperature.

The plant data which appear in the figures are

mean values of the three loops. No data uncertainty was

available. Some calculation results have been filtered. The hot

and cold leg temperatures are filtered by means of a 4 seconds

LAG to evaluate the average temperature recorded by the control

systems. The steam generator-level, feedwater mass flow and steam

mass flow, are lagged 0.25 seconds.

The turbine valve began to close at 0. seconds,

and reactor scrammed. Steam line pressure rose, and steam flow

decreased, until steam dump valves opened.

Fig. 2 compares the calculated reactor power with

the measured neutronic: flux.

The vapour generation in the boiler decreased

following the primary-to-secondary power.'

The global effect was that secondary side pressure

had a maximum at a 6 seconds (Fig. 3). This point depends on the

opening velocity and dead times of the steam dump valves.

Velocities and delays assumed In the calculation are mean values

derived from the measures (in trip mode) taken for each valve

during the preoperational tests program.

After the scram, the heat removed through the

steam dump valves quickly reduced the average temperature. Few

seconds later, the valves were demanded to start closing.

In the lapse between 20 and 70 seconds the

secondary pressure was underestimated. This may be attributed to

an excessive steam release. The pressure took a minimum short

time before the loss of recirculation In the steam generator

(about 26 seconds) (Fig. 4).

12
Fig. 5 shows the filtered hot and cold leg

temperatures, compared with the measured ones. Both are

underestimated from 20 seconds on. The reason may be the

overprediction of the discharge through the steam dump valves.

The average temperature is compared with the plant data in Fig. 6

The calculated feed mass flow coincides fairly

well with the measured one (Fig. 7). It decreased in the early

seconds, due to the mismatch with the steam mass flow. However,

the steam generator liquid level rapidly dropped to zero, and the

valve opened again. When the narrow range level became lower than

zero (Fig. 8), the mass flow remained stable until the valve

tripped. That did not happen in the calculation, where the level

fell faster and the mass flow became larger than the measured

one. This may be attributed to the mentioned mass default In the

steam generator. The trip time was well predicted, due to the

good agreement between the average temperatures.

The auxiliary feedwater flow started, as imposed,

18.6 seconds after the turbine trip.

The pressurizer pressure and level are In a

reasonable agreement with the plant data (Figs. 9 and 10). The

level Is overpredicted In about 5% . During the initial decrease

In the level the calculated primary depressurization rate was

slightly lower than the observed one, In coincidence with strong

flashing in the upper liquid zone of the pressurizer (Fig. 11).

This vapour generation was partly due to a large relative

velocity between both phases (Fig. 12). A sensitivity study has

been performed using the Homogeneous Equilibrium Model In the

pressurizer, to elucidate the influence of the velocity slip in

the depressurization rate (see section 6).

13
6. ADDITIONAL STUDIES.

Some additional studies were performed in this

analysis. First, a model was used consisting of full-scaled

components, with the exception of the hot leg, enlarged on a

scale of 3:1. Second, a run was performed using the 3:1 scheme

and applying the Homogeneous Equilibrium Model (HEM) to the

pressurizer junctions.

6.1 STUDY 1 FULL SCALE COMPONENTS MODEL

A study was performed to verify the Influence of

the spatial scaling on the analysis. A model was employed In

which only the hot leg was enlarged on a scale of 3: 1. The steam

generator, steam line, pressurizer, reactor vessel, pump and cold

leg were kept full-scaled. Two TMDPJUN were used to triplicate

the primary flow entering and leaving the 1:1 cold leg. FIg.13

shows the noding diagram.

This model was built In a straightforward way,

simply by connecting the separated 1:1 component models

previously tested.

Results obtained In this study are showed in Figs.

14 to 17. Until 30 seconds, they are reasonably close to those of

Base Case and plant. Afterwards, an excessive primary system

cooling Is appreciated. The consequence Is a steady decrease in

the pressurizer pressure and level. The reason of this trend Is

that in RELAPS It Is not possible to reproduce exactly the

thermodynamic state of a volume through a TMDPVOL component. So,

the TMDPJUN 184 was Injecting water in the vessel Inlet with an

enthalpy lower than that of the simulated cold leg.

14
6.2 6.2 STUDY 2 :PRESSURIZER WITH HOMOGENEOUS

MODEL.

As mentioned In 5. 2, in the Base Case a primary

depressurization rate slightly lower than the measured one was

observed. This was attributed to the strong flashing found in the

upper part of the pressurizer. Such a vapour generation stems

mainly from the superheated liquid in that upper zone.

RELAP5/MOD2 calculates the bulk vapour generation

through some Interfacial heat transfer coefficients, which are

flow regime dependent [2J. These coefficients are proportional to

the Interfacial area per unit volume, and this area is, for

bubbly flow, proportional to the void fraction and to the square

of the phasic relative velocity. Fig. 12 shows the relative

velocity between the phases In the upper liquid zone of the

pressurizer, and Indicates that the volumetric interfacial area

took a maximum in the early seconds of the transient.

To check the influence of the phasic relative

velocity on the depressurization rate, a run was performed using

the 3:1 scheme and applying the Homogeneous Equilibrium Model

(HEMi) to the pressurizer junctions (by setting h = 2 In the

junction flags). This action should eliminate the flashing due to

the velocity slip effect.

The results of this case were not better than

those of Base Case. Fig. 18 compares the vapour generation in

pressurizer obtained with this model with the Base Case result.

The main effect of the HEM use was the suppression of a peak at

about 5 seconds, due to velocity slip. Nevertheless, the pressure

slope (Fig. 19) basically did not im~prove with this change.

i5
7. RUN STATISTICS.

The calculations were run on a CDC CYBER 830,

owned by the CSN. The operating system was NOS 2.7 . The code

cycle used was 36.04.

Table IV shows the run statistics for the

steady-state run, the Base Case and the additional studies. In

transient runs, It was specified a maximum time step of 0.05

seconds, lower than the Courant limit (about 0.06 seconds

throughout the transient). So the code always used this maximum

value. In the steady state run, the code reduced the time step to

0.025 seconds. The reason was an Incorrect initialization in the

nodes representing the steam dump discharge pipe, which reduced

the Courant limit under 0.05 s.

The CPU time to transient time ratio has been

around 42 (91. for the steady state run). The grind time was

among 17.5 and 18. miliseconds

The CPU time and time step are plotted versus

transient time in Fig. 20 and 21, respectively, for Base Case.

16
8. CONCLUSIONS.

Three calculations have been performed, and the

results are in reasonable agreement with the plant measurements.

The steam dump demand uncertainty has not affected

significantly the results.

The first additional study reveals that a model

consisting of full-scaled components ,one loop and Time

Dependent Junctions that triplicate the primary flow,

reproduces the 3:1 model trends. But this model is unrealistic

because the code cannot replicate exactly the thermodynamic state

of a volume through a TMDPVOL component.

The second additional study shows that the use in

this analysis of Homogeneous Equilibrium Model In pressurizer

does not alter substantially the primary pressure evolution.

17
9. REFERENCES.

1.- "RELAP5/MOD2 Code Manual. Vol. 1 and 2. V. H.

Ransom et al. NUREG/CR-4312. EGG-2396. Rev. 1.

March 1987.

2.- "RELAP5/MOD2 Models and Correlations". R.A.

Dimenna et al. NUREG/CR-5194. EGG-2531. August

1988.

3.- "Simulaci6n de Transitorios de Central PWR

Mediante el C6digo RETRAN-02. Cuaderno de

C.Alculo". UITESA, Julio 1987.

4.- "Informe Final de Seguridad, Rev.1 ".C.N.

Vandell6s-II, Jun10 de 1985.

5.- "Handbook of Hydraulic Resistance ".I.E.

Ideichik. Hemisphere Publishing Corporation.

Second Edition. 1986.

6.- "Nota de C~hculo de la Modelaci6n de C.N.

Vandell6s II mediante RELAPS/MOD2 ". A. Casals et

al. Marzo de 1991.

7.- "Documentac16n de la Copia Oficial de Prueba.

Procedimiento PAN-63, Rev. 1. Disparo de Planta

desde el 100% ". C.N. Vandell6s-II, 29-02-88.

8- "Assessment of TRAC-PF1/MOD1 against a turbine

trip from 100%. power in the Vandell6s II Nuclear

Power Plant". Prepared for ICAP-Spain. A. Querol,

R. de la Fuente, P. Hern~n. UITESA. Ref. Consejo

de Seguridad Nuclear. ICSP-V2R100-T. October 1990.

i8
"Capacidades y experiencia adquiridas en el

juiclo de andlisis realizados con los c6dlgos

termohidrdulicos RB..AP5/MOD2 y TRAC-PF1/MOD12"

J.M. Izquierdo et al. XVI Reun16n Anual SNE.

Oviedo, Octubre de 1990.

19
T A BLE I

MAIN CHARACTERISTICS OF VANDELLOS II PUANT.

Thermal Reactor Power (MWt) ...................... 2775.

Electrical Power (MWe) ........................... 992.

Fuel .............................................U02

Number of assemblies ............................. 157

Number of coolant loops ............................ 3

Cladding Tube Material .................... ZIRCALOY 4

Absorber Material ..................... B4C + Ag-In-Cd

Reactor Operating Pressure (MPa) ................. 15.4

Coolant Average Temperature

Zero Load (K) ........................... 564.8

100% Load (K)...........................582.3

Steam Generator ...................WESTINGHOUSE TYPE F

Number of tubes in SG ............................5626

Total Tube Length (in).................................. 98759.

Inner Diameter Tubes (mn).............................. 0.0156

Tube Material .................................INCONEL

Pumps Type .........................WESTINGHOUSE D 100

Discharge Head of Pumps (bar) .................... 18.8

Design Flow Rate (m3/s) .........................6.156

Speed of Pumps (rad/s) ........................... 155.

Primary Volume (mn3) ............................106.19

Pressurizer Volume (mn3) .........................39.65

Heating Power of the Heaters Rods (KW) .......... 1400.

Maximum Spray Flow (Kg/s) ........................ 44.2

Steam Mass Flow Rate at 100% (Kg/s) ............. 1515.

20
TA B LE 11

STEADY STATE VALUES

PARAMETER MEASURED CALCULATED

PRIMARY SIDE

Core Power M*. 99.1 99.8 C)

Mass Flow Rate (Kg/s) --- 14602.

RCP Speed (Rad/s) --- 158.2

RCP Head (MPa) --- 0.645

Hot Leg Temperature (K) 597.3 596.8

Cold Leg Temperature (K) 564.1 563.7

Average Temperature (K) 580.7 580.2

Delta T M% 99.4 99.2

Pressurizer Pressure (MPa) 15.41 15.33 ()

Pressurizer Level M% 57.2 56.7 C)

SECONDARY SIDE

SG Dome Pressure (MPa) --- 6.69

SG Outlet Pressure CHPa) 6.5 6.59

Collector Pressure (MPa) 6.35 6.56

Feedwater Mass Flow (Kg/s) 1542.9 1513.2

Steam Mass Flow (Kg/s) 1471.8 1514.2 ()

Feedwater Temperature (K) 494.1 493.3 C)

SG Level (Me 50.5 50. C)

Recirculation Ratio --- 2.29

* Average values.

() Controlled or imposed parameters.

21
T AB L E

SEQUENCE OF EVENTS

EVENT TIME (SECONDS)

PLANT RELAPS/MOD2

BASE CASE

TURB3INE TRIP 0.0 0.0

REACTOR TRIP --- 0.0

STEAM DUMP DEMAND SIGNALS:

TO TRIP OPEN (4 BANKS) 0.1 0.0

TO START CLOSING BANK 4 3.5 3.5

TO START CLOSING BANK 3 5.8 5.0

TO START CLOSING BANK 2 8.4 8.0

TO START CLOSING BANK 1 11.8 ()12.5

LOW-LOW LEVEL IN SG. 7.0 8.4

LOW AVERAGE TEMPERATURE 18.5 18.2

MAIN FEEDWATER TRIP 19.0 18.2

AUXILIARY FEEDWATER INJECTION -- 18.6

STEAM DUMP VALVES FULLY CLOSED -- 38.5

()According to average temperature program.

22
T A BLE IV

RUN STATISTICS

CALCULATION TT (S) CPU (S) TS (S) CPU / TT CN TSN GT (inS)

Steady State 91.2 8301.8 0.025 91.0 116 4070 17.58

Base Case 100. 4190.1 0.05 41.9 116 2013 17.94

Additional Study 1 100. 4217.3 0.05 42.2 119 2023 17.52

Additional Study 2 100. 4174.2 0.05 41.7 116 2011 17.89

KEY:

TT : Transient Time

CPU : CPU Time

TS : Maximum Time Step

CN : Cells Number

TSN : Time Steps N~umber

GT : Grind Time ( =CPU/(CN x TSN)
-it-opfiv~ lbov ) fop~e v strowlAtyv *AW-H WP

UA " 44S FP 613 46 /N lw

430 QS 25

423 RELIEF T ArIfr a1i30 -4~~ 67

VALVES I~ JVAvCS '- -

C..VANDELLOS HE ujYjfi1E

P*(SSW)Z[at 141SC-7 0 S

413 171 14131-T~J4 CLWODOSLA

411 43 W-I

- 14131.4

I4) I4s- e '

IIAW I I Si -5 1214 2 - E

50 AM"3 0

s1CLilIo 

-

12101 

titT

__ 

AAI4 RM ~~~~6 4 _ i

FIGURE 1 .RELAP5 NODING; DIAGRAM FOR VANDELLO6S 11 NPP
Iea. 88 v

88.8as

N

Ln.

U.

LdJ

Lii

88.08

40.08

28.88

4 I I I I aiPLANT NEUTROHIC FLUX CX)

A CHTRLVAR 381

A- 8.8 Ie

. 0

- -. - r' w m in in -

2e 48 Be so 188

TIME (SECONDS)

FIGURE 2 . BASE CASE : REACTOR POWER.
6.80

P Y AP SAL GV - PLANT AP 60061

N' 0%

0i

c,,

c,)

-cc

LU

C41)

6.66

a 20 46 Be Be 160

TIME (SECONDS)

FIGURE 3 .BASE CASE :SECONDARY PRESSURE.
5.08

M CNRLVAR 117

4.00

(%J

C3)

ud

s.00

2.06

I .AD

9 26 40 68 so 160

TIME CSECONDS)

FIGURE 4 . BASE CASE : SG. RECIRCULATION RATIO.
see. 0 C3 T RAN CA - REC PLANT

A T RAN FR RE - PLANT

X CNTRLVAR 327

oD CNTRLVAR 326

rict0

0.'

I-c

ILu

ILo

C-3

-d

I.-

03

575.98

sse.99

0 20 46 so so 10e

TIME (SECONDS)

FIGURE 5 .BASE CASE : PRIMARY TEMPERATURES.
60e. o V3T2 MEDIA A CTL PLANT ACNTRLYAR 330

S 575.08

N a

zj

sLUe 96 

s S - 0 

TIE(ECNS

FIUE6 1AECAE--RMR VEAEEPRTR
2488.e88

C3F AG ALrm GV - PLANT

,& CNTRLYAR 201

1 8803.08

w

I-i

U-

(e)

ff)

Li..

1298.88

680.08

8.98

a 48 60 Be too

TIME (SECONDS)

FIGURE 7 . BASE CASE :FEEDWATER MASS FLOW
68.88

C3L RE GV PROT - PLANT

,& Ct4TRLVAR 283

48.88

c~) I-.

eN

Sts

-J

0i

28.88

8 20 49 Be as fee

TIME CSECONDS)

FIGURE 8 .BASE CASE : SG. LEVEL.
16.00

M P PRZ CONTROL P@483

A P 41SO9

Its~e

C.,

t%

4-'

co

(n)

hi

0r

mC

fr

M.

14.00

13. 90

a 28 48 68 88 too

TIME (SECONDS)

FIGURE 9 .BASE CASE : PRESSURIZER PRESSURE.
100.90

L NIVEL PRZ - PLANT

SCNTRLVAR 350

80.90

80.90

e%

-j

0: 40.98

20.90e

8.89

a 20 49 es Be too

TIME (SECONDS)

FIGURE 10 . BASE CASE : PRESSURIZER LEVEL.
Wa.s

0 CHrRLvAR 733

40.9

-C

cc

-1.9 1

29496 s

TIM CECODS

FIUE1 AECS RSUIE AORGNRTO
9.58e

I II mVREL 41S85

AVREL 41586

X VREL 41607

8.48

8. 38

8.29

(119

9.-e

5 1 i 2

TIE(ECNS

FIGURE 12 .BASE CASE : PHASIC RELATIVE VELOCITY IN PRESSURIZER.
TUO~JUM CThORfSIE - 156 S~

(A)

0%

FIGURE 13 .1ST. ADDITIONAL STUDY :NODING DIAGRAM
wP VAP SAL GV - PLANT

B ASE CASE

STHlL MIODEL

CL

aI)

7.08

6.88

0 28 48 68 Be 100

TIME (SECONDS)

FIGURE 14 . 1ST. ADDITIONAL STUDY :SECONDARY PRESSURE.
P PRZ CONTROL P8483

ABASE CASE

X THL MODEL

ls.00

aw

(I)

-c

a.

14.08

13.80e

13 28 48 68 so tee

TIM1E (SECONDS)

FIGURE 15 . 1ST. ADDITIONAL STUDY : PRIMARY AVERAGE TEMPERATURE
Bee. as

MT2 MEDIA A CIL PLANT

A, BASE CASE

X THL MODEL

C')

'0

Lai

-C

uii

I.-

cm

575. .8e

558.88

a 20 48 68 8e 100

TIME CSECONDS)

FIGURE 16 .1ST. ADDITIONAL STUDY : PRESSURIZER PRESSURE.
Ie s.see

x

L NIVEL PRZ - PLANT

BASE CASE

THL MODEL

89.89

608.9

0

U)

-J

(L. 48.99)

293 99

9.99

a 29 49 69 so lee

TIME (SECONDS)

FIGURE 17 . 1ST. ADDITIONAL STUDY :PRESSURIZER LEVEL
so.80

inCNTRLVAR-733

SCNTRLVAR 733 - HEM

49.99

39.98

I-.

Ct,

N.

C,

0

I-'

Iw

z

w

CD

0

0~

29.900

19.90

9.98

-19.91

a 20 49 69 8s ISO

TIME (SECONDS)

Forti tpr 1R ?Nn AnDiTinNAI STUD~Y : PRESSURIZER VAPOUR GENERATION
Wee8

C3P PRZ CONTROL P8483

A~ P 41S09

X P 41S69 - HEN

N

C.I

CL

0:

of

CL

13.08

a 20 46 69 so fee

TIME (SECONDS)

FIGURE 19 . 2ND. ADDITIONAL STUDY : PRESSURIZER PRESSURE.
s888.98

MCPUTItIE 8

4888.88

3888.88

U,

LAJ

c-.1

2888.88

1888.98

8.88

a 28 48 60 88 lee

TIME CSECONDS)

FIGURE 20 .BASE CASE :CPU TIME.
9. tO

[3 CNTRLVAR 7

9.98

I--

cLU

9.9s

e.03

9.98

9 29 49 6o as too

TIME (SECONDS)

FIGURE 21 .BASE CASE: TIME STEP
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89) (Ass~gnd tby NRC. Add Vol., Sumss. Rev..

NRCM 1102, and Addendum Numbers. It wny.) 3201.3202 BIBLIOGRAPHIC DT HE

(See instructions on the reverse) UE/A00

2. TITLE AND SUBTITLE IICSP-V2-R1 00-R

Assessment of RELAP5/MOD2 Against a Turbine Trip From 100% Power

in the Vandellos II Nuclear Power Plant 3. DATE REPORT PUBLISHED

MONTH Y EAR

aune 1 1993 4. FIN OR GRANT NUMBER

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

5. AUTHOR(S) 6. TYPE OF REPORT

C. Ll opi s /A. N. V.

J. Perez, .R. Mendizabal/C.S.N. ___________

7. PERIOD COVERED (inclusive Dares)

n amend mailing address.)

Asociacion Nuclear Vandellos Consejo de Seguridad Nuclear

c/Travesera de Les Corts c/Justo Dorado

39-43.-08020-Barcelona 11.-28040-Madrid

9. SPONSO R ING OR GAN IZAT ION - N AM E AND ADD RESS (if NR C. type '~Same as above",, if contractor. provide NRC Division, Office or Region. U.S. Nuclear Regulatory Commission,

and mailing address.) Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words orklss)

An assessment of RELAP5/MOD2 cycle 36.04 against a turbine trip from 100% power in Vandell6s Il NPP (Spain)

is presented. The work is inscribed in the framework of the Spanish contribution to ICAP Project.

The model used in the simulation consists of a single loop, a steam generator and a steam line up to the steam

header all of them enlarged on a scale of .3: 1; and full-scaled reactor vessel and pressurizer.

The results of the calculations have been in reasonable agreement with plant measurements. An additional study has

been performed, to check the ability of a model in which all the plant components are full-scaled to reproduce the

transient. A second study has been performed using the Homogeneous Equilibrium Model in the pressurizer trying

to elucidate the influence of the velocity slip in the primary depressurization rate.

12 KEY WORDS/DESCR!PTORS ('List words or phrases that will assist researchers in locating the report.) 13. AVAILABILýITY STATEMENT

ICAP, RELAP5/MOD2, Vandellos 11, transient, turbine trip unlimited

14. SECURITY CLASSIFICATION

fThis Pawe)

unclassified

(This Report) unclassified

15. NUMBER OF PAGES

16. PRICE

NRIC FORM 335 (2-89)


Federal Recycling Program

ASSESSMENT OF RELAP5/MOL)2 AGAINSI A 'IUKu1NkN1 'IU!I 1Viturvi NUREG/IA-0 108 100% POWER IN THE VANDELLOS Il NUCLEAR POWE R PLANT

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

JUNE 1993

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300