NUREG/IA-0 109

UNID-9 1-08

International

Agreement Report

Assessment of RELAP5/MOD2

Against a 10% Load Rejection

Transient from 75% Steady State

in theVandello's 11 Nuclear

Power Plant

Prepared by

C. Liopis, Asociaci'on Nuclear Vandello's

J. Perez, Consejo de Seguridad Nuclear

A. Casals, Asociacio'n Nuclear Vandell6s

R. Mendizabal, Consejo de Seguridad Nuclear

Asociaci-n Nuclear Vandell6s

c/Travesera de Les-Corts 39-43.-08028 Barcelona

Consejo de Seguridad Nuclear

c/Justo Dorado 11.-28040-Madrid

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

May 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 thereQf, 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-0109 @ UNID-91-08

International

Agreement Report

Assessment of RELAP5/MOD2

Against a 10% Load Rejection

Transient from 75% Steady State

in theVandelids 11 Nuclear

Power Plant

Prepared by

C. U~opis, Asociaci6n Nuclear Vandell6s

J. Perez, Consejo de Seguridad Nuclear

A. Casals, Asociaci6n Nuclear Vandell6s

R. Mendizabal, Consejo de Seguridad Nuclear

Asociaci~n Nuclear Vandell6s

clTravesera de Les Carts 39-43.-08028 Barcelona

Consejo de Seguridad Nuclear

cliusto Dorado 11.-28040-Madrid

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

May 1993

Prepared as part of

The Agreement on Rescarch Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program ([CAP)

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.
FO0R E WOR D

This report has been prepared by A.N. Vandello's in the

framework of the ICAP-UNESA Project.

The report represents one of the assessment calculations

submitted in fulfilment of the bilateral agreement for cooperation in thermaihydraulic activities betwe -en the Consejo

de Seguridad Nuclear of Spain (CSN) and the United States

Nuclear Regulatory Commission (USNRC) in the form of Spanish

contribution to the International Code Assessment and Applications Program (ICAP) of the USNRC whose main purpose is

the validation of the TRAC and RELAP system codes.

The Consejo de Seguridad Nuclear has promoted a coordinated Spanish 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 organizations:

- Unidad Electrica (UNESA)

- Uni6n Iberoamericana de Tecnologia E16ctrica (UITESA.)

- Empresa Nacional del Uraxaio (ENUSA)

- TECNATOM

- EMPRESARIOS AGRUPADOS

- LOFT-ESPANA

The program is executed by 12 working groups and a generic code review group and is coordinated by the "Comite' de

Coordinacion" This committee has approved the distribution

of this document for ICAP purposes.

D-1 /90-MPNV pg
TABLE OF CONTENTS

PAGE

ABSTRACT ....................... .. ... .. ........... 3

EXECUTIIVE SUMM'A.RY . . ... . .. .. * ............. .. . .. A

1.* INTRODUCTION . . ... .. .. .. .. . ... ... . .. ........ 5

2. PLANT AND TRANSIENT DESCRIPTION........6

2.*1. PLANTI DESCRIPTION . . ............. . 6

2.2. DATA ACQUISITION AND ANALYSIS SYSTEM

DESCRIPTION . .o.. . .. oooo ... o..o.. ... o.7

2.3o.TRANSIENT DESCRIPION ION~o...... ..........

3 . MODEL DESCRIPTION . .... .... .... .. o. .. . 10

3.1. PRIMARY SYSTEM AND STEAM GENERATORS .. 11

3.*2o. SECONDARY SYSTEM . . . . .... o.- . .. o . . .. o. . 13

3.3. CONTROL SYSTEMS . .. .... .o. .o. .o.o. . . .. .. o 14

4. STEADY STATE CALC ULATINoo~o NS.. ....o.o.....16

5o TRANSIENT CALCULATION AND COMPARISON

VERSUS ACTUAL DATA o..o. .ooo .... 17

6o RUN STATISTICS o...... oo...... o..****... oo.... o20

7o CONCLUSIONS .. . .. oo. . . . .oooo. ..... ... o.o....21

8.. BIBLIOGRAPHY .... . .~. .. ..... *...... 22

9o. INDEX OF TABLES . ....... o.... .. o..23

10. INDEX OF FIGURES . .. .. ........ o...... o .. ... o30

page 2
ABSTRACT

The Consejo de Seguridad Nuclear (CSN) and the

Asociaci6n Nuclear Vandell6s have developed a model of

Vandellos II Nuclear Power Plant. The ANy collaboration

consisted in the supply of design and actual data, the

cooperation in the simulation of the control systems and other

model components, as well as in the results analysis.

The obtained model has been assessed against the

following transients occurred in plant:

- A trip from the 100% power level (CSN)

- A load rejection from 100% to 50% (CSN)

- A load rejection from 75% to 65% (MWV)

- A feedwater turbopuinp trip (ANV)

This copy is a report of the load rejection from 75%

to 65% transient simulation. This transient was one of the

tests carried out in Vandell6s II NPP during the startup tests.

page 3
EXECUTIVE SUMMARY

ýThe Vandellos II NPP, *owned by ENDESA (72 %) and

HIDROELECTRICA. ESPAROLA .(28 %), is located in Tarragona

(Spain), by the Mediterranean sea. Its commercial operation

started on March 3,' 1988.-

'The Vandellos II NPP obtained the code RELAP5/HOD2

through the ICAP project. Then, Vandellos II NPP collaborated

with the CSN simulating and analyzing two of the four

tr .ansients -the' CSN had prepared for ICAP. However, Vandellos

II NPP had already some experience in the use of this code due

to previous collaboration agreements with the CSN.

This-transient has been selected because of these two

reasons:

-Enough plant data were available to check the

results.

-The initial steady state is 75 %,instead of the

habitual 100 %, this allows checking the model

behavior in this new power level, so that the

100%, 75%, 65%, and 50% levels have been simulated

and analyzed.

The main conclusions of this analysis are the

following:

- Close agreement between results and data.

- The RELAP5/MOD2 is a valuable tool to simulate the

primary side behavior.

- Basically, the differences between the model

results and the plant data are due to the

secondary side behavior during the transient: high

sensibility to steam. flow fluctuations, the

indeterminateness of plant data and the accuracy

of the reactor kinetics calculations (specially,

the Doppler effect calculations).

page 4
INTRODUCTION

The Asociaci6n Nuclear Vandellos II (ANy) decided, at

the beginning of the commercial operation, to promote ef forts

aiming to study the following topics related to the simulation:

- The analysis of plant actual transients.

- The preparation for future IPE (Individual Plant

Examination) works.

- The support to Vandellos II Operation Department

in simulations of transients and accidents.

- The simulation of FSAR design accidents by means

of a best estimate model, in order to compare them

to the results obtained using conservative codes.

- The colaboration in the ICAP project with the

analysis of two transients.

This work is one of the contributions of Vandellos II

NPP (inside the UNESA group) to the ICAP project.

Other works have been carried out in order to support

Vandell6s II NPP Emergency operation Procedures Rewiev and in

the near term the contribution to the IPE is expected to begin,

the experiencie gained during the collaboration in the ICAP

project is considered to be very valuable for this

contribution.

page 5
2. PLANT AND TRANSIENT DESCRIPTION

2.1. PLANT DESCRIPTION

Vandellos 1I is a three-loop PWR Nuclear Power Plant,

designed by Westinghouse, with a nominal thermal power of 2775

MWt. It: is equipped with three Westinghouse U-tube steam

generators (model F) without preheaters. 'The feedwater is fed

through the upper portion via J-tubes. The vessel is cold head

type. The nominal electrical power is at present 992 MW.

Plant features are shown in table I.

page 6
2.2 PLANT DATA ACQUISITION SYSTEM DESCRIPTION

To record the main parameters of the plant, during

the startup tests period, a temporary data acquisition system

was installed. It consisted of a digital system. with an up to

0.05 seconds and 146 signals trail capacity.

The recorded parameters depended on the test carried

out.

The use of this system permitted a better and faster

review of the test results. Therefore, once the nuclear plant

tests had finished, Vandellos II NPP decided to install a final

similar data acquisition equipment in order to interprete the

plant behavior.

The availability of such a great number of signals

has allowed th use of RELAP to check the control blocks partial

performances, specially the feedwater control block and the rod

control block, which are the main contributors to the plant

evolution in this transient.

page 7
2.3 TRANSIENT DESCRIPTION

The test which is the subject of interest of the

current simulation was performed on January 31, 1988 in

Vandellos II NPP.

The objective of this test was to check that the

reactivity introduced by the rod system was enough to absorb up

to'a 10A% load variation.

At the beginning of the test the plant was

aproximately at 75 % power level steady state: the examinated

and recorded parameters were the following:

- Electrical power - Nuclear power

- Average temperature - Steam pressure - Pressurizer pressure - Pressurizer level - Steam generators levels. - Feedwater flow - Steam flow - Feedwater turbopunips speed

The control of the following systems was in automatic

mode:

- Rod control - Steam generators level control.

- Turbopumps speed control - Pressurizer pressure Control - Steam-dump control - D.E.H.

Once the power level was checked to remain stable, a

10 % load rejection was forced to happen by means of the D.E.H.

at a 200 %/min rate, and a new power level at approximately the

65 % rated conditions was reached. The plant stabilized at

this new level.

page 8
The expected evolution of the main parameters of the

plant were the following:

- Primary pressure variation: 3.5 Kg/cm2- - Steam generators levels variation: ± 5% - Steam pressure variation: ± 1.75%

The test acceptance criteria were the followilng:

- Neither the reactor nor the turbine should trip.

- The safety. injection should not actuate.

- Neither the pressurizer safety valves nor. the

pressurizer relief valves should open.

-The control system must lead the plant to a new

steady state automatically (no hand operation is

allowed). - During the 10 %load rejection the steam dump

should not open (the transient must be absorbed by,

the rods system.

- The plant variables -such as the average

temperature, the steam. pressure, the pressurizer

pressure, the f eedwater, flow, thei steam flow, the

levels - should not oscillate either continuosly o

divergently.

The test results were satisfactory and in close

agreement with the results obtained with RELAP5/MOD2 .

page 9
3. MODEL DESCRIPTION

Figure 2 shows the

primary system of the plant.

junctions, 78 heat structures

nodalization used to simulate the

It consists of 117 volumes, 122

and 107 control variables.

A single loop which simulates the three loops of the

plant has been implemented, the reason of this simplification

is the reduction in the computing time; however, inaccuracy is

not introduced with this simplification. A three loops model

has been developed, and some tests have been carried out in

order to compare the results obtained with this model to the

results obtained with the single loop model. This tests have

susbstantiated the single loop model validity for symmetric

transients.

page 10
3.*1. PRIMARY SYSTEM AND STEAM GENERATORS

This model includes the vessel, the primary loops,

the steam generators, the pumps and the pressurizer.

The single loop model requires triplicating the

volumes, the surfaces and heat structures transmission surfaces

of the primary loops and steam generators.

The components of this model have been ellaborated

and checked singly. For example, the steam generator was

tested separately from other components and with the plant

calorimetric data. The objective of this test was to adjust

the primary - secondary heat transfer and the steam generator

pressure. Another example is the comparison of the pressurizer

behavior versus the plant spray and. he'aters performance.

The main components of the vessel are the following:

- Volume 504:

- Volume 510:

- Volume 520:

- Volume 530:

- Volume 535:

- Volume 540:

- Volume 550:

- Volume 560:.

Downcommer

lower plenum

from lower core support forging to

lower core plate.

core

Between internals core barrel and

baffles, and other core by-pass

from upper core plate to mid loop

elevation.

from mid loop

support assembly.

elevation to upper

from upper support assembly

internals flange elevation.

to

- Volume 580: upper plenum.

The vessel by-pass design flow has been adjusted

through the volume 535 _(core by-pass) and the volumes 502, 500

and 580 (vessel head cooling) by means of the energy loss

coefficients.

page 11
The main components of the steam generator are the

following:

Secondary side:

- Volume 200: Boiler

- Volume 220: expansion zone in the boiler upper

portion.

- Volume 310: downcommer.

- Volume 230: turboseparators tubes lower portion

- Volume 240: turboseparators.

- Volume 280: turboseparators external zone.

Primary side:

- Volume 120 and 140: water boxes.

- Volume 130: steam generator tubes.

The recirculation ratio at 100%, 75% and 65% power

levels has been substantiated to fit the design values.

Besides, vessel loops, steam generators and pumps

pressure drops have been successfully checked.

The pressurizer has been divided into 10 volumes; two

of these divisions match the pressurizer levels at 0% and 100%

power levels.

The pressurizer relief and safety valve controls have

been simulated, but not the valves themselves. This allows

verifying that in this transient these valves do not open.

Talking about kinetics, the moderator temperature

coefficient has been considered to be equal to zero, since when

the test was carried out the core was in the beginning of lif e.

An important observation is that the rod position modifies this

coefficient, but RELAP5/MOD2 does not allow simulating this

variation.

page 12

t,. --
3.2. SECONDARY SYSTEM

In the secondary side, the three steam generators and

the lines to the steam header, have been simulated as a single

steam generator and a single line. The lenghts of the lines

have been averaged since the three lines are not exactly equal.

The steam generator relief valves have been

simulated, and the safety ones have been simulated as a single

TMDPJUN. However, in this transient they do not open.

Downstream of the header, the four turbine admission

valves have been simulated as a single valve. The steam-dump

valves, which in plant are 12 gathered into 4 groups and which

discharge into the three condenser shells, have been simulated

as four valves to simulate four benches. The MSR's, ejectors,

and turbopumps consumptions have also been simulated.

page 13
3.3. CONTROL SYSTEM

The primary basic controls can be groupped into f our

groups:

- Rod control

- Pressurizer pressure and level control

- Feedwater control

- Turbine and steam dump control

The four groups have been simulated according to the

plant design. The plant actual control setting values during

the test have been used as a setpoints.

The control blocks diagrams are shown in figures 3,

4, 5, and 6.

The availability of the signals continous recording

system through the data acquisition system, has allowed

checking all the control systems, and it has been observed that

plant data are in close agreement with RELAP5/MOD2 results.

It has not been possible, however, matching the

reactor kinetics to the plant reponse accurately. This and the

steam f low are the main contributors to the RELAP5/MOD)2 results

and to the plant response mismatching in the load rejection

from 75% to 65% transient simulation.

Another item that has not been possible to adjust

accurately is the rod control system, since the pressure in the

turbine first stage impulse chamber is the signal taken to

measure the turbine power, and this signal 'is not available

with RELAP5/MOD2 since the turbine has not been simulated.

With RELAP5/MOD2 the steam flow has been used as the

secondary power measure, since is a variable more related to

the impulse chamber pressure than the valve position. For

example in figure 20 a porcentual comparison of the following

variables is shown:

- Turbine valve position

- Impulse chamber pressure

- Steam flow

- Electrical power

page 14
and the differences among them can be evaluated.

A transient calculation has been carried out

considering the impulse chamber pressure as a boundary

condition, and the results have been much satisfactory.

Nevertheless, this method has not been used since it would be

useful only in those cases which have occurred in plant and

this signal is available.

page 15
4. INITIAL STEADY STATE CALCULATIONS

Starting from a 100% initial level and simulating a

load rejection transient to the 75 %, by closing the turbine

regulating valve, a new steady state has been reached by means

of the control systems, which have led the model to the new

conditions. The INPUT has been reinitialized with the obtained

data. In fact, this means having simulated a new case of load

rejection from 100 % to 75 %.

The main parameters values obtained with RELAP5/MOD2

have been compared to the plant actual values, as shown in

Table III.

page 16
5. TRANSIENT CALCULATION AND COMPARISON VS ACTUAL DATA

The main purpose of this transient simulation is to

check the model at different power levels. Moreover, being the

transient developement governed by the controls - specially by

the rod control, main feedwater control, and the pressurizer

pressure and level control - another purpose is checking the

behavior of these controls.

The simulation of this transient has been carried out

starting from the initial steady state, reducing the turbine

flow and allowing the control systems to perform automatically.

In plant, this flow reduction is governed by the D.E.H., which

commands a 200 % / min. load rejection, so that in 3 seconds

the turbine admission valves are about to~be at their new

position.

As a result of the order given by the D.E.H. and the

new turbine valve position, the steam flow diminishes (fig. 17).

The second immediate consequence is the reduction in the first

stage impulse chamber pressure, which is being used as a

turbine power reference by the rod control system.

*in fact, a close relationship exists between the

steam flow and the impulse chamber pressure, but it is not a

linear relation during a transient. In our model, the turbine

steam flow has been adopted as the turbine power measure.

Owing to the, variation in the energy production and

evacuation balance of.the primary side, new variations of

pressures and temperatures occur. The reactor will attempt to

adapt the new power. level, by means of the rod control system,

which will move the rods as a result of the power error and the

average temperature (fig.7).

The nuclear flux decreases quickly (fig.8) and, from

there on, the reactor will adopt a new average temperature

according to the temperature program.

The 63% and 65% power levels difference Ibetween plant

and RELAP values, respectively at the end of the simulation, is

caused by the error margin- of the plant power measurement. A

thermal balance carried out at the end of the transient points

out that the RELAP value is more reliable than the plant value.

page 17
The cold leg, hot leg and average temperatures are

shown in figures 10, 11 and 12. At the beginning, the average

temperature increases because of the reactor power production

and the steam generator power evacuation mismatch, later, once

this mismatch has been overcome, the average temperature

decreases down to the new level because of the nuclear flux

reduction.

Figure 13 has been included to show the primary delta

Temperature evolution as a significant indicator of the primary

power evolution. This delta Temperature allows a close

adjustement with RELAP, which indicates that the level reached

is the same than the plant one. However, as it has been seen

before, it has not been possible to adjust the nuclear flux to

the plant values. These two points prove that the plant

nuclear flux measurement has an error margin.

The primary pressure evolution is similar to the

average temperature one, and is shown in figure 14. The same

occurs with the pressurizer level (f ig.15), which is modified

essentially by the density variation in the primary side.

It can be observed an initial pressure peak which is

higher in RELAP than in plant. This fact may b~e caused by the

ýspray efficiency, since the RELAP code does not allow

,-simulating the actual physical phenomenon.

The primary flow (fig.9) evolution is due to the

density variation too. It can be observed that plant measures

have a lot of noise, and this happens because it is calculated

as elbow taps pressure difference, so that small variations in

the fluide regime imply small (0.1%) variations in the

measurements.

The steam generators pressure (fig. 18) increases

initially because of--the turbine valve 'closure, but it

stabilizes when the heat transmission through the steam

generator tubes and the evacuation through the turbine balance.

The steam generators level (f ig.19) has a sudden fall

on starting the transient because of the steam binding but it

recovers on increasing the feedwoater flow (fig. 16).

The curling that can be observed in the steam

generators level is supposed to be due to void fraction

variations owing to any numerical inestability of the

RELAP5/MOD2.

page IS
The outstanding adjustement of the feedwater control

system with RELAP5/MOD2 has been achieved owing to the

availability of intermediary recorded signals of the plant

control systems, which have allowed working with partial

models.'

The sole boundary conditions which have been imposed

to the model in this transient are the turbine steam flow and

the pressure and temperature of the feedwater header, since the

secondary side next to the valves has not been simulated by the

moment.'

page 19
6. RUN STATISTICS

This case has been simulated on an IBM 3090, owned by

ENDESA, located in Madrid.

RELAP5/MOD2 cycle 36.04 has been used in the version

adapted by ISPRA the 1st of November, 1987.

The CPU TIME / REACTOR TIME ratio has been 4.31.

The time step has been constant (0.05 sec.) during

all the transient.

The run statistics are shown in Table V.

page 20

a -
7. CONCLUSIONS

The Dispose of such a great number of digital signals

from the data acquisition system has allowed carrying out

simulations, in which plant signals have been supplied to the

control blocks in order to check each block separately. These

checks have permitted and supported the individual control

blocks assessment prior to assemble the complet model. This

process has been shown to conduct to very, accurate results

predictions.

In this load rejection transient, in which a control

rods insertion occurs, the reactor behaves in a different way

depending on the axial zone we consider. To reproduce the

final power level with RELAP correctly, it has been necessary

to modify the Doppler coefficient design values. This fact

lead us to conclude that the punctual kinetics is a

conservative model that has to be corrected for this kind of

transients.

The power initial level in this transient is the 75%,

instead of the habitual 100%, and this has allowed checking the

model behavior in this new power level. The obtained results

are very satisfactory, as shown in Table III.

The evolution of most of the RELAP main variables in

this transient are in close agreement with plant data.

Besides, in these cases in which mismatches can be observed,

the differences are within the plant instrumentation error

margins.

After having assessed this model against the

following transients:

- A trip from the 100% power level (CSN)

- A load rejection from 100% to 50% (CSN)

- A load rejection from 75% to 65% (ANy)

- A feedwater turbopump trip (ANV)

the RELAP5/MOD2 is considered to be a valuable tool for

transient simulations.

page 21
8. BIBLIOGRAPHY

Precautions, Limitations and Setpoints of Vandell6s II,

rev. 6.

Estudio final de seguridad de Vandellos II

Setpoints Study Vandellos unit 2 (WENX/85/38)

RELAP5/HOD2 Code Manual

Project Information Package

Steam Generator thermohydraulic analysis

Copia oficial de prueba.PAN 55

Copia oficial prueba eficacia calentadores y spray

Copia oficial prueba reactividad de barras

Datos sistema Adquisici6n de Datos de planta

page 22
9. INDEX OF TABLES

TABLE I

TABLE II

TABLE III

TABLE IV

TABLE V

DESCRIPTION OF THE MAIN CHARACTERISTICS OF

VANDELLOS II NUCLEAR POWER PLANT

MAIN EVENTS THAT TOOK PLACE DURING THE

TRANSIlENT

COMPARISON BETWEEN RELAP5/MOD2 VALUES AND

ACTUAL DATA FOR STEADY STATE

DESCRIPTION OF RELAP5/MOD2 VARIABLES

RUN STATISTICS

page 23
TABLE I

MAIN CHARACTERISTICS OF VANDELLOS II NPP

- THERMAL REACTOR-POWER (MWt) .............. .2775

- ELECTRICAL POWER (MWe) ............... . .. . 992

- FUEL o. . . ..... o...o ........ . ...... U02

- NUMBER OF ASSEMBLIES.............................. 157

- NUMBER OF COOLANT LO0OPS ............ 3

- CLADDING TUBE MATERIAL ... o..... o... o ZIRCALOY 4

- ABSORBER MATERIAL o.o....... o..... .o.... B4C + Ag-In-Cd

- REACTOR OPERATING PRESSURE (MPa) ........o....o.. 15.4

- COOLANT TEMPERATURE AT NO LOAD (-K) ....... o.. 564.8

- COOLANT AVERAGE TEMPERATURE AT 100% (OK) ... o 582.3

- STEAM GENERATOR ...... .. * ....... WESTINGHOUSE TIPE F

- NUMBER OF TUBES IN STEAM GENERATOR ..... o..o 5626

- TOTAL TUBE LENGHT(i) .... oo ...... ooo98759

- INNER DIAMETER TUBES ( .) . o.o.ooo...o.. 0.0156

- TUBE MATERIAL...... o... . . ..... o ....... INCONEL

- PUMPS TYPE - ... o....... .o.... oWESTINGHOUSE D 100

- DISCHARGE HEAD OF PUMPS (bar.) . ...o......o18.8

-DESIGN FLOW RATE (m3/s) .... *..*.....oo...... 6.156

-SPEED OF PUMPS (rad/s) ..... o.......... o.....155

page 24
- PRIMIARY V'OLUMHE (m3) .............. 106.*19

- PRESSUJRIZER VOL~UME(m3) ................ 39.65

- HEATING POWER OF THE HEATERS RODS (K(W) ...... 1400

- MAXIMUMSPRAY FLOW (Kg/s) ............. 44.2

- STEAM MASS FOOW RATE AT 100 %6 (Kg/s) ....... 1515

page 25
TABLE II

MAIN EVENTS

TIME

0. 0 SEC.

3. 0 SEC.

APROX. 600 SEC.

EVENT

BEGINNING OF REJECTION LOAD FROM 75%

TO 65 %6

TURBINE VALVE AT A NEW POSITION

REACHED NEW STEADY STATE OF 65 *OF

POWER

page 26
TABLE III

COMPARISON BETWEEN RELAP5/MOD2 VALUES AND ACTUAL DATA

VARIABLE

NUCLEAR POWER

PRIMARY MASS FLOW RATE

COLD LEG TEMPERATURE

HOT LEG TEMPERATURE

AVERAGE TEMPERATURE

DELTA TEMPERATURE

PRESSURIZER PRESSURE

PRESSURIZER LEVEL

FEEDWATER MASS FLOW RATE

STEAM GENERATOR PRESSURE

STEAM GENERATOR LEVEL N.R.

RECIRCULATIO RATIO

M%

(Kg/s)

(OK)

(OK)

(OK)

(*K)

(MPa)

(Kg/s)

(MPa)

M%

RELAP5/MOD2

73.9

14907

564.3

589.5

576.9

25.2

15.50

50.0

1120

6.86

50.0

3 .62

PLANT

73.8

14907

564.5

589.7

577.1

25.2

15.49

49.9

1117

6.86

*50.3

3.6 (2)

(1) CALCULATED DATA

(2) DESIGN DATA

page 27
TABLE IV

DESCRIPTION OF RELAP5/MOD2 VARIABLES FIGURES

CNTRLVAR 340

CNTRLVAR 301

MFLOWJ .180010000

CNTRILVAR 328

CNTRILVAR 327

CNTRLVAR 330

CNTRLVAR 947

P 415090000

CNTRLVAR 350

MFLO0WJ 325000000

MFLOWJ 715000000

P 600010000

CNTRLVAR 203

ROD POSITION

NUCLEAR POWER (PERCENT)

PRIMARY MASS FLOW RATE

TEMPERATURE AT THE COLD LEG

TEMPERATURE AT THE HOT LEG

AVERAGE TEMPERATURE

DELTA TEMPERATURE

PRESSURIZER PRESSURE

PRESSURIZER LEVEL

FEEDWATER MASS FLOW RATE

TURBINE STEAM MASS FLOW RATE

STEAM GENERATOR PRESSURE

STEAM GENERATOR LEVEL (N. R.)

7

8

9

10

11'

12

13

14

15

16

17

18

19

page 28
TABLE V

RUN STATISTICS

COMPUTER

TRANSIENT TIME

CPU TIME

C (TOTAL NUMBER OF ACTIVES VOLUMES)

DT (TOTAL NUMBER OF TIME STEPS)

CPU * :1000

= 1.84------

C * DT

CPU TIME / TRANSIENT TIME

IBM 3090

800 sec

3452 sec

117

16000

4.31

page 29
10.* INDEX OF FIGURES

FIGURE

FIGURE

FIGURE 3.

FIGURE

FIGURE

FIGURE -

FIGURE 7

FIGURE

FIGURE

FIGURE Ii

FIGURE 11

FIGURE 1

FIGURE 1:

FIGURE 1

FIGURE 1

FIGURE 1

FIGURE 1

FIGURE 1

FIGURE 1

FIGURA 2

L.

5.

6.

9.

2.

3.

4.

5.

6.

7.

8.

9.

0.

VANDELLOS II N.P.P. DIAGRAM

NODAIJIZATION OF C.N.VANDELLOS II

ROD CONTROL SYSTEM BLOCK DIAGRAM

PRESSURIZER PRESSURE AND LEVEL SYSTEM

TURBINE CONTROL AND STEAM-DUMP SYSTEMS

FEEDWATER CONTROL SYSTEM

ROD POSITION

NUCLEAR POWER%

PRIMARY MASS FLOW RATE

TEMPERATURE AT THE COLD LEG

TEMPERATURE AT THE HOT LEG

AVERAGE TEMPERATURE

DELTA TEMPERATURE

PRESSURIZER PRESSURE

PRESSURIZER LEVEL

FEEDWATER MASS FLOW RATE

TURBINE STEAM MASS FLOW RATE

STEAM GENERATOR PRESSURE

STEAM GENERATOR LEVEL (N.R.)

POWER COMPARISON

page 30
FF -fTl

PCV-AS449A

QPcv-As-uow ~PCV-AS4476

PCV-AD4489

SPCV-AR-U96

zPCV-AI-432

PCV-A)-4fl

PCV-A*4"C

wA

VANDELLOS II PRIMARY AND MAIN ASSOCIATED SYSTEMS FGFIG. I - I
F

(A,

n"stlE

PugE

INODALIZATION C.N. VANDELLOS III

F IGURE 2

"AamL VE1sL

I

ewI

WM£80
r

CA~
r

330

wA GOV3 KV)

4

~Y czmTum.

4

PMSSUUIER

p 4t5OM O

PPF'ýRIIRTZER PRESSURE AND LEVEL CONTROL FIG. 4
r

MMAWR MW

(A)

01

On/ LW Lt-, Sw

MAIN AND AUXILIARY FEEDWATER CONTROL FGFIG.
r

8750

SETEA PRESSURE 60000RAETYE OPE VALVESPE

TM1DPJUN 710

-*~ ~ ~~C c'1l4 -r"'.m F- A~r ~l WD~~ rrflNTPnl
FIG 7: ROD POSITION

BANK D

200

190

180

170

160

150

140

130

120

0 200 400 600 800

SECONDS

- PLANT - RELAP
FIG 8: REACTOR POWER

wA

80

79 -_ _ _ . -.....- - -

78-

76-

74- ....... ...... . . . . . . . . .

71 _ _

70-- _ _ _

69-

68

64__ _

613 _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ __ _

62 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

0 200 400 600 800

SECONDS t0- Ir
I.

FIG 9: PRIMARY. MASS FLOW RATE

15.2

15.1.

15

w 1-1

,ko tn u)

14.9

14.8

14.7

14.6

0 200 400 600 80

SECONDS

PLANT - RELAP
FIG 10: TEMPERATURE AT THE COLD LEG

570

569

568

567

566

-r11

565 -

564 -

563 -

562 -

561 -

560 -

I

___________-*---------. I---..

I I I -I 4- I ------ I I

0 200 400 600 800

SECONDS
FIG 1 1: TEMPERATURE AT THE HOT LEG

4~.

592

591

590

589

588

587

586

585

584

583

582

.-.---------- 4.-----.-.-----4-.---- I- -

-I-*-*---------- $ I -- ~~~-~-----I -~~---------~- -I

-I I- I - , 1'

0 200

-- PLANT

400

SECONDS

RELAP

600 800
FIG 12:. PRIMARY AVERAGE TEMPERATURE

580

579--- -

578

577

576 4-1

9;-79

574-

573

572 -

571 -

570-

I- l-.---.- 4 I- 4------- 4---.-**-4-------

4 I 4- 4 1 4 -I I I

0 200 400 600 800

SECONDS
FIG 13: VESSEL DELTA TEMPERATURE

30 -I -, -- - -. - 1*

27-

26-

25-

24-

23-

22-

21-

+1-- 4---- 1 ---- 4 4

------ I20 -t------ A -

0 200 400 600 B00

SECONDS

-- PLANT -- RELAP
DC

2l16

15.9

15.5

15.7

15.6

15.5

15.4

15.3

15.2

15.1

15

FIG 14: PRESSURIZER PRESSURE

I I

. I S ---....-t...-......-....----4--.

I I -. _____________ I....- ______-- _______________ ________________ I __ __

I.. -

-I I 1 4- 1 F I

0 200 400

SECONDS

600 B00

-. f a %
FIG 15: PRESSURIZER LEVEL

-P.

60

59

58

57

56

55

54

53

52

51

50

49

48

47

46

45

44

43

42

it 1

40

2i~.

N-

~f.7II777.

N 4-

-I--.----4------------I-~

F~--..-- I

Z I t t 4-

0 200 400 600 800

SECONDS

- -PLANT RELAP
FIG 1 6: FEEDWATER MASS FLOW RATE

1.3

1.2

0.9

0.8

0 200 400 600 800

SECONDOS
FIG 17: STEAM MASS FLOW RATE

1.3 -' ~---~-1 .T -.

1..2.. -f--- - - .. -I "--- ....-

1.1 -t-----*--~--*

WU)

N 2 

I--,

0.9-

0.8 -

Dý6ý

I I 4- F I-. F I

400 600 800 0 200

SECONDS

PLANT RELAP
FIG 18: STEAM GENERATOR PRESSURE

7.5 -

7.4

7.3 -

~~1~'

*1**

e-%

a7.1-

7-

6.9-

F- - I -I-----.--I. *1~~~.--* --

6.8 -

6.7 -

6.6 -

6.5 -

-. - --- ~*1

0 200 400 600 B00

SECONDS __ PCI AP
FIG 19: STEAM GENERATOR LEVEL

(NARROW RANGE)

sn

59 -

58

57

56

55

54

53

52

51

50.

49

48

47

46

45

44

43

42

41

40

L I---.-

Md'd

~i~Lj- I

-...... I

a I I It.. I Ill. J

I -.

'I. -.

liii. 1111111, - ........ .. ~ ~I n .. rIu,.*1.

600 800 0 200 400

PLANT

SECONDS _ RELAP
FIG 20: POWER COMPARISON

(percent)

75

70

65

607

I I

0 200 400 600 800

seconds

50
NR, ORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

128.L) *lAuslmned by NRC. Add Vol.. Supp.. Rev.. NRCM 1102. an~d Addendum Numbers. If any.) 3201, . 302 BIBLIOGRAPHIC DATA SHEET NIJREG/IA-0l 09

(See instructions on the teerestl LJNID-91 -08

2. TITLE AND SUBTITLE

Assessment of RELAP5/MOD2 Against a 10% Load Rejection Transient from 75% 3. DATE REPORT PUJBLISHED

Steady State in the Vandell6s Il Nuclear Power Plant MONTH YEAR

May 1993

4. FIN OR GRANT NUMBER

_______________________________________________________ L2245

5. AUTHOR(S) 6. TYPE OF REPORT

C. Liopis, Asociacibn Nuclear Vandell6s

J. Perez, Consejo de Seguridad Nuclear Technical Report

A. Gasals, Asociacio-n Nuclear Vandell6s 7. PERIOD COVER ED iinclusise Dat*7

R. Mendizabal, Consejo de Seguridad Nuclear

8. PERFORMING ORGANIZATION - NAME AND ADDR ESS O(II RC. provide Division. Officeor Region, U.S. NucleaffRegulItar" Com~mission, and maiiing address.- ifcon tractne. provide.

Pam# e nd mailing aldervLl

Asociaci6n Nuclear Vandell6s

c/Travesera de Les Corts 39-43.-08028 Barcelona

Consejo de Seguridad Nuclear

c/Justo Dorado l1.-28040-Madrid

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (II NRC. type -Sameastabo ye if Contractot. provide NRC Division. Office or Region. U.S. Nucleer Regulatory Com~mission. ansd mailing addressi

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

[ 0 U-PPLEMENTARY NOTES

11. ABSTRACT (200 words of l*ss1

The Consejo de Seguridad Nuclear (CSN) and the Asociaci6n Nuclear Vandell6s (ANY) have developed a model of

Vandell6s 1I Nuclear Power Plant. The ANY collaboration consisted in the supply of design and actual data, the

cooperation in the simulation of the control systems and other model components, as well as in the results analysis..

The obtained model has been assessed against the following transients occurred in plant:

- A trip from the 100% power level (CSN)

- A load rejection from 100% to 50% (CSN)

- A load rejection from 75% to 65% (ANV)

- A feedwater turbopump trip (ANV)

This copy is a report of the load rejection from 75% to 65% transient simulation. This transient was one of the tests

carried out in Vandell6s Il NPP during the startup tests.

12. KEY WO RDS/DESCR 1PTORS (List words or phtrases that will assist researchers In locating the rep~ort. J 13. AVAILABILITY STATEMENT

ICAP, I Tnlimiitnvid

Vandll~s[I,14. SECURITY CLASSIFICATION

RELAP5, (This Pagel

Transient, Unclassified

Load Rejection (hsrot

Unclassified

15. NUMBER OF PAGES

16. PRICE

NFIC FORM 335 12.89)
I.

Federal Recycling Program

NUREGA-0109ASSESSMENT OF RELAPS/MOD2 AGAINST A 10% LOAD REJECTION TRANSIENT

NUREG/A-0109FROM 75% STEADY STATE IN THE VANDELL6s II NUCLEAR POWER PLANT

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

MAY 1993

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300