NUREG/IA-Ol 10

NUREG/IA-0110

-ICSP-V2-TURFW-R

International

Agreement Report

Assessment of RELAP5/MOD2

Against a Main Feedwater

Turbopump Trip Transient in the

Vandellos II Nuclear Power Plant

Prepared by

C. Llopis, A. Casals/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 20555-0001

December 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

NUREG/IA-Ol 10

NUREG/IA-0110 -I trto a ICSP-V2-TURFW-R

International

Agreement Report

Assessment of RELAP5/MOD2

Against a Main Feedwater

Turbopump Trip Transient in the

Vandellos II Nuclear Power Plant

Prepared by

C. Llopis, A. Casals/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 20555-0001

December 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--0110

ICSP-V2-TURFW-R

International

Agreement Report

Assessment of RELAP5/MOD2

Against a Main Feedwater

Turbopump Trip Transient in the

Vandellos II Nuclear Power Plant

Prepared by

C. Llopis, A. Casals/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 20555-0001

December 1993

Prepared as part of

The Agreement on Research Participation and Tbchnical Exchange

under the International Thermal-Hydraulic Code A..essment

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.
FOREWORD

This report has been prepared by A.N. Vandell6s 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 thermalhydraulic activities between 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 El~ctrica (UNESA)

- Uni6n Iberoamericana de Tecnologia Elctrica (UITESA)

- Empresa Nacional del Uranio (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

Coordinaci6n". This committee has approved the distribution

of this document for ICAP purposes.

D-1/90-MPNV
TABLE OF CONTENTS

PAGE

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

EXECUTIVE SUMMARY ..................... o.......... 4

I. INTRODUCTION ........ .................. . .o. 5

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

2. I. PLANT DESCRIPTION ... o...o. .... ..... 6

2.2. PLANT DATA ACQUISITION

SYSTEM DESCRIPTION o ... ......... . ... 7

2.3. TRANSIENT DESCRIPTION ............. . 8

3. MODEL DESCRIPTION .... ... ................. 9

3.1. PRIMARY SYSTEM AND STEAM GENERATORS ... 10

3.2. SECONDARY SYSTEM ... ...... ... . . .... . 12

3.3. CONTROL SYSTEMS . . ..... ............. o 13

4. STEADY STATE CALCULATIONS ..................... 14

5. TRANSIENT CALCULATION AND COMPARISON

VERSUS ACTUAL DATA . . . ...... . ...... o. ...... 15

6. RUN STATISTICS ......... o..............o...... 17

7. CONCLUSIONS ......... ....... . . ....... o .... o 18

8. BIBLIOGRAPHY .... ... o....... .... .... ...... ..... 19

9. INDEX OF TABLES .. ........ .. ... ......... . ..... 20

10. INDEX OF FIGURES ...... . ... .......... ...... ... 27

page 2
ABSTRACT

The Consejo de Seguridad Nuclear (CSN) and the

Asociacion Nuclear Vandellos (ANV) have developed a model of

Vandellos II Nuclear Power Plant. The ANV 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 feedwater turbopump trip transient simulation. This transient ocurred actually in plant

on June 19, 1989.

page 3
EXECUTIVE SUMMARY

The 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/MOD2

through the ICAP project. Then, Vandellos II NPP colaborated

with the CSN simulating and analyzing two of the four

transients that 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.

- This transient causes the steam-dump to open, but

does not cause either the relief or the safety

valves to actuate, so that this allows analyzing

the steam.dump behavior.

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 (ANV) decided, at

the beginning the commercial operation, to promote efforts

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 simulation of FSAR design accidents by means

of a best estiamte model, in order to compare them

to the results obtained using conservative codes.

- The comparison of the FSAR design accidents to the

best estimate model.

- 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 II 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. This is the equipment used to record the

parameters needed to assess this case.

The availability of such a great number of signals has allowed the use of RELAP to check the control blocks

partial performances, specially the feedwater control block and

the rod control block (verifications carried out during the

load rejection from 75% to 65% case), and the steam-dump.

page 7
2.3 TRANSIENT DESCRIPTION

The test which is the subject of the current

simulation ocurred on June 19, 1989, in Vandellos II, being the

plant at the 99.2% power level.

The transient ocurred as a result of a maintenance

operation consisting in the realignement of the turbopumps

lubricating oil cooler. To realign it, an auxiliary three way

valve is needed, and there is a position of this valve in which

a pressure drop occurs. This pressure drop caused the main

feedwater turbopump trip.

This trip of a main feedwater turbopump caused the

runout in the other one, which has the capacity to supply the

85% of the total feedwater volume. In the same manner, this

trip triggered the turbine runback from 100 % to 70% at 200% /

min rate.

All the control systems in the plant were in

automatic mode.

page 8
3. MODEL DESCRIPTION

Figure 2 shows the nodalization used to simulated the

primary system of the plant. It consists of 117 volumes, 122

junctions, 78 heat structures and 155 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 9
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 the 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 heaters 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:

- Volume 580:

Downcommer

lower plenum

from lower core support forging to

lower core plate.

core

between internals core barrel

baffles, and other core by-pass

and

from upper core plate to mid loop

elevation.

from mid loop eleV

support assembly.

from upper support

internals flanges.

ation to upper

assembly to

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 10
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 valves control have

been simulated, but not the valves themselves. This allows

verifying that in this transient these valves do not open.

Talking about kinetics, the Doppler coefficient value

has been adjusted in such a way that for each rod position the

plant nuclear flux level is reached. This adjustement has not

been possible to be made with the Doppler coefficient design values, since RELAP5/MOD2 uses the punctual kinetics model, and

during this transient a rod position shift occurs that implies

a different core behavior depending on the axial height we

consider.

page 11
3.2. SECONDARY SYSTEM

In the secondary side, the three steam generators and

the three 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

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 the four groups. The MSR's,

ejectors, and turbopumps consumptions have also been simulated.

page 12
3.3. CONTROL SYSTEM

The primary basic controls can be groupped into four

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 settings during the

test have been used as setpoints for the model.

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 response accurately. This and

the steam flow are the main contributors to the RELAP5/MOD2

results and to the plant reponse mismatching in the main

feedwater turbopump trip transient simulation.

In this case, however, the rod control system has

been adjusted properly, since the turbine first stage impulse chamber pressure has been imposed as a boundary condition to

represent the turbine power evolution. In the load rejection

from 75% to 65% simulation case, this variable was not used

because it is not calculated by RELAP model, but in this case

it was expected to prove that using this procedure the control

rod behavior is improved preceptibly. This point suggests

extending the model downstream of the turbine valve.

page 13
4. INITIAL STEADY STATE CALCULATIONS

The new steady state has been reached starting from a

100% rated conditions steady state and modifying all the RELAP

variables reinicialization until reaching a similar steady

state to the plant one.

The main parameter values obtained with RELAP5/MOD2

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

Table III.

page 14
5. TRANSIENT CALCULATION AND COMPARISON VS ACTUAL DATA

The main purpose of this transient simulation

assessment is to check the model behavior, specially the

control system (the rod control, the pressurizer level and

pressure control and the steam-dump control) and the

thermohydraulic evolution of the plant main parameters.

The simulation of this transient has been carried out

starting from the initial steady state, imposing the feedwater

flow, pressure and temperature conditions, and reducing the

turbine flow. Another boundary condition that has been imposed

is the turbine first stage impulse chamber pressure, which will

be used by the rod control system as a turbine power reference.

This pressure is the plant actual pressure and is

supplied by means of a table.

The transient starts with the turbine flow decrease,

which causes a variation in the energy production and

evacuation balance of the primary side. Owing to this

variation, some changes in 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 mismatch and the average temperature evolution

(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 plant and RELAP final level discrepancy (70% and

74% approximately) is attributable to the plant power measures

error margin. A thermal balance at the end of the transient

has been carried out, which seems to indicate that RELAP

results are more trustworthy than plant data.

The cold leg, hot leg and average temperatures are

shown in figures 9, 10, and 11. In the beginning, the average

temperature increases because of the power produced by the

reactor and the steam generators 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.

page 15
Figure 12 has been included to show the primary delta

temperature evolution, as a significant indicator of the

primary power evolution.

This delta temperature has a close adjustement Whith

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.

Figures 19, 20 and 21 show the steam dump valves

opening, which operate in temperature mode. It can be observed

that the steam dump opening time in the model is longer than in

plant. Probably, attributable to the fact that the valves

capacity has been underestimated.

Since the steam dump valves flow measure is not

available in plant, it has not been possible to obtain a closer

adjustement; however, the total mass evacuated is considered to

be the correct one owing to the close adjust of the steam

generator pressure and the primary side average temperature.

The primary pressure evolution is similar to the

average temperature one, and is shown in figure 13. The same

occurs with the pressurizer level (figure 14), which is

modified esentially by the density variations in the primary

side.

It can be observed an initial pressure peak which is

higer in RELAP than in plant. This fact may be caused by the

spray efficiency, since the RELAP code does not allow

simulating actual physical phenomenon.

Figure 15 shows the feedwater flow evolution, which

in this case has been imposed as a boundary condition. This

figure shows that, after a sudden fall as a result of a

turbopump trip, the flows tends to recover by means of the

other turbopump, and it stabilizes at a different level because

of the turbine run back.

The steam generators pressure (fig.17) increases

initially because of the turbine valve closure and has a smooth

decrease later, because of the heat transmission balance

through the steam generator tubes, the evacuation through the

turbine and steam dump, and the feedwater contribution.

The steam generators level (fig.18) has a sudden fall

on starting the transient because of the steam binding, but it

recovers later owing to the steam and feedwater flows balance

(figures 15 and 16).

page 16
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 3.00,

which is smaller than the ratio of the load rejection from 75%

to 65% calculation for ICAP, due to the fact that during the

time delay which existed between both calculations the computer

capabilities were improved.

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

all the transient.

The run statistics are shown in Table V.

page 17
7. CONCLUSIONS

The control blocks models, which were individually

assessed in the load rejection transient from 75% to 65%, the

other case prepared for ICAP, have been used in this

calculation without having introduced any modification. This

means that a new validation of the models has been carried out,

which proves again the outstanding performance of these models.

In this transient, which with regard to the primary side is in fact a load rejection from 100% to 70%, 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 lead us

to conclude that the punctual kinetics is a conservative model

that has to be corrected for this kind of transients.

The evolution of most of the RELAP main variables in

this transient are in close agreement with plant data.

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

the differences are within the plant instrumentation error

margins.

The simulation of this transient has allowed

analyzing the steam dump behavior with the RELAP5/MOD2 model.

The opening and closure speeds of all the banks have been

reproduced with accuracy; however, some differences exist with

regard to the time the valves remain opened. Probably, this

fact is due to the valves capacity in the model. A closer

adjustement has not been possible since the plant steam flow

measurement includes the turbine and steam dump consumptions

and other services (MSR's, turbopumps,...).

Anyway, it has been considered that the value of the

total mass evacuated through the steam dump is correct, since

the steam generator pressure and the primary side average

temperature have been closely adjusted.

This model, after being assessed with the

calculations for ICAP:

- A trip from the 100% power level

- A load rejection from 100% to 50%

- A load rejection from 75% to 65%

- A feedwater turbopump trip

is considered to be a valuable tool for transient simulation.

page 18
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/MOD2 Code Manual

Project Information Package

Steam Generator thermohydraulic analysis

Copia oficial prueba eficacia calentadores y spray

Copia oficial prueba reactividad de barras

Datos sistema Adquisici6n de Datos de planta

page 19
9. INDEX OF TABLES

TABLE I

TABLE II

TABLE III

DESCRIPTION OF THE MAIN CHARACTERISTICS OF

VANDELLOS II NUCLEAR POWER PLANT

MAIN EVENTS THAT TOOK PLACE DURING THE

TRANSIENT

COMPARISON BETWEEN RELAP5/MOD2 VALUES AND

ACTUAL DATA FOR STEADY STATE

DESCRIPTION OF RELAP5/MOD2 VARIABLES

RUN STATISTICS

TABLE

TABLE

IV

V

page 20
TABLE I

MAIN CHARACTERISTICS OF VANDELLOS II NPP

- 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 TEMPERATURE AT NO LOAD ('K) .......... 564.8

- COOLANT AVERAGE TEMPERATURE AT 100% (-K) .... 582.3

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

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

- TOTAL TUBE LENGHT (m.) ....................... 98759

- INNER DIAMETER TUBES (m.) .................... 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

page 21
- PRIMARY VOLUME (m3) .......... 106 19

- PRESSURIZER VOLUME (m3) ...................... 39.65

- HEATING POWER OF THE HEATERS RODS (KW) ....... 1400

- MAXIMUM SPRAY FLOW (Kg/s) .................... 44.2

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

page 22
TABLE II

MAIN EVENTS

TIME

0.0 SEC.

3.0 SEC.

APROX. 400 SEC.

EVENT

MAIN FEEDWATER TURBOPUMP TRIP

TURBINE RUNBACK

TURBINE VALVE AT A NEW POSITION

REACHED NEW STEADY STATE OF 65 % OF

POWER

page 23
TABLE III

COMPARISON BETWEEN RELAP5/MOD2 VALUES AND ACTUAL DATA

VARIABLE

NUCLEAR POWER

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

(%)

('K)

("K)

("K)

("K)

(MPa)

(%)

(Kg/s)

(MPa)

(%)

RELAP5/MOD2

99.2

564.8

598.1

581.5

33.2

15.52

59.9

1527

6.69

50.6

3.25

PLANT

99.2 (1)

564.8

598.1

581.4

33.2

15.50

59.8

1533

6.70

50.4

3.27 (2)

(1) CALCULATED DATA

(2) DESIGN DATA

page 24
TABLE IV

DESCRIPTION OF RELAP5/MOD2 VARIABLES FIGURE

CNTRLVAR 340

CNTRLVAR 301

CNTRLVAR 328

CNTRLVAR 327

CNTRLVAR 330

CNTRLVAR 947

P 415090000

CNTRLVAR 350

MFLOWJ 325000000

MFLOWJ 600010000

P 600010000

CNTRLVAR 203

CNTRLVAR 970

CNTRLVAR 971

CNTRLVAR 972

ROD POSITION

NUCLEAR POWER (PERCENT)

TEMPERATURE AT THE COLD LEG

TEMPERATURE AT THE HOT LEG

AVERAGE TEMPERATURE

DELTA TEMPERATURE

PRESSURIZER PRESSURE

PRESSURIZER LEVEL

FEEDWATER MASS FLOW RATE

STEAM MASS FLOW RATE

STEAM GENERATOR PRESSURE

STEAM GENERATOR LEVEL (N.R.)

BANK 1 STEAM-DUMP DEMAND

BANK 2 STEAM-DUMP DEMAND

BANK 3 STEAM-DUMP DEMAND

7

8

9

10

11

12

13

14

15

16

17

18

19

20

21

page 25
TABLE V

RUN STATISTICS

COMPUTER

TRANSIENT TIME

CPU TIME

C (TOTAL NUMBER OF ACTIVES VOLUMES)

DT (TOTAL NUMBER OF TIME STEPS)

CPU * 1000

= 1.28---------- C * DT

CPU TIME / TRANSIENT TIME

IBM 3090

650 sec

1954 sec

117

13000

3.00

page 26
10. INDEX OF FIGURES

FIGURE 1. VANDELLOS II N.P.P. DIAGRAM

FIGURE 2. NODALIZATION OF C.N.VANDELLOS II

FIGURE 3. ROD CONTROL SYSTEM BLOCK DIAGRAM

FIGURE 4. PRESSURIZER PRESSURE AND LEVEL SYSTEM

FIGURE 5. TURBINE CONTROL AND STEAM-DUMP SYSTEMS

FIGURE 6. FEEDWATER CONTROL SYSTEM

FIGURE 7. ROD POSITION

FIGURE 8. NUCLEAR POWER %

FIGURE 9. TEMPERATURE AT THE COLD LEG

FIGURE 10. TEMPERATURE AT THE HOT LEG

FIGURE 11. AVERAGE TEMPERATURE

FIGURE 12. DELTA TEMPERATURE

FIGURE 13. PRESSURIZER PRESSURE

FIGURE 14. PRESSURIZER LEVEL

FIGURE 15. FEEDWATER MASS FLOW RATE

FIGURE 16. STEAM MASS FLOW RATE

FIGURE 17. STEAM GENERATOR PRESSURE

FIGURE 18. STEAM GENERATOR LEVEL (N.R.)

FIGURE 19. GR. 1 STEAM-DUMP DEMAND

FIGURE 20. GR. 2 STEAM-DUMP DEMAND

FIGURE 21. GR. 3 STEAM-DUMP DEMAND

page 27

F 71

VANDELLOS II PRIMARY AND MAIN ASSOCIATED SYSTEMS FIG. 1

_J
f- NTDPvCL TLBIE-1

8MCH

K NODALIZATION H

C.N. VANDELLOS II

FIGURE 2

7UOPV

REACTM VCSSEL

AUXUARY

rn= WAT

BUNCH

PUMP TWOP,•L

L I
[-

NUCLEAR POVER

301 = RKTPOW

TURBINE LOAD

NON LINEAR GAIN FIG. 3

VARIABLE

RODS SPEED PROGRAM

302 = MFLOWJ 600

MFLOWJ 102

334

ROD

ROD CONTROL

ROD REACTIVITY 342
r

ERROR

PRESSURIZER PRESSURE AND LEVEL CONTROL
r

FEEDVA/ER FLnOV

(1/3 L1V LEVEL &G.)

MAIN AND AUXILIARY FEEDI)ATER CONTROL FIG. 5
r

TURBINE AND STEAM DUMP CONTROL FIG. 6
FIG.7 ROD POSITION (BANK D)

210

200 -

190 -

180 -

170

160 -

150

140

130 -

120 -

110

T T T

. . -~-.-..---

-4-- - 4-

4 1 I -t + 1----

0 200 400 600

seconds

PLANT - RELAP
FIG.8 NUCLEAR POWER

100

98-

96

94

92-

90

88-

86-

84

82

80

78

76

74 -

72-

70 -

+ + -+ -______ -- I

•L f ~---t--------+----- -----i -- _________

WK

-- 1 K-- ~~---~- -t. _____

..........

600

. . . . . . . . . . 4 . . . . . . . . . . -

0 200 400

seconds

PLANT -- RELAP
IG.9 TEMPERATURE AT THE COLD LEG

570 -

569 - - .-

568 - ................. .... ........

566- -

566 -

565 -

564-

563 -

0 200 400 600

seconds

PLANT RELAP
601

600

599

598

597

596

595

594

593

592

591

590

589

588

FIG.1O TEMPERATURE AT THE HOT LEG

- .- I • T

i ! . . . ..... !- I ............... . .......... ...... ..... .. ......... 7 7 -. . .. .. . .. . . ... . . .. . ... . .. ... . ....... ... . .

X -i~~~~~~~~~ ....- ............ ......... I ..........4 I ... I... .... ......

X ..... -F.. .................. h . ....... 1............... .. ..... i.........

___'_ -. 1....!... = I .. ......... ..... ..........

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

_ _ _ -t !. 

0 200 400 600

seconds

PLANT RELAP
585 -

584 -

583 -

582 -

'581 -

580 -

579 -

578 -

577 -

576

FIG.1 1 PRIMARY AVERAGE TEMPERATURE

--. -b-.- . . .--- . . .--- . . ~ - -------

I I . 1....

... I ...... -.. - . . -.-... .. .. ... -....... .. .. ......... ... , .... . ..... . -. -. 1.-i... . . . .. ....

N t .. .... . .. .... .I ... . - . .

. ..........-.... . . . . . . . . . . .. i.... ........

4 + I I -,...-4- -_ __

0 200

PLANT

400 600

seconds - RELAP
FIG. 12 VESSEL DELTA TEMPERATURE

34 ,

33 _ .

_ I _ _

I I

50-

28 - 1 . ...

2 6 -" ... . ... .. . . . .

25 -2OO4O

0 200 400 300

seconds

PLANT - RELAP
FIG.13 PRESSURIZER PRESSURE

IR9 -t -.- -- I------

16.1 -

C')

oC

16

15.9

15.8

15.7

15.6

15.5

15.4

15.3

15.2

15.1

--F .- "~___ ____-- [--.-----.--2- _ - ________ ~.. ...---- ~-- _______ I ---.---.---.

I I _

~1~~~~ - ________

- ~1~~~_________ -1- ________________ -- ~~---.

-- 4 -- .1~ _________ // I

I I I I I

0 200

PLANT

400 600

seconds

RELAP
FIG.14 PRESSURIZER LEVEL

67-

II '

65- _ _

64--

63 -_ _

57 I

63O - __ __ __ _L*-- - .. ..........-...........

5-I

62 - 1 . . . . ~.. . . . .. • ..... .... .. ....

54 I

51- 1

0 200 400

seconds

600

-- PLANT RELAP
FIG.1 5 FEEDWATER MASS FLOW RATE

1.6

1.5 1- ____* *--...~.{-.-..-.....-- ..---.

1.4"

1.3 -

1.2-

U)

T\~1f1~-*------------- 

/ I.---.-I-----------...--l-..---------.--.-....- 

NX .~ ___

V

1.1

0.9- I I I -- - i - -i0 200 400 600

seconds

PLANT RELAP
FIG.1 6 STEAM MASS FLOW RATE

1.6

1.5 -

1.4.

..

i- ik -- 1-- --- +-I------ I

CIul

'NV 1.3 V\

1.2

1.1 -I 777ý

1 -t 4 4- 4 4

0 200 400 600

seconds

- PLANT P RELAP
FIG.17 STEAM GENERATOR PRESSURE

7.4

7.3

7.2

7.1

~...... 

0I

7

6.9 -

6.8 -

6.7-

6.6 1* I

0 200 400 600

seconds

PLANT RELAP
FIG.18 STEAM GENERATOR LEVEL

56

54

52

50

48

46

44

4'2

40

38

36

34

32

30

28

26

24

22

I i

[... .............. ......... F.--

t~-'PAAi

U- I ---- 4------------± J

t ..

A)-

~--.---.... 

11. _

LXI 7.L I7__Z_ V I

_- -I I iii l -- :

L..-

-+ 4. 4 4. .. 4..... .. 4. -

0 200 400 600

seconds

PLANT - RELAP
FIG.19 STEAM DUMP GR.1 VALVE POSITION

1

0.9 4

0.8 --

0.7-4

-- ---. I-.... I 0.6

0.5-

0.4

0.3

0.2-

0.1 -

0

K -____

............. ........... I 

-,-.--- -. 1I--- . .it.----------.I ....

*1* Y ¶ Y

0 200 400 600

seconds

PLANT RELAP
FIG.20 STEAM DUMP GFR.2 VALVE POSITIO",

0.9

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

9

0 200 400 600

seconds

PLANT - RELAP
FIG.21 STEAMI DUMP GR.3 VALVE POSITION

1 ~1*~* -,. r

0.9 +-----------

0.8

0.7

0.6

HI

0.5 -

0.4 -1--

0.3 -r- KA,

0.2 -

0.1 -

I'

0 -, II 4 I. ,t.

0 200 400 600

seconds

PLANT RELAP

NFRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12-89) (Aseloned by NRC. Add Vol.. &pp., Rev.. NRCM 1102. and Addendum Number, If eny.)

3201.3202 BIBLIOGRAPHIC DATA SHEET

Isfe i,,..ions on the ,eversel NUREG/IA-0110

2. TITLE AND SUBTITLE ICAP00219

Assessment of RELAP5/MOD2 Against a Main Feedwater Turbopump 3. DATE REPORT PUBLISHED

Trip Transient in the Vandell6s II Nuclear Power Plant MONTH I YEAR

December 1993

4. FIN OR GRANT NUMBER

L2245

5. AUTHORIS) 6. TYPE OF REPORT

Carlos Llopis (ANV), Julio Perez (CSN), Albert Casals (ANV), and

Rafael Mendizabal (CSN) Technical Report

7. PERIOD COVERED Dosclust Dta7l r 

8. PERFORMING ORGANIZATION - NAM E AND ADDR ESS (If NRC. provide Division. Office or Region. U.$. Nuclear Regulatory Commisuion. and mailing address.:lt contractor, pro vile nwa and mailing addmsy.

Asociaci6n Nuclear Vandell6s,

Consejo de Seguridad Nuclear,

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

c/Justo Dorado 1l.-28040-Madrid

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (Ii NRC. type 'Tome as above: if contractor. provide NRC Division, Office or Region. U.S. Nuclear Regulatorv Commission.

oad malling addnus5.1

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT f200 words of l/ul

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

Vandell6s II Nuclear Power Plant. The ANV 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 feedwater turbopump trip transient simulation. This transient occurred actually in plant

on June 19, 1989.

12. KEY WORDS/DESCRIPTORS IList words or phraes that will nist researcrher In locating the report.1 13. AVAI LABILITY STATEMENT

ICAP,, Tnlimitpd

Vandell6s IL, 14. SECURITY CLASSIFICATION

RELAP5, ITht Page

Transient Unclassified

Turbopump Trip (his Report,

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 33S 12401


Federal Recycling Program

NURtEG/IA-0110 ASSI-'SSMiLN I' U1. I{(LAI'I•/IUI2 A•iAIN.o IA MIAIN P LUWAIL, it 'UiUu'UUmI'

TRIP TRANSIENT IN THE VANDELLOS II NUCLEAR POWER PLANT

DECEN1Iil5I 1993

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

AlfIiA\I kJVA

 Ig-LAI

.)/Ak11U4U AllN61 A4 iItýIAIP V' i:LldA I 1111 I UIJSlJ0I'UJlVII' TRIP TRANSIE'NT IN THE V'ANDELLOS If NUCLEAR POWER P~LANT

OUL5101AL~( IYY.3

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300