NUREG/IA-0 107

- -. ICSP-V2-R50-R

International

Agreement Report

Assessment of RELAP5/MOD2

Against a Load Rejection From

100% to 50% Power in the

Vandellos 11 Nuclear Power Plant

Prepared by

C. Llopis/A.N.V.

R. Mend izabal, J. Perez/C.S.N.

Asociacion Nuclear de Vandello's (A.N.V.)

Travesera de les Corts

39-43, 08020 Madrid, Spain

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

Justo Dorado

11-28040 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-0 107 -

JCSP-V2-R50-R

*Intern ational

Agreement Report

Assessment of RELAP5/MOD2

Against a Load Rejection From

100% to 50% Power in the

Vandellos 11 Nuclear Power Plant

Prepared by

C. Llopis/A.N.V.

R. Mendizabal, J. PerezlC.S.N.

Asociacion Nuclear de Vandellos (A.N.V.)

Travesera de les Corts

39-43, 08020 Madrid, Spain

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

Justo Dorado

11-28040 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-Hlydraulic 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.
ABSTRACT.

An assessment of RELAP5/MOD2 cycle 36.04 against a

load rejection from 100% to 50% power in Vandell6s II NP? (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.

All major trends and phenomena are correctly

reproduced. The discrepancies between calculated nuclear power

and plant data are likely due to wrong values of the Doppler

coefficient, and In less degree to rod worth uncertainty.

iii

LIST OF CONTENTS

PACE

ABSTRACT.........................................i

LIST OF TABLES ................................... vi

LIST OF FIGURES .................................. vii

FOREWORD ....................................... ix

EXECUTIVE SUMMARY ................ i

1. INTRODUCTION..................................... o1

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

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

2.2 PLANT SIGNAL ACQUISITION SYSTEM

DESCRIPTION........................ o o ..... .... 2

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

3. CODE INPUT MODEL DESCRIPTION.......... o ........ o.. . .. 4

3.1 'PRIMARY SYSTEM.......................... o..5 ...

3.2 SECONDARY SYSTEM............ .o.............. 6

3.3 CONTROL SYSTEMS ................... o ooo ..... ... 7

4. STEADY STATE CALCULATION............ o............. 7

5. TRANSIENT RESULTS.......................... o ...... 8

5.1 BOUNDARY CONDITIONS ......................... 8

5.2 TRANSIENT RESULTS..................... ...... 9

6. RUN STATISTICS..... ................... o ........... 11

7. CONCLUSIONS..... ................................. 12

8. REFERENCES ... .................................... 13

V
L IST OF TABLES

PAC E

TAB3LE I MA IN CHARACTER I ST I CS OF VANDELIos I I PLANT ......15

TABLE I I : STEADY STATE VALUES..............16

TABLE I II SEQUENCE OF EVENTS..............17

TABLE IV: RUN STAT ISTICS 18

vi
LIST OF FIGURES.

PAC E

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

1

2

3

4

5

6

7

8

9

RELAP5 NOD ING D IAGRAM FOR VANDELLbsI11NPP.....

TUR13IN E VALVE POSI T ION.............

FEED WATER TEMPERATURE...........

FEED WATER PRESSURE............

REACTOR POWER................

CONTROL RODS POSITION...........

REACTIVITY COMPONENTS...........

PRIMARY TEMPERATURES............

PRIMARY AVERAGE TEMPERATURES.......

19

20

21

22

23

24

25

26

27

FIGURE 10:

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

FIGURE

11

12

13

14

15

16

17

18

19

20

21

22

STEAM Dump TEMPERATURE ERROR.......

ROD CONTROL TEMPERATURE ERROR.......

STEAM Dump DEMAND.............

CONTROL RODS VELOCITY...........

SECONDARY SIDE PRESSURE.....................

PRESSURIZER LIQUID LVL........

PRIMARY SIDE PRESSURE...........

STEAM GENERATOR LEVEL...........

FEED WATER MASS FLOW RATE........

STEAM MASS FLOW RATE (SG. OUTLET) ............

STEAM MASS FLOW RATE (TURB INE)......

CPU TIME..................

TIME STEP..................

28

29

30

31

32

33

34

35

36

37

38

39

40

Vii

FOREWORD

This report represents one of the assessment/application

calculations submitted in fu If ilment of the bilateral -

agreement for cooperation -in thermaihydraulic 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 (1CM') 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 organizations:

- EMPRESARIOS AGRUPADOS, S.A.

- Unidad E16ctrica (UNESA)

- Uni6n Iberoamericana de Tecnologia Elictrica (UITESA)

- Empresa Nacional. del Uranio (ENUSA)

- TECNATOM

- LOFT-ESPANRA

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

review group and is coordinated by the "Comit6 de Coordinaci6n".

This committee has approved the distribution of this document -

for ICAP purposes.

ix

EXECUTIVE SUMMARY

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

load rejection from 100 %' to 50% power in the Vandell6s 11

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 ESPAROLA (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; In this transient

all Important control systems took into action.

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; and comparison with plant

measurements.

Calculations were carried out using Cycle 36.04 of

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

the CSN.

The calculation results are in reasonable

agreement with the plant measurements.

All major trends and phenomena are correctly

reproduced. The discrepancies between calculated nuclear power

and plant data are likely due to wrong values of the Doppler

coefficient, and In less degree to rod worth uncertainty.

xi

1. INTRODUCTION

The results of an assessment of the RELAP5/MOD2

code against a load rejection 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 Vandell6s 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; and comparison with the

plant measurements.

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 load rejection test has been analyzed

using the TRAC-PF1/MOD1 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 HIDROELECTRICA

ESPAfA0LA (28%). It Is located in Tarragona, in the North-East of

Sp .ain. 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 HWt).

1
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 downcorner 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 SYSTEM 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.

For this reason, once: the nuclear plant tests had *finished,

Vandell6s II NP? 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
2.3 TRANSIENT DESCRIPTION

The- transient under study is a startup ,load

rejection from 100% to 50% power. It was conducted on February

th

27 , 1988.

Objectives of this test were to verify the ability

of the plant to accept a 50% 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 load rejection

programmed with the Digital Electro-Hydraulic system (D.E.H.)

which reduced mass steam flow with a rate of 200% per minute.

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 load rejection 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 primary temperatures, during

the early seconds of the transient.

As a result of the load rejection, the reference

temperature suddenly changed from full load to 50% load, and there

was a significant temperature 'error which produced the quick

opening of the steam,dump valves and control rods Insertion at the'.-.

maximum speed. The combined effect of both. systems drove the

primary average temperature- to the new reference value.

3
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 an early fall

of the downcomer liquid level. The corresponding control system

recovered the level to Its reference value (50% narrow range).

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.

There were not plant values to know the 'actual

response (dead times and movement velocity) of the steam dump

valves, and the valve positions were not recorded.

3. CODE INPUT MODEL DESCRIPTION

The plant model (Fig.1) 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. 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 appropriate to

the transient under study, which is basically symmetric. The

nodalization includes 118 hydrodynamic volumes, 123 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 (THDPVOL).

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 take place as the control rods are

going up or down through the core and the effect that this change

produces in reactivity coefficients.

4
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

Includes the reactor vessel, loops, steam

generator primary side, pumps and pressurizer.

The loop Is scaled-up 3:1. excluding the vessel

and pressurizer, which are full-scale. 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). RELAP5/MOD2 cannot account for

the change of this coefficient with control rods position. The

Doppler coefficient value and the rod worth were not well known.

Design values for beginning of life and all rods off were used.

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 r epresent the- heat

losses through the vessel walls.

5
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

motor torque have been triplicated.

The primary side of the steam generator has been

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

chambers. The U-tubes have 10 nodes, with increasing length in

the direction of flow, In order to reproduce in detail the

temperature profile and enhance the pr imary- 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.

6
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 steam header, 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 =36% of the full power steam

mass flow at nominal pressure. A Time Dependent Junction accounts

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.

- Pressurizer pressure control.

- Steam dump control.

- Steam generator level control.

The five groups have been simulated according to

the plant design [6]. The plant actual control setting values

during, the test have been uspj as setpoints., L N>

The CVCS charge was simulated by means of a VALVE

and a TMDPVOL. The discharge was represented by a ThDPJUN

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.

7
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 [10].

.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 [31.

- 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 [5], and then tuned to adjust pressure

losses or bypass flows. For instance, the core bypass mass flows

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.

8
Known shortcomings In the RELAP5/MOD2 heat

transfer correlations [2] forced 'to Increase the

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

the desired steady state.

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 [9]; 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. TRANSIENT RESULTS.

5.1 BOUNDARY CONDITIONS

The simulation started with the turbine valve

closure from 37.32% to 17.39% (of around 53%) In =15.5 seconds,

(Fig. 2). The valve position was Imposed In order to match the

turbine steam mass flow rate.

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.

Header measured temperature and pressure were

imposed as boundary conditions in the TMDPVOL representing the

main feedwater source (Fig. 3 and 4).

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.

9
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 control valve began to close at 0.

seconds. Thus the control rods started to be Inserted, the steam

flow decreased, and the steam line pressure rose.

Fig. 5 compares the calculated reactor power with

the measured neutronic flux. The agreement was good until =40.

seconds. Afterwards, the power was overpredicted, although the

rod movement was well reproduced until =80 seconds (Fig. 6). The

disagreement is likely due to wrong values of the Doppler

coefficient, and in less degree to rod worth uncertainty (Fig.

7). This fact biased the evolution of main calculated variables.

In the early seconds, both hot and cold leg

temperatures Increased (Fig. 8), due to the degradation of the

primary-to-secondary heat transfer. After =40 seconds both were

overpredicted. Fig. 9 compares the calculated primary average

temperature with the measured one. This Figure also includes the

compensated average temperature, and the reference temperature.

The large difference between them (Fig. 10 and 11) produced the

early steam dump valves opening (Fig. 12), and the control rods

Insertion at the maximum speed (Fig. 13).

The vapour generation In the steam generator

boiler decreased following the primary-to-secondary power. The

combined effect of vaporization, steam dump performance and steam.

flow to the turbine produced a maximum in secondary side pressure

at =48 seconds (Fig. 14). The position of this maximum 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.

10
The combined action of steam dump and control rods

brought the primary temperatures down to the 50% power values.

The steam dump valves were demanded to start closing in the

calculation 17 seconds later than in the plant. The demand of

full closure delayed 360 seconds, due to the temperature

overprediction.

Following the temperature trend, the calculated

pressurizer level was greater than the plant data from =40

seconds on (Fig. 15).

The primary pressure is shown in Fig. 16. The peak

was truncated due to the PORV and spray actuation. At long term

the pressure control system activated heaters at full power and

closed the spray valve trying to match the nominal value.

The steam generator level fell at the beginning of

the transient due to void collapse. Calculated level dropped

faster than measured one (Fig. 17). This may be attributed to the

mentioned mass default In the steam generator. Through the

whole transient the level was underestimated , but the calculated

one matched the reference level from -100 seconds on.

The Fig. 18 shows the feed water mass flow rate

that Is consistent with the calculated steam mass flow rate and

and steam generator level.

The steam mass flow rate at the steam generator

outlet (Fig. 19) was overpredicted. The turbine flow rate matched

the plant data (Fig. 20). So, the quoted discrepancy was due to a

excessive steam dump capacity.

* 6. 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.

11
Table IV shows the run statistics for the

steady-state run, and the transient run. Both in steady state and

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.

The CPU time to transient time ratio has been

around 42 . The grind time was 17.89 miliseconds

The CPU time and time step are plotted versus

transient time in Fig. 2i and 22, respectively.

7. CONCLUSIONS

The availability of a large number of plant

variables through the Signal Acquisition System has allowed the

performance of the present assessment exercise.

The calculation results are in reasonable

agreement with the plant measurements.

All major trends and phenomena are correctly

reproduced. The discrepancies between calculated nuclear power

and plant data are likely due to wrong values of the Doppler

coefficient, and In less degree to rod worth uncertainty.

12
8.REFERENCES.

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

Ransom et al. NtJREG/CR-4312. EGG-2396. Rev.i.

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-OZ. Cuaderno de

CUlculo". UITESA. Julio 1987.

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

Vandell6s-II, Junlo de 1985.

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

Idelchik. Hemisphere Publishing Corporation.

Second Edition. 1986.

6.- "Nota de C.Alculo de la Modelac16n de C.N.

Vandellc~s II mediante RELAP5/MOD2 ". A. Casals et

al. Marzo de 1991.

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

Procedimiento PAN-59, Rev. 1. Disparo de Planta

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

8.- "Assessment of TRAC-PFl/MOD1 against a load

rejection from 100% to 50% power In the Vandell6s

II Nuclear Power Plant". Prepared for ICAP-Spain.

A. Querol, R. de la Fuente, P. Herndn. UITESA.

Ref. Consejo de Seguridad Nuclear: ICSP-V2R100-T.

October 1990.

13
9.- "Capacidades y experiencia adquiridas en el

juiclo de and1isis realizados con los c6digos

termohidrdulicos RELAP5/MOD2 y TRAC-PF1/MOD1.2

J.M. Izquierdo et al. XVI Reuni6n Anual SNE.

Oviedo, Octubre de 1990.

10.- "Assessment of RELAP5/MOD2 against a 10% load

rejection transient from 75% steady state in the

Vandell6s II nuclear power plant". Prepared for

ICAP-Spain. C. Llopls, A. Casals (A.N.V.), J.

Perez, R. Mendizdbal (C.S.N.). Ref. Consejo de

Seguridad Nuclear: ICSP-V2R1O-R. 1991.

14
T AB LE I

MAIN CHARACTERISTICS OF VANDELLOS II PLANT.

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

Electrical Power (Me) .............................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 (in)..............................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. (m)............................106.19

Pressurizer Volume (m3) ..........................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.

15
TA B LE 11

STEADY STATE VALUES

PARAMETER MEASURED CALCULATED

PRIMARY SIDE

Core Power M% 99.1 100.69 C)

Mass Flow Rate (Kg/s) --- 14638.

RCP Speed (Rad/s) --- 158.6

RCP Head (MPa) --- 0.649

Hot Leg Temperature (K) 597.3 597.1

Cold Leg Temperature (K) 564.1 563.9

Average Temperature (K) 580.7 580.5

Delta T (W. 99.4 99.2

Pressurizer Pressure (MPa) 15.41 15.45 C)

Pressurizer Level M% 57.2 56.7 (

SECONDARY SIDE

SG Dome Pressure (MPa) --- 6.70

SG Outlet Pressure (HPa) 6.5 6.60

Collector Pressure (M~a) 6.35 6.57

Feedwater Mass Flow (Kg/s) 1542.9 1525.4

Steam Mass Flow (Kg/s) 1471.8 1528.5 C)

Feedwater Temperature (K) 494.1 494.0 C)

SG Level M% 50.5 50.2 ()

Recirculation Ratio - -- 2.29

* Average values.

C) Controlled or imposed parameters.

16
T A BLE III

SEQUENCE OF EVENTS.

EVENT TIME (SECONDS)

PLANT RELAPS/MOD2

LOAD REJECTION

TURBINE VALVE START TO CLOSE

PERMISSIVE C-7 ON

FIRST INSERTION OF CONTROL RODS

STEAM DUMP DEMAND SIGNALS:

OPENING BANK 1

OPENING BANK 2

OPENING BANK 3

OPENING BANK 4

CLOSING BANK 4

CLOSING BANK 3

CLOSING BANK 2

CLOSING BANK 1

SPRAY OPENING DEMAND

TURBINE VALVE STOP TO CLOSE

OPENING DEMAND OF PORV

CLOSING DEMAND OF PORV

SPRAY CLOSING DEMAND

SECOND INSERTION CONTROL RODS

PERMISSIVE C-7 OFF

FULL CLOSING DEMAND OF STEAM DUMP

0. 0

0.0

2. 0

0.0

0. 0

2.2-82.0

6. 5-9. 5

9.5-11.5

11.5-13.5

13.5-15.5

70. -78.

78. -87.

87.5-116.

116. -138

7.7

14. 1

23. 0

38.5

103. -127.

137. 5

1.5-88.

9.5-12.

11.5-14.

13.5-15. 5

15.5-16.5

87. 5-98. 5

98.5-175.5

175.5-297.

297. -498.

10.

15.5

23. 5

24. 0

62. 0

103.5-

498.

17
TA B LE IV

CALCULAT ION TT CS) CPU CS) TS CS) CPU / TT CN TSN GT (mS)

Steady State

Transient

KEY

TT

CPU

TS

CN

TSN

GT

228.4 9505.6

300. 12665.8

Transient Time

CPU Time

Maximum Time Step

Cells Number

Time Steps Number

Grind Time (= CPtJ/(CN x TSN)

0.05

0.05

41.6

42. 2

116

118

4570

6000

17.93

17.89
K

FNODALIZATION ~

L±N. VANDELLOS 11J

Eý-~J ., 421

420

PRCUMINAlC

413 (0) 1415

45() 1413

K 411 111 3411

413 (4) 1413

11411-11 411 II) 1413

iI-,I 411 (3) 14153

i.-

(0

OLL

SNCLAJOJ 402

Li' BRANCH BRANCH4

1460 (LIPs

93) 15, 460

~4cu~jo rim ~ ACCLOMAIIT

SW.0MLVDLt Sao fRAAJCI *6m204 SAFETY __j MAj6CIClI

It*L. X'0

o.WsPvI. faO6'VA. S(cco)oAvI A. 71S

VALVIS

705 4 10~L170{ 21 4 j ®

119 '7152

0-1

'0-

.2-2 2

7O 0 2 1 0 2 0

364,004ý 12017 2 I I's IUPC f I * 0 eoa $ 323()

-® k~wJ MAIN

FEED WA.T FEE WATER'' *4,

320(*(0

rT SHO.............

-I

LK -1 L'.~ r~

FIGURE 1 .RELAP5 NODING DIAGRAM FOR VANDELLOs 11 NPP
180.08 I I 1 PLANT 71

0

C3,

0.

0i

tLii

-J

~ii

1-4

4-

z

s.00

8.88

0 too 200 300

TIME (SECONDS)

L L FIGURE 2 : TUR131NE VALVE POSITION -1
r -I

T AG ALIN1 GY - PLANTr

STEHPF 32391

490.08

N'

CL

I.-

I.-

480.00O

470.00

460. ea

458.88

8 lee 288 388

TIME (SECONDS)

FIGURE 3 :FEED WATER TEMPERATURE --1
r -I

9.99

P COLECTOR AA POSBI - PLANT

AP 32391

X:

tLJ

en

en

L.J

0.-

ntJ

L&J

LLJ

u8.99

7.99

a too 289 399

TIME (SECONDS)

L FIGURE 4 :FEED WATER PRESSURE -1
r

188.00

~jPLANT

SCNTRLVAR 30t

L&J

0x

0i

-cc

(-)

So.308

0.06

a too 20e

TIME (SECONDS)

FIGURE 5 :REACTOR POWER

300

L -j
r

208.88

150.00

[)PLANT

A CNrRLVAR 340

W'

aLii

0,

z

m

0

'-4

I-

'-I

(I,

0

a100.00

50.88

0 fee 200 300

TIMlE (SECONDS)

L FIGURE 6 : CONTROL RODS POSITION -i
r, 1i

I 660

c3DOPPLER (66)

A MODERATOR DENSITY C56)

X CONTROL RODS (342)

CD RKREAC

IN)

Ln

JI.-

8.0

-I

lee 296 396

FIME (SECONDS)

L L ~FIGURE 7 :REACTI V ITY COMPONENTS -.1
680.00

112 A

0

T RA" FR RE - PLANT

T RA" CA RE - PLANT

CNTRLVAR 328

CNTRLVAR 327

0)

'I)

I.-

.4

1.1

aL.i

I0

ILl

-J

-J

C)

C..)

I0

575.88

550.88

a too 200 3800

TIME (SECONDS)

FIGURE 8 :PRIMARY TEMPERATURES
600.00

x

0n

T2 fIEDrA PLANT

TMEDIAX COMP PLANT

TREF COMP PLANT

CNTRLVAR 330

CNTRLVAR 334

CHTRLVAR 31S

-.23

w

-t

hi

aE

hi

I.-

S75. flA

556.86

a 100 200 300

TIME (SECONDS)

FIGURE 9 :PRIMARY AVERAGE TEMPERATURES
r

10.00

0.88

-10. 00

SAL TY-408 HPI

ACNTRLVAR 312

SCNTRLVAR 3e7

oCNTRLVAR 3e8

SCNYRIVAR 310

c:)

cw

tx

-2 .800

-38.00

-48.80

-58.09

8 too 200 300

TIME (SECONDS)

L FIGURE 10 : STEAM Dump TEMPERATURE ERROR -i
r

-I

5.88

0.08

SDEII ?OV BARRA (K)

ACNTRLYAR 338

I.-

-10.

a too 200 300

TIME (SECONDS)

FIGURE 11 :ROD CONTROL TEMPERATURE ERROR -

. i
188.88

x

x

GR I

GR 2

CR 3

CR 4

GR I

GR 2

GR 3

CR 4

PLANT

RELAPS/flOD2

88.98

0

w

L.J

WJ

0i

88.88

40.08

28.88

8.88

0 log 208 308

TIMIE (SECONDS)

FIGURE 12 : STEAM Dump DEMAND
A C3SVEL CNTRLVAR BAR SA 338GfSee

Cu) '-4

Cl)

'-4

()

V

-J

tAJ 25.08

9.98

a 199 20e 399

TIME (SECONDS)

FIGURE 13 : CONTROL RODS VELOCITY
8.88

19P1 YAP SAL GY - PLANT

A P 60001

L.J

C,)

LAJ

0.

6.00

0 S188 208 300

TIME (SECONDS)

FIGURE 14 : SECONDARY S IDE PRESSURE
r 

1

-I

80.00

13 PRZ PROT L0482 PLANT

A.CNTRLYAR 3Se

-j

I.]

Id

--I

NI

0~

42.08

9 tee 290 308

TINE (SECONDS)

FIGURE 15 : PRESSURIZER LIQUID LEVEL -i
r -I

17 .08

0Pt PRZ PROTEC P0489

A P 41599

(&J

tJJ

I-4

CL

Is.08

13.00

0 too 208 300

TIME (SECONDS)

L FIGURE 16 :PRIMARY SIDE PRESSURE -i
r -I

88.88

68.88

~jLI RE 6Y PROT - PLANT

SCHTRLVAR 283

(a)

Ln

-J

L.J

w

-S

(D

C,)

49.98

100 2018 388

TIME (SECONDS)

L L FIGURE 17 :STEAM GENERATOR LEVEL
r r -I

2400.09

1806.03

DFt AG AIIM GV - PLANT

A tFLOWJ 32509

C,,

I.-

0

Li..

C,,

C,,

L.

1200.09

6090 0

0.90

0 100 209 300

TIME (SECONDS)

L L 18 : FEED WATER MASS FLOW RATE -JFIGURE 

-i
r -I

2488.88

1898.88

C3Ft YAP SAL GY - PLANT

A IIFLOUJ 08881

-13 cIr

I-.i

01)

I.-

c,)

1289.88

608.88

8.88

a fee 280 308

TIME (SECONDS)

L. L 19 : ~FIGURE STEAM MASS FLOW RATE (SG. OUTLET) --.1
r -1

I so. e

P IER ETA TUR %

SMFLCUJ 71SO8 %

uJ

01

-J

uiA

I-C

P-C

so.6of

0.68

0 100 266 300

TIME (SECONDS)

L FIGURE 20 : STEAM MASS FLOW RATE (TURBINE)
r -I

isewees

t oeec. on

C3PUTIME 8

C,)

4-)

5666, 66

e.66e

0 tee 200

TIME (SECONDS)

FIGURE 21 : CPU T IME

300

1. i
-I

0 10

i I I LT CNTRLYAR 7

0 01

aC,)

"I

11

0.43S Lk,.Lrfl" -- a

0.08 I I I I I

0 108 200 300

TIME (SECONDS)

FIGURE 22 :TIME STEP L -i
NOC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

12-89) lAssigned by NRC. Add Vol.. Supp,* Rev.. NRCM 1102. DAA d Addendum Numbers, it any.) 3201. 3202 BIBLIOGRAPHIC DAASHEET

Isee'i -;fuuion on the reversel NUREG/IA-0107

2. TITLE AND SUBTITLE ICSP-V2-R50-R

Assessment of RELAP5IMOD2 Against a Load Rejection from 100% to 50% 3. DATE REPORT PUBLISHED

Power in the Vandell6s II Nuclear Power Plant MONTH "VEAR

June 1993

4. FINJ OR GRANT NUMBER

________________________________________________________ L2245

5. AUTHOR (S) 6. TYPE OF REPORT

Technical

C. Llopis/A.N.V. 7. PERIOD COVERED tinclusiee Dares

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

8. PERFORMING ORGAN IZATION - NAME AND ADODRESS (11t NRC. provide Division. Olfiecf r eglon. U.S. Nuclear Regulatory Commnimslon. and mailing address: it contractor, rvd

name. end mailing eddeml, odid

Asociacion Nuclear de Vandellos (A.N.V.) Consejo de Seguridad Nuclear (C.S.N.)

Travesera de les Corts Justo Dorado

39-43, 08020 Madrid, Spain 11-28040 Madrid, Spain

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (Ii NRC. type "Sjme as above~; It contractor, provide NRC Division. Otliceor Region. U.S. Nuclear Regulatory Commission.

and mailing address.)

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washiington, DC 20555.

10. SUPPLEMENTARY NOTES

11. ABSTRACT 1200 words or l/ss)

An assessment of RELAP5/MOD2 cycle 36.04 against a load rejection from 100% to 50% 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.

12. KEY WORDSDESCR:PTORS (List words or phrases that will assist researchers In locating the report.) 13. AVAILABILITY STATEMENT

ICAP, Vandell6s 1I, RELAP5, Unlimited

Transient, Load Rejection 14. SECURITY CLASSIFICATION

IThit Page)

Unclassified

fThis Reflors

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 12-891


Federal Recycling Program
I
NUREG/IA-0107 ASSESSMENT OF RELAP5/MOD2 AGAINST A LOAD REJECTION FROM

100% TO 50% POWER IN THE VANDELLOS II NUCLEAR POWER 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