NUREG/IA-0123

ICSP-AL-BOUT-R

International

Agreement Report

Application of Full Power

Blackout for C. N. Almaraz

with RELAP5/MOD2

Prepared by

A. L. Lechas

C. N. Almaraz I y II

c/Claudio Coello, 123

28006 - 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-0123

ICSP-AL-BOUT-R

International

* Agreement Report

Application of Full Power

Blackout for C. N. Almaraz

with RELAP5/MOD2

Prepared by

A. L Lechas

C. N. Almaraz I y II

c/Claudio Coello, 123

28006 - Madrid, Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

June 1993

Prepared as part of

The Agreement on Research Participation and Technical Exchange

under the International Thermal-Hydraulic Code Assessment

and Application Program (ICAP)

Published by

U.S. Nuclear Regulatory Commission
NOTICE

This report documents work performed under the sponsorship of the Consejo De

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

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

United States and Spain (Technical Exchange and Cooperation Agreement Between

the United States Nuclear Regulatory Commission and the Consejo De Seguridad

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

November 1985). Spain has consented to the publication of this report as a

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

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

Spain or any agency thereof, or any of their employees, makes any warranty,

expressed or implied, or assumes any legal liability of responsibility for

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

apparatus, product or process disclosed in this report, or represents that

its use by such third party would not infringe privately owned rights.
ABSTRACT.-

The analysis group of Almaraz Nuclear Power Plant has developed a model of the

plant with RELAP 5 / MOD. 2 / 36:04. This model is the result of the work-experience on the code RELAP 5 / MOD. 1 / that was the standar code during the

period 1.984/1.989. Different solutions were adopted in the network to adecuate the model to RELAP 5 MOD. 2 Computer Code.

This transient was selected for ICAP because it presents an experience with

the same transient calculated with RELAP 5 / MOD. i / CY 29 computer code.

The comparison between both analysis will be interesting.

i
EXECUTIVE SUMMARY.-

Almaraz Nuclear Power Plant are two units P.W.R. designed by Westinghouse.

Commercial operation started in April 1.981 the Unit I and in September 1.983

the Unit II.

Almaraz Nuclear Power Plant has a nominal reactor power of 2696 MWt and three

loops (each). Steam generators are typical U-tubes model D-3 with preheater

designed by Westinghouse. Reactor coolant pumps are type single stage, centrifugal model W-11011-Al (93-D) designed also by Westinghouse.

The total electrical power output is 930 MWe.

Almaraz nuclear power plant received the computer code Relap 5 / Mod. 2 /

36:04 through the ICAP project. In exchange Almaraz Nuclear Power Plant

should send to ICAP the results analysis selected for that purpose two transients.

The first one is the blackout from 100% rated power to be described in this

report.

The Almaraz nuclear power plant model, with Relap 5 / Mod. 2 / 36:04, consists

of 210 control volumes, 220 junctions, 37 heat structures and 189 control variables.

All calculations have been carried out on a computer C.D.C. Cyber 180/830.

The C.P.U. time versus real time event was 22.6.

ii
ICAP PROJECT

TABLE OF CONTENTS

Page

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

EXECUTIVE SUMMARY ....................................................... ii

TABLE OF CONTENTS ....................................... ............. iii

INDEX OF TABLES .....................................................

INDEX OF FIGURES ......... .............................................. vi

1. INTRODUCCION ...................................................... 1

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

2.1 PLANT DESCRIPTION ............................................. 2

2.2 TRANSIENT DESCRIPTION ......................................... 3

3. MODEL DESCRIPTION .................................................. . 4

3.1 Primary System ................................................ .4

3.1.1 Hot and Cold Legs Nodalization ......................... 4

3.1.2 Reactor Vessel Nodalization ........................... 6

3.1.2.1 Hot fuel rod condition considerations ......... 7

3.1.2.2 Reactor shutdown process conditions .......... 8

3.2 Secondary System .............................................. 8

3.2.1 Steam generator secondary side nodalization ............ 8

3.2.1.1 Rise chamber ................................... 8

3.2.1.2 Downcomer annulus ............................ 9

3.2.1.3 Separators ................................... 9

3.2.1.4 Steam generators head .......................... 10

3.2.1.5 Main steam line ................................. 10

3.2.1.6 Steam header ................................... 11

3.2.1.7 Auxiliary components ........................... 12

3.2.1.7.1 Feedwater systems .................... 12

3.2.1.7.2 Protection systems .................. 14

3.2.1.7.3 Turbine ............................. .14

3.3 Pressurizer and Surge Line Nodalization ........................ 14

3.3.1 Pressurizer Nodalization .............................. 15

3.3.2 Surge Line Nodalization ................................. 15

3.3.3 Spray System .......................................... 15

3.4 Controls ................................... 15

3.4.1 Incorporation of reactor trip ........................... 16

3.4.2 Incorporation of safety injection trips ............... 17

3.4.3 Incorporation of auxiliary feedwater flow initiation

for steam generators ................................. 18

iii
TABLE OF CONTENTS (2)

Page

3.5 Valves .. ...................................................... 19

3.5.1 Steam generator secondary side valves ................ 19

3.5.2 Pressurizer safety and relief valves ................... 21

3.5.2.1 Pressurizer valve hysterisis curves ......... 23

3.6 Positive and Negative Moderator Temperature Coefficient

Considerations ................................................ 24

3.7 Emergency Core Cooling System ................................. 24

4. STEADY STATE CALCULATION ............................................ 25

5. TRANSIENT CALCULATION ............................................... 25

5.1 Event Sequence ................................................ 25

5.2 Evolution of Reactivity During the Simulated Period ......... 26

5.3 Nuclear Power (Evolution throughout the transient) .......... 26

5.4 Evolution of System Pressure .................................. 26

5.5 Evolution of System Temperature ............................... 29

5.6 Evolution of Primary Mass Flow ................................. 30

5.7 Level Behaviour and Pressurizer Valve Opening ............... 30

5.8 Evolution of Secondary System Mass Flow ....................... 31

5.9 Evolution of Steam Generator Level ............................ 32

6. RUN STATISTICS ..................................................... 32

7. CONCLUSIONS ........................................................ 33

8. REFERENCES ......................................................... 33

iv
ICAP PROJECT

INDEX OF TABLES

Page

TABLE 1.- DESCRIPTION OF THE MAIN CHARACTERISTICS OF ALMARAZ I NUCLEAR

STATION ....................................................... 37

TABLE 2.- PROJECT REACTOR COOLANT SYSTEM PRESSURE SETPOINTS ............ 47

TABLE 3.- EVENT SEQUENCE FOR LOSS OF ALL OFF-SITE POWER AT 100% POWER

OPERATION ..................................................... 48

TABLE 4.- RUN STATISTICS ................................................ 49

TABLE 5.- STEADY STATE RESULTS AT 100% POWER ............................ 50

V
ICAP PROJECT

INDEX OF FIGURES

Page

FIGURE 1 HOT LEG MODEL 51

FIGURE 2 STEAM GENERATOR MODEL 52

FIGURE 3 COLD LEG MODEL 53

FIGURE 4 VESSEL MODEL. 54

FIGURE 5 MODEL OF STEAM GENERATOR VALVES 55

FIGURE 6 PRESSURIZER SAFETY AND RELIEF VALVES 56

FIGURE 7 PRESSURIZER SAFETY AND RELIEF VALVES SYSTEM 57

FIGURE 8 PRESSURIZER SAFETY AND RELIEF VALVES 58

FIGURE 9 PRESSURIZER SAFETY AND RELIEF VALVES 59

FIGURE 10 HISTERISIS CYCLE FOR PRESSURIZER SAFETY VALVES 60

FIGURE 11 HISTERISIS CYCLE FOR PRESSURIZER RELIEF VALVES 61

FIGURE 12 REACTIVITY CURVE (PCM/C) 62

FIGURE 13 REACTIVITY VERSUS TEMPERATURE CURVE C•) 63

FIGURE 14 REACTIVITY CURVE ($) 64

FIGURE 15 EMERGENCY CORE COOLING SYSTEM 65

FIGURE 16 C.P.U. TIME 66

FIGURE 17 REACTIVITY 67

FIGURE 18 REACTOR POWER 68

FIGURE 19 PRIMARY PRESSURE (HOT LEG) 69

FIGURE 20 SECONDARY PRESSURE (STEAM LINE) 70

FIGURE 21 PRIMARY PRESSURE (10. SECONDS) 71

FIGURE 22 RELIEF VALVE MASS FLOW 72

FIGURE 23 SECONDARY PRESSURE (DOWNCOMER) 73

FIGURE 24 PRIMARY TEMPERATURE (U-TUBES) 74

FIGURE 25 SECONDARY TEMPERATURE (RISE CHAMBER) 75

FIGURE 26 HOT CHANNEL TEMPERATURE 76

FIGURE 27 PRIMARY MASS FLOW (HOT LEG) 77

FIGURE 28 PRIMARY MASS FLOW (COLD LEG) 78

FIGURE 29 PUMP VELOCITY 79

FIGURE 30 PRESSURIZER LEVEL 80

FIGURE 31 MAIN FEEDWATER MASS FLOW 81

FIGURE 32 ISOLATION VALVE MASS FLOW 82

FIGURE 33 AUXILIARY FEEDWATER FLOW 83

FIGURE 34 STEAM GENERATOR LEVEL (SINGLE LOOP) 84

FIGURE 35 STEAM GENERATOR LEVEL (DOUBLE LOOP) 85

vi
FOREWORD

This report has been prepared by Central Nuclear de Almaraz in the framework of the ICAP-UNESA Project.

The report represents one of the application calculations

submitted in fulfilment of the bilateral agreement for cooperation in thermal hydraulic 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 Electrica (UNESA)

- Uni6n Iberoamericana de Tecnologia El~ctrica (UITESA)

- Empresa Nacional del Uranio (ENUSA)

- TECNATOM

- Empresarios Agrupados

- LOFT-ESPARA

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

Coordinaci6n". This committee has approved the distribution

of this document for ICAP purposes.

D-1/90-MPNV Viii

1, INTRODUCTION

Within the framework of the Spanish contribution to the I.C.A.P. project,

which is promoted by the Spanish Nuclear Safety Council (Consejo de Seguridad Nuclear) and in which participation by the Electrical Sector is led

by U.N.E.S.A., the different nuclear power plants, electrical utilities

and engineering companies mainly dedicated to development of Nuclear

Energy were invited to collaborate.

In this respect CENTRAL NUCLEAR DE ALMARAZ is one organization in Spain

dedicated to the nuclear-based generation of electrical energy. The company's two 930 MWe units are one of the main contributions to the country's installed electrical power.

Almaraz Nuclear Power Plant began to perform calculations using the Relap

5 Mod. 1 calculation Code in 1984, with a dual application: on the one

hand, due to the need for a Best Estimate Calculation Code for the calculation of accidents and transients, and on the other for the training of

operations personnel, an area in which the code has been widely used since 1984.

With respect to thermohydraulics, versions 19, 21, 25 and 29 of the above-mentioned Relap 5 mode 1 code have been used, the model utilized being

variable depending on the transient to be calculated, but generally

equipped with 210 control volumes.

Mode 2 of the Relap 5 calculation code was installed in the Almaraz NPP

computer in April 1988, this initiating knowledge of the code and

attempts to acquire the experience necessary for future replacement of

Relap 5 Mode 1 C and 29, although at present the standard calculation code at Almaraz continues to be mode 1.

1.
On the basis of the above, the first complete calculation of an accident

was undertaken. Our organization decided to dedicate this initial experience to normal blackout (loss of offsite power, Loop), as this sequence

presents a series of highly important characteristics that may be summed

up in the following two areas:

- Experience accumulated in previous analysis using Relap 5 Mode 1.

- The possibility of comparing results with an actual event that occurred

at Almaraz NPP in December 1981.

2. PLANT AND TRANSIE DESCRIPTION

2.1 PLANT DESCRIPTION

C.N. Almaraz I and II is a PWR nuclear power plant designed by Westinghouse of the USA. The plant is owned by a consortium of four

spanish utilities: COMPARIA SEVILLANA DE ELECTRICIDAD, S.A.; HIDROELECTRICA ESPANOLA, S.A.; IBERDUERO, S.A. and UNION ELECTRICA-FENOSA,

S.A.

Comertial operation started in April 1.981 the Unit I and in September 1.983 the Unit II.

C.N. Almaraz I and II has a nominal reactor power of 2.686 MW (thermal) each. The core contains 157 fuel elements, each one with 17x17

standard fuel rods. Steam generators are three, one per loop, the

model is D-3 designed by Westinghouse, with typical U-tubes and preheater.

The rated generator electrical output is 930 MW. Reactor coolant

pumps are type single stage, centrifugal, model W-11011-Al (93-D)

designed also by Westinghouse.

The unit modeled was the Unit I, the difference between Unit I and

Unit II is located in the vessel upflow versus downflow.

A more complete description is given in Table I.

2.
2.2 TRANSIENT DESCRIPTION

This transient is a normal Blackout (loss of external electrical power) at full power. It is a calculation of a very similar event

that took place on December 31, 1981, and classified as not Programmed, but is not possible to make a detailed comparison because

in the nuclear power plant, at this date, there was not very sophisticated computer data acquisiton system, as there is now.

This transient was characterized by a loss of forced primary coolant

flow. The condenser is not available because the condensate pumps

trip and there is not vacuum at the condenser.

The turbine valves close and the pressure increases at the secondary

side.

After 8 seconds the main feedwater is tripped and the auxiliary

feedwater started, 60 seconds after Blackout signal.

The rising pressure and temperature in the secondary side result in

a corresponding temperature and pressure increase on the primary side.

The pressurizer level increases to 7.9 meters with a maximum 7.5 seconds after the Blackout signal. The water level in the steam generators drops to a minimum of about 11 m after that.

The residual heat was calculated by the Relap 5 computer code and

the results were very close to the curve contained in A.N.S. 5.1.

The pressurizer spray was not taken into account due to the fact

that the results of the different analysis showed that the contribution to the transient results is negligible.

The parameters of the reactor coolant pump were apported by Westinghouse included the pump homologous curves.

3.
3. MODEL DESCRIPTION

During the nodalization process, previous experiences with the Relap 5

Mod. I computer code and the results obtained by different users were taken into consideration along, with EG&G remarks with respect to optimization of RELAP 5.

The plant was modeled by two loops: the first one (the double loop) containing the pressurizer, and the second one, one single loop. This model

was prepared for asimetrics calculations due to one single failure, for

instance a failure in one steam generator. The model with three loops

does not have special contributions.

3.1 Primary System

3.1.1. Hot and Cold Legs Nodalization

The final configuration (see Figure 1) for nodalization of

the hot legs was accomplished by means of 3 components which

divided each of the legs into 5 control volumes.

As regards correct simulation of the hydrodynamic model, the

above-mentioned control volumes were affected by their corresponding hydraulic diameter (diameter of one single leg).

The first component corresponding to the reactor vessel outlet nozzle was nodalized by means of a branch with a control

volume. The reason for nodalizing the nozzle by means of a

branch is that it is in this control volume that the high

pressure safety injection is connected (percentage injected

into the hot leg).

The straight section of the hot leg was nodalized by means of

a pipe with 3 control volumes connected to the curved section

of the hot leg by means of a branch with a single control volume.

4.
The steam generator inlet chambers were nodalized by means of

a single volume (SNGLVOL). This volume is in fact a double

volume for twin loops, attaining its real magnitude as a single volume for the case of the single loop.

The steam generator tubes were nodalized by means of a pipe

with 8 control volumes - four risers and four letdowns, (see

figure 2).

Heat transfer from the primary to the secondary side was

accomplished by means of heat plates simulating the entire

tube surface with respect to heat transfer.

Heat transfer in relation to the tube support plate was represented by means of two heat plates simulating heat generation across this component.

The steam generator outlet chambers were represented by a

single volume having the same geodesic head as the actual

steam generator inlet chamber. This was represented by a

single volume (SNGLVOL).

The section of the cold leg located between the outlet of the

steam generator and the main coolant pump was nodalized by a

single component (PIPE), which was divided into 5 control volumes.

The first three of these five control volumes sloped downwards, the fourth being horizontal and the fifth ascending.

The final control volume was connected to the main coolant

pump - nodalized by the component pump. The pump curves were

incorporated into the input data without using the Westinghouse pump curves included in RELAP 5 and the input date were

aported also by Westinghouse.

5.
The pump outlets were connected to the cold legs, each of

which was nodalized by means of 3 components, 2 pipes and 1

branch, representing 5 control volumes.

The objective underlying the incorporation of the branch was

to connect the high and low pressure emergency core cooling

systems and the accumulators (see Figure 3).

3.1.2. Reactor vessel nodalization

The core, downcomer and bypass reflector were divided by

means of volumes having the same length and geodesic head.

No heat structures were considered in the upper and lower

areas of .the core (influence of springs and plugs) (Figure 4).

The heat structures (fuel rods) corresponding to the active

part of the core are simulated by including the heat source

(U02 ) and the properties of U02, fuel-to-cladding gap and

zircaloy for heat transfer to the coolant.

The collection zone for water coming from the inlet nozzles

is represented by a single control volume which has the same

geodesic head as the zone for distribution of water to the

outlet nozzles forming part of the upper plenum.

The downcomer inlet zone is also represented by a single control volume. This volume has the same length and geodesic

head as the core outlet zone (lower parts of the upper plenum).

The downcomer itself is represented by 2 control volumes having the same length and geodesic head as the adjacent areas

of the core and bypass reflector, which are also represented

by 2 control volumes.

6.
The lower plenum is represented by one control volume.

(BRANCH with four JUNCTIONS).

The upper plenum is divided into 5 control volumes. The core

water collection zone was divided into different control volumes due to the fact that, otherwise, we would have two interconnected volumes, one excessively large and the other excessively small, for which reason the results corresponding

to this control volume would be insufficiently reliable.

The mass transfer through the guides columms upper internals

was not taken into account.

In order to achieve complete homogenization of the fluid, a

control volume (branch 223 and 224) has been derived from the

upper part of the upper plenum, which is also connected to

the single a double, outlet nozzles.

The reactor vessel head has been divided into 2 control volumes in order to assure a coherent configuration with respect

to the dimensions of the adjacent volumes.

3.1.2.1. Hot fuel rod condition considerations

In order to calculate heat loads for any of the fuel

rods, a truncated cosine distribution based on an

integral value of 177 W/cm was used.

In order to account for the maximum load conditions

of a centrally located rod, a heating structure was

calculated for this rod with a maximum heat coefficient of 410 W/cm axially orientated from the intermediate zone.

7.
3.1.2.2. Reactor shutdown process conditions

The reactor may be conditionally controlled by the

model by means of control logic processes.

This control may be time-dependent or depend on system variables. These variables may be assumed to be

instantaneous or delayed by means of disconnection

process result logic connections, depending on the

complex system conditions to which they refer (model

auto-control).

All questions (trips) may be applied in the reversible or irreversible mode. These possibilities are

to be applied depending on the plant operational mode and the problem to be dealt with.

3.2. Secondary System

3.2.1. Steam generator secondary side nodalization

The steam generators were nodalized taking into account their

different fundamental parts, (see figure 2).

These working zones include fundamentally the following:

- Rise chamber

- Downcomer annulus

- Separators

- Steam generator head

- Auxiliary components

3.2.1.1. Rise chamber

This chamber was nodalized by means of a component

(pipe) divided into 6 control volumes.

8.
The first four control volumes - as from the tube

support plate - have the same geodesic head as the

control volumes corresponding to the primary tubes.

The fifth control volume runs from the end of the

tube bundle to the inlet of the 12 risers (tubes

channelling the mixture of steam and water to the

steam separators). This fifth volume has the same

geodesic head as the first zone of the downcomer

annulus.

The sixth control volume corresponds to the 12 risers channelling the mixture to the separators.

These risers have been nodalized by means of a single control volume (the sixth) to which has been

assigned the hydraulic diameter corresponding to

each of the 12 tubes.

3.2.1.2 Downcomer annulus

The downcomer annulus has been nodalized by means of

the component "ANNULUS", which has been divided into

5 control volumes and whose geometric characteristics have been conmmented on above.

3.2.1.3 Separators

The model D-3 steam generator installed at Almaraz

NPP has 12 steam separators, and corresponding steam

dryers, located in the steam generator head.

The new separator model (SEPARATOR) incorporated in

RELAP 5 Mod. 2 contemplates inlet of a mixture of

water plus steam,: which is divided at the outlet and

sent in two directions: upwards, in the case of

steam with a small portion of water, and downwards,

in the case of water with steam.

9.
In this respect, the volume adopted for this component is indifferent and does not represent the real

conditions of the model D-3 steam generator, steam

separators and dryers; however, the results achieved

accurately represent the path followed by the fluid.

The way in which the separator is incorporated into

the RELAP 5 model does not correspond to the real

geodesic situation. This is due to the fact that

experience suggests that the most appropriate form

of nodalization is that observed in Figure 2.

3.2.1.4 Steam generator head

The steam generator heads have been nodalized by

means of two components - one volume (SNGLVOL) and

one branch (BRANCH) -, dividing this area into 2

control volumes connected to the steam generator via

the separator (SEPARATOR).

The upper part of the steam generator head is

connected to the main steam line and auxiliary components.

3.2.1.5 Main steam line

The main steam lines corresponding to each of the 3

steam generators installed at Almaraz NPP are not of

the same length; consequently, the steam path is

different in each of these lines.

The value used is the average length of loops 2 and

3 up to the steam header, this fictitious length

being subsequently nodalized as a twin loop.

The single loop has been nodalized using the exact

length of loop 1.

10.
With respect to this nodalization, four different

parts should be distinguished:

a) Nodalization from the steam generator head to

containment, nodalized by means of a "PIPE".

b) Nodalization from containment to the steam isolation valve. This zone has been nodalized by

means of a "BRANCH" in which are applied the

steam isolation valve, the relief valve and the 5

safety valves, all of which have been jointly nodalized by means of a single "VALVE".

Both the relief and steam isolation valves have

been nodalized with the components called "VALVE".

c) Nodalization from the steam isolation valve to

the steam header. This zone has been nodalized

by means of a "PIPE".

d) The four 24-inch steam lines channelling the

steam to the high pressure turbine have been nodalized by means of a control volume to which has

been assigned the corresponding hydraulic diameter.

This nodalization makes it possible to correctly

assembly the high steam line differential pressure

trip.

3.2.1.6 Steam header

The steam header has been nodalized by means of a

branch with 2 inputs, one for the twin loop steam

line and the other for the single loop.

11.
The 24-inch line has been joined to the steam header; as has been pointed out above, this line is also connected to the turbine by means of the control

and shutdown valves and the turbine bypass valves

leading to the condenser.

3.2.1.7 Auxiliary components

The steam generator auxiliary components may be subdivided into the feedwater system protection systems

and turbine.

A detailed description of each is given below.

3.2.1.7.1 Feedwater systems

Basically, Almaraz NPP possesses a main

feedwater system and an emergency (or auxiliary) feedwater system.

The main feedwater system has two steam

turbine-driven pumps for the three steam

generators. In order to save computer

time, the feedwater simulated at the inlet of each steam generator is 2/3 capacity of each pump, taking into consideration the losses experienced in the feedwater line from the pump to the inlet to

the steam generator.

Consequently, and with respect to main

feedwater, we have a hypothetical pump

feeding each steam generator. This pump

is determined by the hypothetical real

pump curves at the inlet to each steam

generator.

12.
Injection by this hypothetical pump is

given depending on steam generator level,

injecting rated flow (497.6 kg/s) at rated level (12.5 m) from the tube sheet.

The table containing the hypothetical

pump curve is given by a time-dependent

junction (TMDPJUN), and is governed by a

trip aimed at interrupting feedwater flow

whenever the event should make this necessary.

Fluid inlet conditions are given by a time-dependent volume (TMDPVOL) giving the

exact pressure and temperature at the inlet to the steam generator.

Emergency cooling injection (auxiliary

feedwater) has been performed analogously

to the above. This injection is provided

by 2 motor-driven pumps and 1 turbine-

-driven pump. With respect to these

pumps, and given their different characteristics, the methodology used is analogous to the above, the result being the

overall joint characteristics at the

steam generator inlet.

Pipe losses have been taken into account

in the curve for the pumps. A time-dependent junction (TMDPJUN) has been calculated (characteristics of the assumed

assembly at the steam generator inlet),

along with a time-dependent volume

(TMDPVOL) (for inlet conditions). (See

Figure 2.)

13.
3.2.1.7.2 Protection systems

The safety and relief valves have been

simulated by means of their corresponding

assemblies (VALVE), and the hypothetical

atmospheric steam dump by means of a time-dependent constant volume (TMDPVOL),

which gives us the pressure and temperature conditions outside containment.

Likewise, steam generator shutdown due to

dumping of steam to the condenser has

been simulated, represented by a time-dependent volume (TMDPVOL), along with the

actuation valve which was represented by

the component (VALVE).

3.2.1.7.3 Turbine

Downstream of the four 24-inch lines

branching off from the steam header is

the turbine with its upstream located

control and shutdown valves.

Only the shutdown valve has been represented nodalized by its component, VALVE. The connection is direct to the turbine, which is represented by a time-dependent volume (TMDPVOL) including turbine conditions.

3.3. Pressurizer and surge line nodalization

The pressurizer model nodalized is highly simplified. Computer system limitations have been a determining factor in this aspect; however, the model meets the requirements perfectly.

14.
3.3.1. Pressurizer nodalization

Nodalization of the pressurizer was carried out using three

components: two components BRANCHES and one component PIPE.

The two BRANCHES represent the upper and lower heads of the

pressurizer. The lower head (BRANCH 510) is joined to the

lower cylindrical end of the pressurizer, on one side (junction 2), and on the other to the surge line (junction 1).

The upper head (BRANCH 530) is joined to the upper part of

the cylindrical shell of the pressurizer.

The cylindrical part of the pressurizer was represented by

means of a "PIPE" with 5 control volumes.

The pressurizer has been provided with heat structures relating to heat transfer across its walls.

3.3.2 Surge line nodalization

The surge line was nodalized by means of a PIPE component

with 4 control volumes, connection to the twin loop hot let

being accomplished by means of a junction (SNGLJUN).

3.3.3 Spray System

The spray has not been taken into account because the RCPs

were stopped at the begining of the transient.

3.4 Controls

The calculation code RELAP 5 Mod. 2 offers enormous possibilities

with respect to plant controls. It is common practice, when studying an accident or transient, to determine beforehand when the

trip will occur and the causes for this trip.

15.
In order to accomplish a real model, all reactor trips were included

as weLl as all causes possibly leading to safety injection.

This fact is of great use as there are events for which the cause of

reactor trip may not be strictly determined. Evidently, this is not

particularly important from the point of view of transient analysis,

but very important as regards operation.

The safety and relief valves have been nodalized and provided with

opening or closing controls.

Also incorporated are the logics for auxiliary feedwater start-up

and main feedwater isolation. This was accomplished on the basis of

the same requirements governing the previous trip, i.e. incorporating all the causes that might give rise to these trips.

Tripping of the main coolant pumps have been taken into account; likewise, the model incorporates all causes of turbine trip. The only

trips not included are those due to electrical causes such as, for

example, low frequency or low voltage trips.

All controls referring to safety injection, actuation valves, etc.

have been incorporated in the problem.

Almaraz NPP has also been provided in this nodalized mode with controls in those places in which this has been considered appropriate,

both for control in the strictest sense - such as in the case of the

pressurizer or the steam generators - and for refuelling water storage tank level. The core and upper plenum have also been provided

with controls, although in this case it is naturally only possible

to speak of level in the case of a loss of coolant accident.

3.4.1 Incorporation of reactor trip

As has been pointed out above, all reactor trips have been

incorporated. These trips are as follows:

- Nuclear overpower trip

16.
- Primary high average temperature trip

- Low pressurizer pressure trip

- High pressurizer pressure trip

- High pressurizer level trip

- Low primary coolant flow trip

- Low steam generator level trip

- Safety injection trip

All of the above causes for trip have been incorporated in

the Almaraz NPP model.

These trips have been incorporated with non-exclusive logic;

in other words, if one occurs, the reactor trips; but on the

other hand, it is possible to see at which moment in time the

other trips would be produced, although as can be logically

supposed no actuation is possible as the order would have

been given previously.

As is already known, reactor trip implies direct tripping of

the turbine, interruption of the main feedwater flow, and actuation of the auxiliary feedwater to the steam generators

(due to tripping of the main feedwater turbine-driven pump).

All the above has been appropriately incorporated in order to

simulate real actuation in Almaraz NPP. Following generation

of the trip signal, the control and shutdown banks would be

inserted (a scram occurs), the main coolant pumps would trip

(in the case of blackout), and the sequence of events described above would occur.

3.4.2 Incorporation of safety injection trips

The trips giving rise to safety injection incorporated are as

follows:

17.
- Low-low pressurizer pressure

- High steam line differential pressure

- High mass flow in the main steam lines coincident with low

steam line pressure

- High mass flow in the main steam lines coincident with low

average primary temperature

- Safety injection actuation on high containment pressure has

not been contemplated

As described above, for reactor trips safety injection can be

initiated if any of the above-mentioned events occur. As has

been pointed out, if safety injection occurs without reactor

trip, this trip will take place on safety injection initiation.

3.4.3 Incorporation of auxiliary feedwater flow initiation for

steam generators

The auxiliary feedwater system to the steam generators may

actuate under normal operating conditions or as a result of

accidents or transients.

The possible causes for auxiliary feedwater system start-up

and shutdown are as follows:

- Safety injection

- Main feedwater pump trip

- Low steam generator level

- Normal operation from 0 to 30% power

The code has been provided with all logics and controls necessary for simulation of all these possibilities.

18.
Consequently, the Almaraz NPP model using RELAP 5 Mod. 2 code, would copy plant operation in the automatic mode.

3.5 Valves

In the calculation for blackout special attention has been given to

modelling the safety and relief valves, both those of the steam generators and the pressurizer, as a series of problems have been encountered which will be described below.

The version 004, Cy 36 of the Relap 5 Mod. 2 computer code - which

is currently the version used for Almaraz NPP - does not accurately

calculate the critical discharge flow. The calculated flow is much

higher than the real value. There are techniques that can be used

to correct this defect. The first of these being the experience Of

certain laboratories with respect to mass flow from the valve closure position to valve open position, assuming the steam to be a compressible gas, and calculating the critical discharge flow for the

valve area, assuming isoentropic evolution of the fluid via the

break. These data could be incorporated in a time-dependent junction, and then, by inserting a valve opening and closing, can be governed and the critical discharge flow can be controlled (see Figure

5a).

Optimization could be accomplished by incorporating an opening trip

in the time-dependent junction and then assuming zero mass flow on

closure trip. (See nodalization figure.)

3.5.1 Steam generator secondary valves

The steam generator secondary side valves have been nodalized

by the specific component '"ALVE". The relief valves can be

set from the control room, the setpoint established for operation with respect to station blackout being the value corresponding to operation at 100% power, which is the value

that the operators usually have inserted. Opening pressure

is 76.9 x 105 Nw/m 2 , closing pressure is 68.276 x 105

2

INw/r , and which flow is 52.08 kg/s.

19.
The safety valves actuate against the force applied by a

spring on the basis of static pressure. Closing pressure

will be identical to opening pressure, but when the valve is

open the dynamic pressure term will be other than zero. Consequently, the static closure pressure will be lower. These

calculations have been carried out and have given the following results:

First safety valve:

Opening pressure 80.55 x 105 Nw/m2

Closing pressure 74.21 x 105 Nw/m2

Maximum discharge flow 41.19 kg/s

Second safety valve:

Opening pressure 81.93 x 165 Nw/m2

Closing pressure 75.28 x 105 Nw/m2

Maximum discharge flow 41.89 kg/s

Third safety valve:

Opening pressure 83.00 x 105 Nw/rm2

Closing pressure 76.2 x 105 Nw/m2

Maximum discharge flow 42.41 kg/s

Fourth safety valve:

Opening pressure 83.7 x 105 Nw/m2

Closing pressure 76.76 x 105 Nw/m2

Maximum discharge flow 42.75 kg/s

Fifth safety valve:

Opening pressure 84.28 x 105 Nw/m2

Closing pressure 77.24 x 105 Nw/m2

Maximum discharge flow 43.08 kg/s

20.
Nodalization of the loop is different inasmuch as the mass

flow in the model loop (which is double) is clearly double.

3.5.2 Pressurizer safety and relief valves

The pressurizer safety and relief valves are located circumferentially on the same parallel, i.e. parallel 440 13'

from the pole.

These valves are off-set 660 one from the other (see Figure

6a).

The nozzles of both the safety and the relief valves have a

diameter of 6 inches.

According to the final safety analysis:

There are 3 safety valves and 2 relief valves discharging to

a common header.

a

As can be appreciated in Figure 7 , the different safety

valves form a loop in order to assure the hydraulic seal and

maintain forces of reaction on the nozzles located downstream.

Let us look at each of the systems in more detail.

Table II shows all the pressurizer setpoints in great detail. For this particular accident, it is extremely

important that the discharge of both the pressurizer safety

and relief valves is correctly modelled, as on the basis of

the discharged mass it is possible to evaluate if the

pressurizer relief tank rupture disc has broken, and if there

is a discharge to the containment atmosphere, which should

certainly not be the case but, however, should be demonstrated. What is expected of the calculation performed is that

the relief valve opens for a very short time, or not at all,

although it is not expected that there should be any discharge across the safety valves.

21.
As regards the relief valves, the configuration for Almaraz

NPP is that shown in Figure 8.

There are 2 relief valves, their relief capacity being:

(At 165 kg/cm )

95254 kg/h ==> 26.5 kg/s each

with the fluid being saturated steam.

There are 3 safety valves, their maximum capacity being:

172364 kg/h ==> 47.88 kg/s each

Consequently, the relief valves have a fictitious capacity of

52.92 kg/s.

The 3 safety valves have a fictitious capacity of 143.636 kg/s

The opening pressure in Nw/m2 would be as follows:

PA = 171.2 Nw/mr2

and the closing pressure will be

P = 170.2 Nw/mr2

c

In this respect we believe that it would be more than acceptable to have both opening and closing occurring at the same

pressure, which would imply a saving on control volumes.

In this case, the definitive nodalization would be as shown

in Figure 9.

The conditions in the pressurizer relief tank could be incorporated, although this aspect is of little importance due to

the short transient period to be calculated.

Relief valve opening pressure is 162.03 x 105 Nw/m2 and

the closing pressure is 160.92 x 105 Nw/m2

22.
3.5.2.1 Pressurizer valve hysterisis curves

In order to provide a real evolution for mass flow

across the pressurizer safety and relief valves,

curves have been defined to show the typical evolution of mass flow on valve opening. This evolution

does not correspond to any real event observed at

Almaraz NPP, as no experience is available with the

above mentioned valves, but has been based on the

experiences of the RESEARCH PROJECT V102-9 for September 1981, which gave rise to the document "FLOW

VISUALIZATION TESTS AND ANALYSIS OF SAFETY AND PRESSURE RELIEF VALVES", which was prepared by EPRI.

This document includes experiences with safety and

relief valves analogous to those installed at Almaraz NPP.

The above mentioned hysterisis curves have been calculated by extrapolating the results of the above

mentioned document, and are represented in Figs. 10

and 11. As can be appreciated, on both curves the

pressure increase following opening to the outlet

mass flow is very small as at this moment the velocity is still very low. Subsequently, and as velocity increases, mass flow increases up to the critical discharge flow value. As can be seen in Figs.

10 and 11, this evaluation corresponds only to opening, as closure is considered to be instantaneous

because of the spring.

23.
3.6 Positive and Negative Moderator TemDerature Coefficient Considerations

It has been considered appropriate to carry out the calculation

using two different moderator temperature coefficients. The first

of these is for the initial Almaraz 1 cycle which corresponded to a

negative moderator temperature coefficient, the second calculation

corresponding to cycle 5 of Almaraz 1, which gave a slightly positive reactivity coefficient at the beginning of life.

As in previous cases, the calculations carried out are divided into

two parts. One zone is. that given by the Fuel Designer for cycle 5,

which is shown in Figure 12. Figure 13 shows the corresponding calculations and conversion from degrees centigrades to degrees Kelvin,

while Figure 14 shows the definitive form used as input to the code,

with reactivity in dollars depending on the density of the moderator

as requested by the code.

Following performance of these calculations it was concluded that

repetition would be unnecessary as the moderator temperature coefficient is only very slightly positive, and the calculations would be,

in a sense, repetitive.

3.7 Emergency Core Cooling System

The emergency core cooling system installed in Almaraz NPP is made

up as indicated in Figure 15. This figure, which is self-explanatory, shows the system connected to the emergency bus.

24.
4. STEADY STATE CALCULATION

The final objective is to get a steady-state condition of normal operation of the plant at full power, from which the transient calculation can

be initiated.

A stable state was obtained with the "steady-state" option of RELAP 5/MOD

2 code from the input data.

Particular conditions could be defined for this purpose having a stabilizing effect on the sequence of the steady-state calculation, like a

TDPVOL at the head pressurizer or other techniques to acelerate the

Steady State calculation. Those techniques have not been used in this

case.

Reactivity and reactor kinetics feedback due to moderator density, moderator temperature and fuel temperature are specified as zero for the

steady state calculation in order to maintain the power at a constant level.

A stable condition for the model has been reached after 186.45 seconds of

calculation. RELAP 5/MOD 2 code stops the steady state calculation once

the stability has been accepted by its internal checking procedure. The

main results of the calculated values versus operating values are represented in Table V. We have some discrepancies with regard to the heat

transfer coefficient in the Steam Generator (Secondary Side), the mass

inventory and the isentropic coefficient in valves discharges. Justification of discrepancies are included in the references 20 and 21.

5. TRANSIENT CALCULATION

5.1 Event Sequence

The sequence of events occurring during blackout can be seen in Table III and associated text, and requires no further explanation.

25.
5.2 Evolution of Reactivity During the Simulated Period

On the basis of the hypothesis developed previously, attempts have

been made to carry out these calculations using both positive and

negative moderator reactivity coefficient. The positive coefficient

hypothesis had been calculated for cycle 5 of Almaraz I. On the basis of the results - which showed a very small difference - it was

decided not to perform both calculations, as the results would necessarily have been very similar.

Figure 17 shows the evolution of core reactivity. For study, this

curve can be divided into two parts. The first arises as a result

of the insertion of negative reactivity due to insertion of the control and shutdown banks. The second corresponds to reactivity insertion as a result of variation of the moderator density (cooldown).

5.3 Nuclear Power (Evolution throughout the transient)

As regards nuclear power during evolution of a blackout event, total

nuclear power has been represented. Graph 18 shows total reactor

power.

The above-mentioned figure shows the delay in the low primary mass

flow signal as well as the lag between the electrical signal and the

mechanical signal. As can be appreciated in the graph, initiation

of rod drop occurs 2.3 seconds into the accident.

5.4 Evolution of System Pressure

Pressure is transmitted throughout the entire circuit at the speed

of sound. Figures 19 and 20 show the evolution of the pressure on

the primary and secondary sides of both steam generators. This evolution is as follows: as a result of the normal blackout (LOOP) and

closure of the turbine shutdown valves there is a pressure increase

on the secondary side resulting from the fact that the dynamic pressure term in the equation showing quantity of movement is transformed into static pressure, with the additional aggravation that the

turbine bypass to the condenser is not available.

26.
Consequently, the pressure increases up to the setpoint of the safety and, relief valves, which are the valves opening in. the case of

Almaraz Nuclear Power Plant for this accident. Once a certain quantity of steam has been dumped the safety valves begin to close while

the relief valves open in order to maintain the corresponding setpoint pressure, which is somewhat higher than the zero load pressure. This evolution can be seen in Figures 19 and 20. As regards

primary pressure, the evolution is reflected in Figure 19 which

shows a pressure increase at the beginning of the transient, and a

subsequent decrease.

Let us now analyze the reason for this increase. When blackout

occurs there is a temperature increase in the hot channels because

of the fuel, this being due to the fact that the trip of the three

main pumps causes a rapid increase in fluid temperature in the core. This, in turn, applies a reduction in the fluid density, for

which reason the pressurizer level increases, compressing the bubble

and increasing pressure. This also causes steam to be condensed into liquid; however, the pressure increase is faster, and eventually

exceeds the relief valve setpoint, the valve opening for 6.2 seconds, and then closing. The valve does not re-open throughout the

entire transient. Figure 21 shows a detail of the evolution of

pressurizer pressure and how this parameter exceeds the relief valve

setpoint. It should be pointed out that in this figure the units

2 2

are not kg/cm but the international Nw/rm. Figure 22 shows

mass flow across the pressurizer relief valve, along with the

corresponding setpoint and valve opening. Subsequently, primary

pressure decreases with reactor shutdown and rod insertion. When

the primary pressure decreases the third phenomenon occurring in

this evolution takes place, this being the way in which primary pressure follows secondary pressure.

This is due to the fact that the secondary side pressure is high and

prevents faster cooldown of the primary, the secondary side pressure

thus leading the primary pressure. There is a fourth phenomenon,

i.e. shutdown of the main coolant pumps, which makes residual heat

removal more difficult. Consequently, during the second part of the

transient the core must be cooled by natural circulation.

27.
Having said the above, it would be advisable to present a systematic

synopsis of the phenomena occurring, and the reasons.

From 0 -- > 6 seconds

Pressure increase

From 6 -- > 12 seconds

Raoid pressure decrease (

Because of shutdown of the main coolant pumps, the core temperature increases, this implying a reduction in

density, an increase in level and pressure, and opening of the relief valve

The control rods are fully inserted

and the reactor cools down as a result

of the lack of fission events, the density increases and pressure decreases;

there is a drop in pressurizer level

From 12 -- > 74 seconds

Gradual pressure reduction

The main pumps are coasting down but

are still pumping some 800 kg/s; pri-

< mary pressure follows secondary pressure

From 74 -> 100 seconds I

Even more gradual pressure <

decrease I

Mass flow is decreasing; the process of

heat transfer to the secondary is masked by the low coolant flow. The pressure decreases more gradually

Figure 23 shows the evolution of the pressure in the secondary lines

reflecting the increase in pressure due to closure of the turbine

shutdown valves and the non-availability of the turbine bypass system to the condenser. Subsequently, the safety and relief valves

open, the pressure decreases, and 60 seconds after blackout auxiliary feedwater flow is initiated to the steam generators. As this is

cold water, the overall density increases and the pressure drop is

more rapid, subsequently stabilizing during the steaming process.

If feedwater is colder, the complete shutdown process will be completed sometime earlier.

28.
5.5 Evolution of System Temperature

The phenomenon of secondary primary temperature leading or following

- which has been studied in the previous paragraph - can be observed

in Figures 24 and 25, which show all the phenomena described above.

The temperature on the secondary side follows pressure, as we are

working under saturation conditions. The figure also shows valve

opening, the ingress of auxiliary feedwater to the steam generators

(second 60), and the events described above.

The phenomena occurring in the primary have been described above and

are reflected in Figure 26. This figure shows the increase in temperature in the channel as a result of the pumps tripping, and the

subsequent availability of less mass coolant flow. Subsequently, on

reactor trip, the coolant cools down and the temperature in the active core channels decreases. Also represented is the evolution of

the temperature in the upper primary side zone of the steam generator tube (see Figure 24). It can be seen that the temperature peak

for this parameter lags behind the peak value in the core (see Figure 26), this being a result of the fact that the temperature related

information is carried by the fluid itself. As can be observed, in

this zone we have no secondary lead as the peak value occurs before

the peak temperature on the secondary side. Consequently, Figure 26

shows that this is a primary side phenomenon.

It is also interesting to show as from what moment these lead phenomena occur. Close observation of Figure 24 shows how there is a

temperature decrease on injection of auxiliary feedwater. This temperature decrease is also felt by the primary, although the core is

practically unaffected by the phenomenon (see Figure 26) as the information is carried by the fluid itself and the quantity of this

fluid decreases less because of the main coolant pump trip.

29.
5.6 Evolution of Mass Flow

The evolution of mass flow during the transient has been represented

by three figures. The first of these figures corresponds to the

evolution in the hot leg; the second shows the evolution in the cold

leg, and the third corresponds to the speed of the main pumps and

shows the evolution for pump shutdown.

Figure 27 shows the evolution of mass flow in the hot leg. As can

be seen in this figure, 100 seconds after blackout the mass flow

reaches a value of 582 kg/s.

Figure 28 shows the evolution of the mass flow in the cold leg, the

value in this case being the same.

Figure 29 shows the evolution of the speed of the pumps, which go

from 153 rad/s - the normal operation speed - to 20 rad/s 100 seconds into the transient. Stabilization would be achieved when natural circulation occurs under stable conditions. In this respect,

it should be remembered that natural circulation, according to calculations carried out by C.N. Almaraz and corroborated by Westinghouse, is approximately 200 kg/s.

5.7 Level Behaviour and Opening of Pressurizer Valves

Point 5.4 has already provided sufficient information on the phenomena occurring in the primary circuit with respect to pressure and

its interaction with level. This point will include a description

of the processes taking place. As a result of station blackout, the

main coolant pumps trip, the circuit instantly feels the effects of

the loss of mass flow. As a result of this event, the fluid in the

core heats up. The process of temperature increase is one of reduction of density, i.e. the volume that before was X will now be X+Y;

in other words, there is expansion. This expansion is felt by the

entire circuit, in which the pressurizer acts as an expansion chamber.

30.
Consequently, pressurizer level increases. This increase in level

causes the bubble in the pressurizer to contract, increasing pressure and consequently the saturation temperature limit, which gives

rise to condensation. Subsequently, due to insertion of the control

and shutdown banks, the core begins to cooldown. For both this reason and because of the steam generators (refer to point 5.4), the

primary fluid contracts and steam generator level decreases with time into the transient (see Figure 30).

The pressure described above eventually exceeds the relief valve

setpoint, producing valve opening at the mass flow forecasted by

Westinghouse. Relief valve opening occurs 1.8 seconds after station

blackout, the valves then closing 8 seconds into the event.

5.8 Evolution of Mass Flow in the Secondary

Three graphs have been used to represent the evolution of the mass

flow in different areas of the secondary side, in the steam generator. This evolution can be seen in graphs 31, 32 and 33.

Figure 31 shows the interruption of main feedwater supply to the

steam generator. As can be seen, the loops start with different

mass flows but the shut-off time is the same. Evolution of steam

flow can also be seen in the figure 32. Given the closure of the

shutdown valves and the non-availability of the turbine bypass to

the condenser, there is a pressure increase which increases the saturation temperature limit, this meaning a reduction in the generation of steam. Subsequently, the relief valves open and the mass

steam flow increases as a result of the reduction in pressure, and

consequently saturation temperature limit. Opening of the safety

valves can also be seen in the evolution of the mass steam flow. As

can be seen, closure of these valves causes steam flow to fall to

zero; also shown is the subsequent repeated opening and closing of

the steam generator relief valves. Initial supply of auxiliary

feedwater to the steam generators can be observed at second 60 into

the transient in both graphs, (see figure 33).

31.
5.9 Evolution of Steam Generator Level

Figures 34 and 35 show the evolution of the level versus transient

time. As a result of the pressure increase experienced by the steam

generator secondary side, for the reasons already pointed out above,

level collapses as a result of pressure increase, shut-off of the

auxiliary feedwater also contributing to this effect. Opening of

the steam generator safety and relief valves causes the slope of the

level curve to change because of the depressurization caused by opening of these valves. After second 20, level continues to drop because main feedwater flow has been cut off and the valves continue

to release steam to the atmosphere. Apart from this, pressure in

the steam generator is still very high and the level attempts to

find an equilibrium position. At second 60. into the transient, auxiliary feedwater injection into the steam generators begins, consequently, level is recuperated and attempts to achieve an equilibrium

value corresponding to the pressure determined by the steam generator.

6, RUN STATISTICS

Figure 16 shows the time taken by the computer versus the real time corresponding to the transient. This calculation was carried out using a

CDC cyber 830 computer which Almaraz NPP has installed in his Headquarter

Central Office at Madrid. The operating system used is the NOS, Release

2.5. The editor is the "FULL SCREEN EDITOR". As can be observed, the

relationship existing between the CPU time and real transient time is in

the ratio 3098/100; in other words, a time of 30.9 : 1. Although this

calculation time is not excessive, it can be observed that it is still

far from attaining a simulation time, i.e. a ratio of 1 : 1.

In table IV a typical run statistics has been summarized for blackout

analysis. The table includes the ICAP required number that was calculated based on the transient time, the total number of active volumes in

the model (time dependent volumes were not considered), the total number

of time steps and the total C.P.U. time.

32.
7. CONCLUSIONS

The study of the behaviour of Almaraz NPP in response to station blackout

does not show any increase in fuel temperature, with only the cladding

being somewhat affected by.the heat-up in the primary system.

As regards the steam generators, the relief valves and 3 safety valves

per steam generator open.

Therefore, if any event of these characteristics were to occur the plant

could be re-started without any need for revision.

Consequently, the plant is capable of withstanding an accident of these

characteristics without sustaining any damage.

8. REFERCES

I - Relap 5/Mod. 1 Code Manual Volume 1

System Models and Numerical Methods

Idaho National Engineering Laboratory, September 1981.

2 - Relap 5/Mod. I Code Manual Volume 2

Users Guide and Input Requirements

Idaho National Engineering Laboratory, September 1981.

3 - Efficiency of the Emergency Core Cooling System of the PWR C.N. Trilo I (1040 MWe) Medium and Small Leaks Nodalization, Steady State

Initial Solution KWU - Work Report

Hr. Dr. Kochskamper, R. Martinez, A. L6pez Lechas

Erlangen 20.9.83 R 15/83/E2190

33.
4 - Efficiency of the Emergency Core Cooling System of the PWR C.N. Trilio I (1040 MWe) Medium and Small Leaks

Code - Specific Boundary Conditions for Accident Analysis (Relap 5)

KWU - Work Report R-153

Hr. Dr. Kochskamper, A. L6pez Lechas, R. Martinez

Erlangen 20.9.83

- Efficiency of the Emergency Core Cooling System of a 1040 MW PWR

C.N. Trillo I Medium and Small Leaks

2

39.6 cm Leak between Reactor Coolant Pump and Reactor Pressure

Vessel KWU - Work Report R-153

A. L6pez Lechas.

6- Efficiency of the Emergency Core Cooling System of a 1040 MW PWR

C.N. Trillo I Medium and Small Leaks

475 cm2 Leak between Reactor Pressure Vessel and Steam Generator

(Hot Leg)

KWU - Work Report R-153

A. L6pez Lechas.

7 - Efficiency of the Emergency Core Cooling System of the PWR C.N. Trilio I (1040 MW) Medium and Small Leaks

21 cm2 Leak at the Pressurizer (Relief Valve)

KWU - Work Report R-153

A. L6pez Lechas.

8 - Efficiency of the Emergency Core Cooling System of a 1300 MW PWR

Konvoi Project KKP-II

Medium and Small Leaks

422 cm2 Leak between Reactor Pressure Vessel and Steam Generator

(Hot Leg)

KWU - Work Report R-153

A. L6pez Lechas.

34.
9 - Anilisis de la eficiencia de los sistemas de emergencia y protecci6n

de un PWR C.N. Almaraz (930 MWe)

Nodalizaci6n Estado Estacionario Soluci6n Inicial

[Analysis of PWR Plant Emergency and Protection System Efficiency -

Almaraz NPP (930 MWe)] - [Nodalization Steady State Initial Solution] C.N. Almaraz - WORK REPORT

A. L6pez Lechas - Juan Manuel Blanco

Madrid 11-3-85

10 - Anilisis de la eficiencia de los sistemas de emergencia y protecci6n

de un PWR C.N. Almaraz (930 MWe)

Disparo de Turbina al 100% de potencia

[Turbine Trip at 100% Power]

C.N. Almaraz - WORK REPORT

A. L6pez Lechas - Juan Manuel Blanco

Madrid 28-5-85

11 - Movimiento en conductos de secci6n variable [Movement along Variable

Section Piping]

ETSIA Abril 1976

Crespo, A. LifiAn.

12 - Turbulencia [Turbulence]

ETSIA

Gregorio Millin.

13 - Movimiento turbulento de liquidos en tubos de secci6n constante

[Turbulent Movement of Liquids in Piping of Constant Cross-Section]

ETSIA

Gregorio Millin.

14 - Movimiento de fluidos ideales [Ideal Fluid Movement]

ETSIA Abril 1968

A. LiBin.

35.
15 - Movimientos con superficies de continuidad

[Movements with Continuity Surfaces]

ETSIA

Gregorio Millin.

16 - CinemAtica de Fluidos [Fluid Kinetics]

ETSIA Octubre 1967

A. Lifiin.

17 - Tensor de velocidades de deformaci6n, tensor de esfuerzos, fen6menos

de transporte [Deformation Velocity Tensor, Stress Tensor, Transport

Phenomena]

ETSIA Octubre 1967

A. Lifiin.

18 - Movimientos unidireccionales [One-directional Movements]

ETSIA Septiembre 1969

A. LiiiAn.

19 - Anilisis dimensionales y semejanza fisica (Dimensional Analysis and

Physical Similarity]

ETSIA

G. Millin.

20 - Assessment study of RELAP 5 / Mod. 2 cycle 36.04 based on pressurizer safety and relief valve test.

G.J. Stubbe and L. Vanhoenacker

NUREG/IA - 0034

21 - Anilisis de las deficiencias observadas en los estudios realizados

con el c6digo de cilculo Relap 5 Mod. 2 y su comparaci6n con el Relap 5 Mod. 1 y Relap 5 Mod. 2 - Work Report.

Alfredo L6pez Lechas - C.N.A.

36.
TABLE I

Number of Loops

Core

Number of Assemblies

Rod Array

Rod OD, cm.

Number or grids

Active Fuel Length, cm.

Number of control rods, FL

Steam Generator

Model

Shell design pressure, Nw/m 2

Reactor coolant pump

Model

Pump motor, hp

3

157

17 x 17

0.95

8R

365.76

48

D3-1

82.74 x 105

93D

7000

Thermal Design Paramters

NSSS power, MWt

Reactor power, MWT

Thermal Design Flow, Loop m3 /s

Reactor Coolant Pressure, Nw/m 2

Reactor Coolant Temperature, *K

Core outlet

Vessel outlet

Core average

Vessel average

Vessel/core inlet

Steam generator outlet

Steam Generator

Steam Temperature, *K

Steam Pressure, Nw/m2

Steam Flow, Tm/hr total

Feed Temperature, *K

Moisture, % max

Zero Load Temperature, °K

Hydraulic Design Paramters

Best Estimate Flow, m3 /s

Mechanical Design Flow, m3 /s

Pump design Point Flow, m3 /s/head (mts)

2696

2686

5.93

155.139 x 105

601.

599.6

583.

582

564.4

564.3

556

68.05 x 105

5379.5

496.8

.25

564.9

6.17

6.43

5.93/86.25

37.
TABLE I (Cont. 1°)

THERMAL AND HYDRAULIC DATA

General

Total core power

Number of loops

Best estimate flow per loop

Mechanical design flow per loop

Thermal design flow per loop

Design core bypass flow

Design vessel and core inlet

temp./enthalpy

System pressure

Best estimate vessel pressure

drop based on thermal/best

estimate flow

Best estimate core pressure

drop based on thermal/best

estimate flow

2686 MWt

3

6,17 m3 /seg

6,43 m3 /seg

5,93 m3 /seg

4.5%

564.4°K/1291.6 KJ/Kgr

155.139 x 105 Nw/m2

2.826 x 105/Nw/m 2

1.613 x 105 Nw/m2

4. FLUID SYSTEMS DATA*

4.1. RCS Geometry and Hydraulic Data

4.1.1. Reactor vessel

4.1.1.1. Volumes

2.

3.

4.

5.

6.

6.

Core fuel region

Upper plenum

Downcomer

3.1 above nozzle bottom

3.2 below nozzle bottom

+ barrel baffle volume

Lower plenum

Upper head

Inlet nozzle (incl. cold leg piping)

Outlet nozzle (incl. hot leg piping)

15.07 m3

26.06 m3

4.24 m3

18.05 m3

24.748 m3

12.82 m3

3.038 m3

2.8316 m3

38.
TABLE I (Cont. 20)

Pressure Losses

Operating conditions:

Pressure = 155.130 x 105 Nw/m2

Inlet temperature = 564.4*K

Flow = 5.93 m 3/s/loop

Rejion

2

Pressure droD (Nw/rn

inlet nozzle

barrel - vessel annulus

lower plenum

support plate

diffuser plate

core

outlet nozzle

plenum

0.5254 x 105

0.04 x 105

0.35579 x 105

0.05516 x 105

0.00

1.3976 x 105

0.1303 x 105

0.03034 x 105

Hot Leg

Segment from reactor vessel to S.G. inlet

- weight of steel in contact with

primary water:

- height (difference from cold leg):

- length:

- flow area:

- hydraulic diameter

- volume (including outlet nozzle)

- steady state pressure losses

K:

138783,84 Nw

1.3 m

5.514 m

2

0.426 m

0.737 m

3

2.831 m

-6 3/) 35.3 x 10 rn/(rn /hr)-

39.
TABLE I (Cont. 30)

Cross Over leg

Segment from S.G. out let to pump inlet

- weight of steel in contact with

primary water:

- height (difference from cold leg):

- length:

- flow area:

- hydraulic diameter:

- volume:

- steady state pressure losses

K:

74849.86 Nw

0.314 m

3.83 m

2

0.4867 m

0.787 m

3.36 m3

64.5 x 10-10 m/(m 3/hr)2

Cold Let

Segment from reactor coolant pump outlet to reactor vessel inlet

- weight and area of steel in coni

primary water:

- height (difference from cold le

- length:

- flow area:

- hydraulic diameter:

- volume (including inlet nozzle)

- steady state pressure losses

K:

tact with

g):

128917,73 Nw

0. m

6.925 m

0.383 m

0.6979 m

3

3.0389 m

35.3 x 10-10 m/(m 3/hr)2

40.
TABLE I (Cont. 40)

4.1.5. Pressurizer Surge Line

- area of steel in contact with

primary water:

- height (difference from cold leg)

- length:

- flow area:

- hydraulic diameter:

- volume:

- equivalent length

- steady state pressure losses

delta P at max. surge line flow

0.064 m2

0. m

17.06 m

0.64 m0.284 m

3

1.217 m

147.9

4.1577 x 105 Nw/m2

4.1.6. Pressurizer Data

3

39,6 m pressurizer general assembly

Pressurizer as built

Pressurizer

a. Number of relief valves: 2

b. Relief valve rated flowrate (each):

c. Relief valve rated pressure:

d. Number of safety valves

95.253 Tm/hr

162.03 x 10 5Nw/m2

3

41.
TABLE I (Cont. 50)

e. Safety valve rated flowrate when fully

opened (each)

f. Safety valve rated pressure

g. Number of spray valves

h. Spray valve capacity (fully open)

i. Heaters capacity

ON/OFF

Proportional

Steam Generator Volumes

- Inlet plenums including nozzles

- Tubes

- Outlet plenums including nozzles

2 Loop Flowrate Data

Active loop flow

Inactive loop flow

CVCS and Charging System Data

Maximum possible flowrate to RCS at power

Maximum possible flowrate to RCS at refueling

Minimum RCS water volume for RHR operation

Maximum letdown flow

172.352 Tm/hr

172.378 x 105 Nw/m2

2

0.022 m 3/seg/valve

1023 kW

377 kW

4.332 m 

3

17.78 m3

4.332 m3

6.27 m3 /seg

1.38 m3 /seg

0.0104 m3 /seg

0.018 m 3/seg

3

91.746 m3

0.0076 m3 /seg

42.
TABLE I (Cont. 6°)

Safety Injection System Data

Accumulator

a.

b.

C.

d.

e.

f.

g.

h.

1.

j.1

Total volume:

Water volume (LOCA)

Gas pressure (LOCA)

(design)

Minimum boron concentration (SLB)

Temperature (SLB)

Equivalent length/diameter

Line volume from accumulator to check

valve nearest RCS

Line volume from check valve to Reactor Vessel

Point of entry to reactor coolant system:

Line characteristics

hydraulic diameter

area

41.059 m3

27.6 m3

41.37 x 105 Nw/m2

48.26 x 105 Nw/M 2

2000 ppm

310.7 0K

132 m

0.996 m3

0.215 m3

cold leg

0.2667 m

0.05586 m3

SjafeýyInjetction~System

a. Startup time of safety injection

pumps; time from start to full speed

b. Diesel generator starting and sequences

loading delay time

c. Boron injection volume - from cold-leg SI pump to boron

injection tank

- from boron injection tank to a

header where flow branches and

is fed to each cold-leg connection

12 sec

12 sec

0.566 m3

0.7928 m3

* from the header to cold-leg 0.00566 m3

connections (total of all lines)

43.
TABLE I (Cont. 70)

d. Boron concentration in BIT 20000 ppm

e. BIT enthalpy (min) 111,6 KJ/Kgr

f. Boron concentration in RWST 2000 ppm

g. RWST temperature (LOCA) 273.6=K

RWST temperature (design) 338.5 0 K

Secondary side

A. Minimum water level in shell side of the SG

required to completely cover the tubes, from

tube sheet 8,43 m

B. Water and Steam masses

100 % power case - normal water level

at Tsec : 556.8°K

Steam mass 3920 Kgr

Water mass 42936.8 Kgr

Total mass 46857.6 Kgr

50% power case -

at Tsec : 562.4*K

Steam mass 4562.2 Kgr

Water mass 37379. Kgr

Total mass 41942. Kgr

0% power case -

at Psec : 76.329 x 105 Nw/m2

Steam mass 4318 Kgr

Water mass 44597.8 Kgr

Total mass 48916.5 Kgr

44.
TABLE I (Cont. 80)

Water level is 12.49 m above tubesheet

Preheater Data

A. Preheat heat transfer surface

- surface area of tubes in parallel-flow region 170 m2

2 - surface area of tubes in inlet region 156.3 mi

2 - surface area of tubes in counter flow region 312.6 m2

2 - surface area of tubes in cold leg boiling region 156.3 m2

B. Heat transfer resistance

- associated with preheater primary - 0.029971

2

side film resistance, m 2K/KW

- prehetater secondary - side film 0.07052

2-

resistance, m 2K/KW

- preheater fouling factor, 0.01763

2

m °K/KW

- overall preheater heat transfer 0.1918

resistance, m 2K/KW

C. Total volume of steam and water in preheater 7.518 m 3

45.
TABLE I (Cont. 90)

Steam Generator Data

1.

2.

3.

4.

5.

6.

7.

8.

9.

10.

11.

12.

13.

14.

15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

27.

28.

29.

30.

31.

32.

33.

34.

35.

36.

37.

38.

39.

40.

41.

42.

43.

Heat Transferred - MWt - 898.66

Heat Transfer load - MWt - 897.8

Primary flow rate - m3 /s - 5.93

Primary Flow Rate - Kgr/seg - 4.423

Coolant Inlet Temp. - °K = 599.6

Coolant Outlet Temp. - °K = 564.3

Average Coolant Temp. - °K = 582

Primary Operating Pressure - x 105 Nw/m 2 - 155.139

Feedwater Temperature - °K = 497

Steam Flow Rate - Kgr/seg - 496

Steam Pressure - x 105 Nw/m 2 - 68.05

Steam Temperature - °K = 556

Primary Pressure Drop - x 105 Nw/m 2 - 2.208

Circulation Ratio = 2.31

Total Heat Transfer Surface - m2 = 4459.3

Preheat Heat Transfer Sdrface - m2 = 841

Number Counterflow Passes = 4

Number Parallel Flow Passes = 3

O.D. & Wall of Tubes - cm = 3/4 x 0.1092

Tube Material - Inconel 600

Number of U-tubes - 4674

Tube Pitch - cm = 2.6987

Primary Volume - m3 = 26.47

Secondary Volume - m3 = 168.4

Primary Design Pressure - x 105 Nw/m 2 = 171.343

Secondary Design Pressure - x 105 Nw/m 2 = 81.707

Number Tube Supports = 8

Circulation Hole Diameter - cm = 1.27

Average Tube Length - m = 17.039

Maximum Tube Length - m = 19.29

Tube Straight Length - m = 7.528

Minimum Bend Radius - cm = 5.715

Maximum Bend Radius - cm = 153.255

Weight of Tubes - Nw = 427916.8

Weight Tube Plate Forging - Nw = 404786.2

Weight Channel Casting - Nw = 267781.6

Weight Elliptical Head - Nw = 182376.2

Weight Stub Barrel - Nw = 161024.8

Weight Lower Shell - Nw = 419910.

Weight Transition Cone - Nw = 183710.6

Weight Upper Shell - Nw = 591610.

Weight Misc. Internals - Nw = 493305

Total Dry Weight - Nw = 3020327

46.
TABLE II

PROJECT REACTOR COOLANT SYSTEM PRESSURE SETPOINTS

(10' Nw/m )

Project pressure 171.2

Operation pressure 155.131

Safety valves 171.2

Power operated relief valves

Pressurizer spray valves (begin to open) 155.72

Pressurizer spray valves (completely open) 159.1

High pressure trip 164.3

High pressure alarm 160.9

Low pressure trip (typical but variable) 130.

Low pressure alarm (interlocked with

pressurizer relief valve) 150.5

Hydrostatic test pressure 214.0

Backup heaters connected 152.29

Proportional heaters (begin to operate) 155.0

Proportional heaters (total operation) 152.97

47.
TABLE III

EVENT SEOUENCE FOR LOSS OF ALL OFF-SITE

POWER AT 100% POWER OPERATION

I)

2)

3)

EVENT

Blackout signal (main pump shutdown)

Pressurizer relief valve opening

VALUE

0.025 sec.

1.8 sec.

4)

5)

6)

7)

8)

9)

Reactor trip signal

On low mass flow in the primary

Control rod fully inserted

Steam generator relief valve open

High pressurizer pressure signal (without effect)

Pressurizer relief valve closure

Main feedwater (cut off)

Auxiliary feedwater (turbine-driver pump start-up)

2 sec.

4.125 sec.

4.2 sec.

6.125 sec.

8 sec.

8 sec.

60.025 sec.

48.
TABLE IV

RUN STATISTICS

CASE .....................................................

COMPUTER .................................................

CPU TIME (S) .............................................

REACTOR TIME .............................................

C (TOTAL NUMBER OF ACTIVES VOLUMES IN THE MODEL) .........

DT (TOTAL NUMBER OF TIME STEPS) ..........................

(CPU X10 3 ) / (CXDT) ......................................

CPU TIME / REACTOR TIME ..................................

BLACKOUT

CYBER 180/830

3098

100

173

1136

15.76

30.98

49.
TABLE V

Steady state results at 100% power

Almaraz NuVariable Units clear Power RELAP 5/MOD.2

Plant (calculated)

Reactor power (Mw) 2686 2686

RCS average temperature (K) 308.4 308

Pressurizer level (%) 56 55

Pressurizer pressure (MPa) 15.14 15.73

RCS mass flow rate (per loop) (Kg/s) 4423 4425

Reactor coolant pump speed (RAD/s) 153.4 153.1

Steam generator pressure (MPa) 6.805 6.79

Steam generator circulation rate (-) 2.1 Z.8

Steam generator collapsed liquid level (W) 12.491 12.697

Steam flow rate (Kg/s) 496 477

Feedwater temperature (K) 497 497

(INPUT DATA)

50.
PROYECTO ICAP (ICAP PROYECT)

; i I ; . I 'I I I I~ I I I. 11.11

RAMA CAL.IENTE (HOT LEG)

LONGI TUD (LENGTH)

) .. ,. W ,( ;.AR

AREA (FLOW AREA,) U'

I-.

"-1I 1.-4

I0

C, ' , I t",

DIAMETRO HI

(HYDRAUL I C

( '2 ." i '.I I ", 

DRAUL I CO

D IAMETER)

', 4 i ;' F ri )

•-'Ilpi ~'IPt. HI~A,~k.;i I
PROYECTO ICAP (ICAP PROYECT)

I'1I 1 I IIl

GENERADOR DE VAPOR

I0

1-n
PROYECTO I CAP (I CAP PROYECT)

RED [JE.Ti- ' F 'i- •J0 -. .. IHL- '1 " j

RAMA FRIA

". PE

(COLD LEG)

5P•,'•L A D- ,,

1---12, .; F -A1."7 j0~

PIPE• 1I

wI.3-9 21 2!I •

P I Pi

V*I'./ URL .._
PROYECTO ICAP (ICAP PROYECT)

: T I ' ,, "i-r , . .lFt [i.i11 it I

VASIJA DEL REACTOR

2-0

In1

1 G~O

101

410 c')
PROYECTO ICAP (ICAP PROYECT)

TMDPVO L

~E;I*I '1Li~.(ZJ

Ul

tJI

F I GUF1PE i..-..

• ;,,
PROYECTO ICAP (ICAP PROYECT)

SA.FETY* ,"-".I-.ID REL I Er VAI._"../E:S

-Il
F. a- 7.

II

T.Ia.

V

~I~F *~ - -- a -

1~

t....

1 %~a* - *-

-

L.~-±~- -ci

ul.

a-I ~ ~ ~ ~ ~ r ------.- 11J~S~j.

-ez j* zFtT

r-. Av______,a

ED.)

it -" a . fa.

*lnq
,0 Owf p~ressuri zer re 1~ 4 e t-nk

Air-operated valve

Motor-operated valve

PVC-444A

To tlh. pressurizer head

F IGURE-8

58
PROYECTO ICAP (ICAP PROYECT)

~....F E. I' .~FIL.. 1L -1 ElF .A L %v. L. -z.i

TMDPVOL

TMDPJ UN

TMDPVOL

U' TMDPJUN

530 BRAINICH

I* GI..)RE 9
PRESSURE

I NW/m

2 )

172.20

171.95.

171,70-

171.45-

171.20.

I ____________________________________________________________________________ ____________________________________________________________________________ I

I

I

I

/

/

/

/

/

/

4-,

.10

9 +

HYSTERISIS CURVE

SAFETY VALVES

CONSIDERED JOINTLY

/ /

/

- 4 i

/

/

I

I

I

I .1. h I I I I

I £ Itoo I I I I

50 ISO

MASS FLOW

(Kgr/see)

FIGURA-I0

60
PRESSURE

( Nw/m2)

162,92

162,42

161,92

161,42

160.92.

HYSTERISIS CYCLE

FOR PRESSURIZER RELIEF VALVES

I

10 20 30 40 50 60 MASS FLOW

61

FIGURE -tl-
(PCM IC)

MODERATOR TEMPERATURE

62

FIGURE -12
Jig)A

ON

1. 0CCc

.8

'.6

p.4-

-cP.2

-t

-w

-I 4 1 1 

I

0 100 200 300 400 T-OC

273 373 473 573 6;3 T-*K

FIGURE 

- 13
S[IX

2.0

1.0

0.0

-1.0

FOR POSITIVE NYC

IN THE CASE OF ALMARAZ I CYCLE 5, THE COEFFICIENT IS

ONLY VERY SLIGHTLY POSITIVE, AND ANALYSIS IS UNNECESSARY

I I I I -2.0 -

-2.

-3.0

-4.0

-5.0

-I -6.0 .4 I-- -1------- I------

I I + 4 4 4

0,1 02 d.3 I1.4 0.5 0.6

0 0I 0.7. 0.8 0.9 I.0 S [ur /CM2]

FIGURE -14
ALMARAZ NUCLEAR POWER PLANT

CONTAINMET

SPROY PUMP

ESSENTIAL

SERVICE WATER PUMP ESSENTIAL

SERVICE WATER PUMP

AUXILIARY FEED

WATER PUMP

AHIGH PRESSURE 3SAF ETY INJECTION

IMP I PUMP I

RESIDUAL HE

REMOVAL PU

SP

NTAINMET

'ROY PUMP

a' u'

I

DIESEL 12 DIESEL 22

FIGURE- 1b
,"I

H

0)

ON

ON

0

z

0

(I?

7-

L

3047.0-

2748.8 -

2450.6-

2162,47

1 854,2-

1556.0-

1257.8-

661 .4-

363.2-

65.0

SECCION DE TERMOHIDRAULICA

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SECCION DE TERMOHIDRAULICA

CENTIRAIL Nr-UCL.EFAR [I.L rL..HRA Z ),_..

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T I ME (SECONDS)

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SECCION DE TERMOHIDRAULICA

CENTRAlL NUCLEAR DE ALMRI:IZ

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SECCION DE TERMOHIDRAULICA

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ItsCK u
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I. REPORT NUMBER

124- 1) IA iwipw. by NRC. Add Vol.. S•zp.. Rev.,

NRCM 1102. and Addendurm Numbers. It nlv.)

X101.3202 BIBLIOGRAPHIC DATA SHEET NUREG/IA-0123

(Se i;nsu:cions on the rvs IICSP-AL-BOUT-R

2. TITLE AND SUBTITLE

Application of Full Power Blackout for C. N. Almaraz 3. DATE REPORTPUBLISHED

with RELAP5/MOD2 MONTH 'YEAR

June 1993

4. FIN OR GRANT NUMBER

L2245

S. AUTHOR(S) 6. TYPE OF REPORT

A. L. Lechas Technical.

7. PERIOD COVERED tcctlumn Dates)

8. PER run MII•E3 4r11,145AlN.• a W14 - NAMEAND• . VORE.5 .))fl (•lI~r NC.- provide UwOiono Office orrt~tn A J.U.1 Nuclear Regulatory Cmmssonn. a•nd mlloJisng loonrr".'itcontr1ctrorpov#oe

C. N. Almaraz I y II

c/Claudio Coello, 123

28006 - Madrid, Spain

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (ii NRC. type -Sa as coove"i;f contrator. Provide NRC Divition. Office or Region. U.&. NusclaRogulatrory Commiesion. n d anddooma

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 Wwors, en)

The analysis group of Almaraz Nuclear Power Plant has developed a model of the plant with

RELAP5/MOD2/36.04. This model is the result of the work-experience on the code RELAP5/MOD1. that was the

standard code during the period 1984/1989. Different solutions were adopted in the network to adequate the model

to RELAP5/MOD2 Computer Code. This transient was selected for ICAP because it presents an experience with

the same transient calculated with RELAPS/MOD1/CY 29 Computer Code. The comparison between both analysis

will be interesting.

12. KEY WORDS/DESCRIPTORS (List w• srooo rparns cost will measit mnearthn locamotng hereporr.I 13. AVAI LABILITY STATEMENT

ICAP Unlimited

Alm araz 14. SECURITY CLASSIFICATION

RELAP5 ,ro,,•.,

Blackout Unclassified

(Trni Reooortr

Unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 33S 12491


Federal Recycling Program

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

SPECIAL FOURTH-CLASS RATE

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300