NUREG/IA-0067

- ICSP-GA-LOCA-T

International

Agreement Report

Recirculation Suction Large Break

LOCA Analysis of the Santa Maria

De Garoia Nuclear Power Plant

Using TRAC-BF1 (GIJi)

Prepared by

J. V. Lopez, J. Blanco, Polytechnical University of Madrid

J. L. Crespo, University of Cantabria

R. A. Fernindez, Nuclenor, S. A.

Polytechnical University of Madrid

c/Josh Abascal, 2

28006-Madrid

Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

August 1992

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-0067

ICSP-GA-LOCA-T

International

Agreement Report

Recirculation Suction Large Break

LOCA Analysis of the Santa Maria

De Garoiia Nuclear Power Plant

Using TRAC-BF1 (GIJ1)

Prepared by

J. V. Lopez, J. Blanco, Polytechnical University of Madrid

J. L Crespo, University of Cantabria

R. A. Fernandez, Nuclenor, S. A.

Polytechnical University of Madrid

c/Josh Abascal, 2

28006-Madrid

Spain

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, DC 20555

August 1992

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.
FOREWORD

This report has been prepared by NUCLENOR 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 thermalhydraulic activities between the Consejo

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

Nuclear Regulatory Commission (USNRC) in the form of Spanish

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

the validation of the TRAC and RELAP system codes.

The Consejo de Seguridad Nuclear has promoted a coordinated Spanish Nuclear Industry effort (ICAP-SPAIN) aiming to

satisfy the requirements of this agreement and to improve

the quality of the technical support groups at the Spanish

Utilities,'Spanish Research Establishments, Regulatory Staff

and Engineering Companies, for safety purposes.

This ICAP-SPAIN national program includes agreements

between CSN and each of the following organizations:

- Unidad Elictrica (UNESA)

- Uni6n Iberoamericana de Tecnologia Elictrica (UITESA)

- Empresa Nacional del Uranio (ENUSA)

- TECNATOM

- LOFT-ESPAIA

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

Coordinaci6n". This committee has approved the distribution

of this document for ICAP purposes.

D-1/90-MPNV iiii

ABSTRACT

This document presents a recirculation suction large break loss of coolant accident analysis

of Santa Maria de Garofa Nuclear Power Plant.

The plant is a General Electric Boiling Water Reactor 3. containment type Mark I. It Is operated

by NUCLENOR, S.A. and was connected to the grid in 1971.

The analysis has been performed by the Chair of Nuclear Technology from the Polytechnical

University of Madrid, the Applied Physics Department from the University of Cantabria and the

Analysis and Operation Section from NUCLENOR, S.A. as a part of an agreedment for

developing an engineering simulator of operational transients and accidents for Santa Maria de

Garoha Power Plant.

The analysis was performed using the frozen version of TRAC-BF1 (GIJ1) code and is the

first of two NUCLENOR contributions to the International Code Applications and Assessment

Program (ICAP).

The code was run in a CDC Cyber 830 with operating system NOS 2.6, property of

NUCLENOR, S.A. Additionally the code was updated into a CDC Cyber 932 operating system

NOS VE, in order to solve the problems related to central memory limitations found at early

stages of the work. A programming effort was carried out in order to provide suitable graphics

capabilities to the code.

V

CONTENTS

FOREWORD .............................................. iii

ABSTRACT .............................................. v

LIST OF TABLES .......................... .............. viii

LIST OF FIGURES ...................................... ix

EXECUTIVE SUMMARY .................................... xi

I.- INTRODUCTION ....................................................................................... 1

II.- PLANT AND TRANSIENT DESCRIPTION .................................................... 2

Ill.- CODE INPUT MODEL DESCRIPTION ..................................................... S

IV.- CALCULATIONS RESULTS .................................................................... 9

V.- RUN STATISTICS .................................................................................... 14

VI.- CONCLUSIONS ....................................................................................... is

VII.- REFERENCES ........................................................................................ 16

vii
LIST OF TABLES

Page

TABLE 11-1

TABLE I11-1

TABLE IV-1

TABLE IV-2

TABLE V-1

Garofia NPP Design Characteristics

Description of the TRAC components

Initial Conditions

Chronology of Events

Run Statistics

17

19

20

21

23

v,.i°i
LIST OF FIGURES Page

Fig. 11-1 Garofta Emergency Core Cooling System and Isolation Condenser System 24

Fig. 111-1 TRAC-BF1 Axial Nodalization 25

Fig. 111-2 TRAC-BF1 Radial Nodalization 26

Fig. 111-3 Fuel Element model 27

Fig. IV-1 Steam Dome Pressure 28

Fig. IV-2 Reactor Power during the first seconds 29

Fig. IV-3 Reactor Power 30

Fig. IV-4 Downcomer Uquid Level 31

Fig. IV-5 Broken Loop Mass Flow Rates 32

Fig.' IV-6 Low-Powered Channel Mass Flow Rates 33

Fig. IV-7 Average-Powered Channel Mass Flow Rates 34

Fig. IV-8 High-Powered Channel Flow Rates 35

Fig. IV-9 Recirculation Pumps Speed 36

Fig. IV-1O HPCI and LPCS Mass Flow Rates 37

Fig. IV-11 Lower Plenum Void Fraction 38

Fig. IV-12 a, b Hot Channel Uquid Heat Transfer Coefficient 39, 40

Fig. IV-13 Hot Channel Vapor Heat Transfer Coefficient 41

Fig. IV-14 Hot Channel Clad Temperature 42

Fig. IV-15 Hot Channel Void Fraction 43

Fig. IV-16 Upper Tie Plate Void Fraction In Inner Ring 44

ix
LIST OF FIGURES (Cont.)

Fig. IV-17

Fig. IV-18

Fig. IV-19

Fig. IV-20

Fig. V-1

Maximum Peak Rod Temperature in Channels

Channel Collapsed Liquid Level

Rod Temperature at upper levels of Hot Channel

Hot Channel Mixture Density

TRAC Time Step

Page

45

46

47

48

49

X
EXECUTIVE SUMMARY

A best estimate analysis of a Recirculation Suction Pipe Large Break Loss of Coolant

Accident Analysis for Santa Maria de Garofla Nuclear Power Plant using TRAC-BF1 code is

presented.

An investigation of the existing safety margins respect to the available Garofla licensing

analysis based on GE SAFER-GESTR Methodology was the main objective of this work.

Boundary and initital conditions of the analysis were consistent to those ones used in SAFER

analysis.

A margin of 160°C in peak clad temperature (PCT) respect to the higher PCT calculated by

SAFER code was obtained with TRAC-BFl.

As a result of the analysis, considerable knowlegde on LOCA phenomena and expertise in

using TRAC-BF1 code has been obtained. It is believed that the developed Garo•la LOCA

model is qualified to performe best estimate analysis according to the state of the art of the

current BWR LOCA technology.

xi

1. INTRODUCTION

NUCLENOR, a Spanish Electrical Utility which owns the Santa Maria de Garofla Nuclear

Power Plant made a decision in 1987 of carrying out a project to develop a specific engineering

simulator of accidents and transients based on TRAC-BF1 and BWR-LTAS codes. In order to

obtain the TRAC-BF1 code NUCLENOR joined ICAP-Spain program through UNESA under

the compromise of carrying out one plant-specific application case and one plant-specific

assessment case with the code.

NUCLENOR selected as an application case the Analysis of a Large Break Loss of Coolant

Accident. The analysis, described in the present report, was carried out by the Chair of Nuclear

Technology, belonging to the Polytechnical University of Madrid, the Applied Physics

Department of University of Cantabria and the Analysis and Operations Section of NUCLENOR.

The application analysis consisted a postulated large double-ended guillotine break in the

piping on the suction side of a recirculation loop. It was intented to simulate the limiting LOCA

case derived from the Garofla application of General Electric SAFER-GESTR Methodology in

order to evaluate the existing safety margins in licensing.

Besides the postulated break, the following main hypothesis were assumed in order to use

the same boundary conditions as SAFER calculation:

1. Loss of off-site power coincident with the LOCA initiation. So that recirculation pumps

would coastdown and the Emergency Core Coolant Systems (ECCS) would be electrically

supplied from Diesel Generators.

2. The worst single failure is also postulated to occur. In this case the failure of Low Pressure

Coolant Injection (LPCI) System, remaining available the Core Spray (CS) System, High

Pressure Coolant Injection (HPCI) System and Automatic Despresurizatdon System (ADS).

Additionally, a programming effort was carded out in order to provide suitable graphics

capabilities to the code.

A summary description of the Santa Maria de Garotla Nuclear Power Plant and the simulated

accident is given in Section II. Section III describes the code input model nodalization.

Calculation results and discusion of main phenomena are presented In Section IV. Run

statistics are summarized in Section V and the conclusions are given in Section VI.

I
II. PLANT AND TRANSIENT DESCRIPTION

11.1 PLANT DESCRIPTION

Santa Maria de Garof~a Nuclear Power Plant is a General Electric Boilng Water Reactor 3, with

containment type Mark I. The plant is operated by NUCLENOR, which is a subsidiary of

lberduem, S.A. and Electra de Viesgo, S.A., and was connected to the grid in 1971. The plant

is rated at 1380 Mw (thermal) and is located in the province of Burgos (Spain).

The nuclear boiler assembly consists of the reactor pressure vessel and internal reactor

components such as the core structure, steam dryer assembly, fuel supports and control guide

tubes.

The reactor core is made up of 400 fuel assemblies and 97 control rod blades. Each fuel

assembly has 64 rods in a square array (8 x 8). Each control rod blade consists of sheathed

cruciform array of vertical absorber rods made of boron carbide (B4C). A complete description

of relevant parameters of the plant Is shown in Table I1-1.

A Recirculation System, consisting of two external centrifugal recirculation pumps and

twenty reactor vessel Internal jet pumps, provides the core coolant flow to meet the plant

thermal rating. This coolant consists of saturated water rejected from the steam separators and

dryers that has mixed with subcooled feedwater entering the vessel at the steam separator

elevation. The recirculation pumps suction exits the vessel at an elevation just above the

shroud support ring, goes through the outside loop and re-enters the vessel through a riser

pipe to become the driving flow for the jet pumps. The remainder of the coolant is thus

entrained in the jet pump, mixing with the driven flow in the pump throat section. Flow then

exits the jet pump via the diffuser section and Is directed to the core inlet plenum.

Main Steam System consist of four lines that penetrate the reactor vessel and provide steam

to the turbine at rated operating conditions. Three relief valves (RV's), two safety-relief (SRV's)

valves discharging into the Suppression Pool, and seven Safety Valves (SV's) discharging into

the Drywell, are installed on the steam lines. As well as, there are two isolation valves per line

and one flow restrictor per steam line downstream RV's and SRVs.

The reactor vessel is Inside an Inerted containment called Drywell and the Supression Pool Is

connected to the drywell through vent pipes.

An Isolation Condenser System allows for condensating steam from the reactor vessel and

returning the condensate to one of the recirculation loops, when the vessel is isolated from

turbine and main condenser.

2
The Garofa Emergency Core Cooling Systems, as shown in Fig. I1-1, consist of following

systems:

High Pressure Cooling Injection System (HPCI)

Automatic Depresurization System (ADS)

Low Pressure Cooling Injection System (LPCI)

Core Spary System (CS)

The HPCI System has a turbine-drven pump which injects water into the vessel, coming from

Condensate Storage Tank first, and then from supression pool. Either high pressure in Drywell

or low low level in the vessel signal produces automatic starting of the system.

The actuation of ADS consist of automatic opening of three relief valves, discharging steam

from the vessel into the supression pool, when high drywell pressure and Low Low Level

signal exist for two minutes. After the ADS actuation, the low pressure systems, LPCI and CS,

are able to inject water into the vessel.

The LPCI System has four pumps and two interconnected hydraulic trains injecting in each

recirculation loop. An electric logic exists for selecting the unbroken loop, so that permiting the

entre flow the four pumps inject into those one, the water enters to the core form the bottom.

The CS System has two independent hydraulic trains, each one consisting of one pump

which takes water from the supression pool and inject It directly on to the core through a nozzle

sparger.

Both LPCI and CS starts automatically when either a drywell high pressure, or Low Low

vessel level signal exists, and when the vessel pressure falls below 21kg/cm2 .

11.2 TRANSIENT DESCRIPTION

A full double-ended guillotine break in the piping on suction of a recirculation pump was

postulated to occur at the time zero.

A loss-of-off site power was assumed to occur simultaneously with the break. This caused a

trip of the recirculation pumps, condensate and feedwater pumps and as a result isolation of

main steam lines.

The safety system single failure criteria was applied to the admission valve in the LPCI line at

the Intact recirculatlon loop. The LPCI admission valve was assumed to fail closed which

prevented any LPCI from reaching the reactor vessel. This single failure thus effectively failed

the entire LPCI. Therefore the available emergency systems were ADS, HPCI and CS.

3
Provided that the main objective was to know the existing safety margin respect to the

licensing calculations by SAFER-GESTR GE Methodology, the boundary conditions were

intended to match as much as possible to those ones used in the mentioned method applied

to Garofa.

According to that, the core was assumed to be operating at 104% rated thermal power at the

start of the transient. As well as maximun average planer linear heat generation rate

(MAPLHGR) of 12,5 kw/ft was assumed. Gap conductance and axial peaking factors for the

high power bundle were consistent to those ones used In SAFER analysis.

4
Ill. CODE INPUT MODEL DESCRIPTION

The Transient Reactor Analysis Code (TRAC) is an advanced best estimate code for

analyzing Light Water Reactor (LWR) accidents. The version for Boling Water Reactors was

developed at INEL (Idaho) under the sponsorship of the Reactor Safety Research Division of

the US Nuclear Regulatory Commission (NRC).

Specifically, TRAC-BF1/G1J1 was developed to analyze postulated transient and accidents

in BWRs. It uses a full developed two fluid model with one-dimensional geometry, except for

the vessel where the resolution is three dimensional. The two-fluid model, in conjuntion with a

model of stratified flow regime, handles a countercurrent flow treatment in a best estimate way

that the drift-flux model used in the past versions of TRAC-B.

Several different types of hydrodynamic components, including PIPE, VALVE, CHAN,

VESSEL, PUMP, TEE, JETPUMP, BREAK and FILL components, are used in TRAC-BF1. The

BREAK and FILL components are used to impose thermalhydraulic boundary conditions. The

other components can be used to represent different types of hardware such as a pipe, valve,

fuel channel, reactor pressure vessel, jet pump, or pump. The user can node all the

components, except BREAK and FILL components, with as many hydrodynamic cells as

desired. All those different types of TRAC-BF1 components were used In the LOCA model of

Garorla. The reactor vessel, both recirculation loops, and portions of the feedwater, steam, and

safety systems were represented in the TRAC-BF1 model.

The development of TRAC-BF1 input deck for the analysis of the large break loss of coolant

accident at the Santa Maria de Garorla Nuclear Power Plant was based on data taken from

drawings (Ref. 2), and specific technical documents related with the nuclear fuel design (Ref.

3).

The model contains 33 components with 43 fluid junctions. Table Il1-1 Includes a listing of all

the TRAC components used in the model. Figure Il1-1 shows the Identification number and the

relative location of each component within the TRAC-BF1 model. The illustrated model

represents Garofla for normal operation only. No broken or failed components are modeled in

the Figure.

The reactor vessel has been modeled by the VESSEL component with the nodalization

shown in Figures Il1-1 and 111-2. This nodalization is made up of eight axial levels, four radial

segments and one azimuthal sector.

The radial segments are distributed such that the inner three rings extend over the core

region with the first ring containing 84 real channels, the second ring containing 240 real

channels and, the third ring containing 76 real channels. The fourth radial ring models the

reactor vessel downcomer.

5
The first axial level extends from the vessel bottom to the jetpumps discharge support ring.

The second one goes from these support ring to the core bottom. The core is divided into two

axial levels (3 and 4) of different lenght. The remaining four axial levels 5 through 8, represent

the steam dome, upper plenum, with the CSs connections being made at axial level 5, the

feedwater system at axial level 6 and the main steam line at the axial level 7. The downcomer

extends from axial level 2 to axial level 7.

Instead of individual steam separator and dryer components, a perfect separator option is

used in axial levels 6 and 7. This component allows the vapor for continuing upward into the

axial level 8 and liquid draining radially outward into the downcommer region. Double-sided slab

models account for the heat capacity and transmission within vessel internals.

Contained in the lower levels of the VESSEL, axial level 1 and axial level 2, the PIPE

components number 2, 3 and 4 model the control rod guide tubes. These three pipes

represent 97 real control guide tubes. The guide tubes were modeled from the top of the

control rod drive housing to the core plate. The volume of the guide tubes and the vessel cells

were computed assuming that the control rods were fully inserted into the core. The heat

transfer between the fluid In the guide tubes and the lower plenum has been modeled.

Fuel bundle modeling is accomplished using four CHAN components simulating the 76

peripheral bundles by CHAN component number 36, 240 average bundles by CHAN

component number 38, and 84 hot bundles by CHAN components numbers 40 and 43.

As shown in Figure 111-3, the actual 8 x 8 rod array channel with two water rod Is modeled by a

geometrically identical 2 x 2 rod array CHAN component, with one water rod. All four CHAN

components have this structure. The axial distribution of the CHAN components consist of 15

axial levels, 10 of them in the active core region, with a bypass flow path modeled by an explicit

leak path. The total leak flow from all the CHAN components represent a 10% of the total flow

through the core.

Axial and radial power profiles agree with data from General Electric Propietary Information

(Ref. 5). Radial fuel rod dimensions represent beginning of life values. Reactor decay power

history was based on the American Nuclear Society decay heat standart (ANS-5.1) (Ref. 4) built

into the code. Additionally, a decay and fission product neutron capture was modeled.

Although the control rods are not physically modeled, the negative reactivity Insertion

associated with the rods Is accounted for a scram reactivity vs. time table derived from

Reference 3.

The use of begining of life radial fuel rod dimensions yields the maximum size for the gap

between the fuel and the cladding, since creep down, fuel craking, or fuel swelling is not

considered. As a result of using the maximum gap size, the fuel stored energy is the maximun

value for a given value of reactor power provided that the cladding has not previously been

ballooned.

6
Each recirculation loop have been modeled by six TRAC-BF1 components. The intact loop

(broken loop) is modeled by the PIPE component number 20 (number 10), representing the

suction piping end connected to the vessel, the VALVE component number 21 (number 11),

representing the isolation suction valve, the PUMP component number 22 (number 32) and

the VALVE component number 23 (number 31), simulating the recirculation pump and the

isolation discharge valve. Two PIPE components, numbers 24 and (numbers 41 and ) take into

account the five actual risers which supplie water into the jetpumps. Finally, the JETPUMP

component number 25 (number 26) simulates the actual 10 jetpumps. The JETPUMP

discharge is connected to the level 1, ring 4 of the vessel.

The suction pipes at the recirculation loop was represented by four nodes, which is a fine

nodalization in order to follow better phenomena of critical flow. The PUMP component

represents only the fluid volume between the inlet and outlet nozzles of the pump. The

internal friction is accounted for the pump homologus curves.

Piping comprising the feedwater system, main steam line, and core emergency cooling

system are also connected to the vessel. The feedwater line is modeled using a FILL

component (FILL number 61) connected to the VESSEL at the level 6 into the ring 4 via one

pipe (PIPE number 60). Also connected to this pipe is the HPCI (FILL number 62) by one leak

path-system.

The CS Is simulated by two FILLs (FILL numbers 73 and 71) connected to two pipes (PIPE

number 72 and 70). The simulation of two CS Is due that the real plant has two pumps with small

differences in their behavoir. The main steam line is modeled by two pipes (PIPE number 50

and 54) and the main steam isolation valve (VALVE number 52) cuts steam flow out of the

vessel. The BREAK component number 58 provides a back pressure representative of the

turbine.

The reactor point kinectics option was turned on the Power Cards to calculate the core

power rather than specifying it as a funtion of time after a trip. The Information used is contained

in Reference 3 to specify the programmed reactivity associated with the control rods and to

specify the reactivity-feedback coefficient for changes in core-averaged fuel, coolant

temperature, void fraction and core-average boron concentration.

The main option selected in the components was the critical flow model that was applied at

the jet pump nozzle and the BREAK juntions in the broken loop. The counter current flow

limiting (CCFL) model was applied at the side entry orifice of the bundle, at the upper tie plate

and at the core bypass.

Several control blocks were added to the input in order to adjust the recirculation pump

speed during the steady state calculation to yield a desired core flow and, as well as all the

necessary control blocks to control the downcomer level.

7
The initial conditions predicted by TRAC-BF1/GIJ1 for the analysis of the LOCA are

compared with tipical measured data at plant in Table IV-1. Calculations produced stable initial

conditions.

In order to run the transient, a modification consisting in two BREAK components in junction

number 210, was introduced in the model. As well as the following boundary conditions were

assumed:

a.- Decay Heat Generation rate was used the American Nuclear Society (ANS) 5.1 with a

multiplier factor of 1.0.

b.- No consider heat losses to the drywell environment.

c.- A double-sided slab in the VESSEL component was available to simulate the cylindrical

wall and at each guide tube cell the heat transfer between the fluid in the guide tubes and

lower plenum is modeled.

d.- The backpressure at the double-ended break was assumed constant.

e.- Loss of off site power coincident with the LOCA Initiation was postulated, so that

recirculation pumps would coastdown and pumped ECC Injection would be delayed a time

equivalent to that needed for a starting diesel generators.

f.- Only the high pressure coolant injection system (HPCi) and the low pressure coolant spray

(CS) are allowed.

g.- Peaking factors were consistent to those ones used in SAFER Analysis.

h.- Feedwater flow was assumed zero since the LOCA Initiation.

8
IV. CALCULATION RESULTS

The calculation of the LOCA with TRAC-BFI/G1J1 has been developed In two steps. First of

all, a steady state calculation was carried out at 104% of nominal thermal power. The second

step consisted of the transient calculation.

The steady state calculation was also developed in two steps. A first run of 300 s real time

with the built-in control system governing the level in the vessel, the recirculation pumps speed

and the mass flow rate in the core. Secondly a null transient calculation including point kinetics

was performed. After 100 seg', the simulation got a stable behaviour. Table IV-1 shows the

values obtained at stable conditions and typical values measured at plant

The convergence criteria for the outer iteration (EPSO), the vessel iteration (EPSI), and the

steady state calculation (EPSS) were 1.OE-04, 1.01E-04, 1.OE-04, respectively. The maximum

number of the outer iterations (OITMAX) and VESSEL iterations (IITMAX) were 10 and 100

respectively.

From the initial conditions described in Table IV-1, and the assumptions described in chapter

Ill, the transient calculation was run.

The chronology of events Is summarized in the Table IV-2 and the main results are

represented in the figures IV-1 through IV-18.

Calculated steam dome (level 8 in the model) pressure vs. time is shown in Figure IV-1. The

pressure began to decrease following the break due to the fluid mass discharge through the

two break sides. The main steam isolation valves closure produced a slighty recovery of

pressure at 4.5 seconds. Once the valves had closed the pressure rate was being controlled

by the core steam generation and the mass flow rate through the break sides. The uncovery of

the recirculation suction line takes place at about 8.0 sec. As a result, pressure rate increase

first, provided that steam is being discharged through both break sides, and then decrease

due to the lower plenum flashing ocurring from Instant 9.0 sec. until 11.0 sec. Thereafter the

pressure continues decreasing until reaching the postulated constant drywell pressure.

The calculated reactor power vs time is shown In Figure IV-2. and IV-3 Figure IV-2 shows the

reactor power during the first seconds. The decreasing core flow caused an increasing void

fraction and consequently a decreasing reactor power. The scram was generated at 0.5 s due

to the low water level signal. The control rod Insertion began at the same time and the rods

were fully inserted 2.9 s after, at that time the core power was all decay heat. According to that

the power goes down slowly for the rest of the calculation (Fig. IV-3).

The calculated downcomer liquid level Is shown in Figure IV-4. The level corresponds to a

collapsed liquid level based on downcomer cells void fractions. The liquid level decreased

9
quickly due to the loss of fluid through the break until the instant 8.0 of the transient. At that

time level of liquid was 0.73 m above the jet pump supporting ring. The small increase in the

level from the instant 9.0 to 12.5 s was caused mainly by the flashing in the lower plenum which

forced liquid into the downcomer. There were two flow paths of liquid to the downcomer, the

first one through the separator-dryer and the second one through the jet pump of the intact

loop. A 1.23 m. level peak occurred at 11.05 sec., then the level went down and remained just

above the downcomer bottom.

The calculated mass flow rates in the broken loop are shown in the Figure IV-5. The vesselside break mass flow rate exceed the pump-side break mass flow from the onset of the

transient. The difference between these mass flows are due to the different frictions in both

paths coming out of the vessel. On the other hand, the flow rate out the break pump side was

limited by the small area of the jet pump nozzles and hence was significatly smaller than the flow

through the vessel-side break. At about 10.5 s of the transient, the mass flow rate out is

increased as a result of the flashing of liquid in lower plenum. Due to the uncovery of the

recirculation suction, once liquid mass was exhausted (at 13.5 sec) the mass flow rate drops

again.

Critical flow was expected at both break sides and the jetpumps nozzles. As we know, if the

pressure gradient applied to a fluid Is continually Increased, the corresponding increase in the

mass flow will suddenly stop when the sonic limit of the fluid is reached. To model this

phenomena, TRAC code has a homogeneus equilibrium model (HEM) and the Algamgir-Jone

Lenhard correlation, and assumes that.urihomogeneus or nonequilibrium process are not

significant, with the exception of the Algamgir-Jones-Uenhard correlation which considers

turbulence and nucleation in the break. These correlations are not adecuated when the

stagnation state is in the region of subcooled liquid or very low quality.

The mass flow rate at the inlet and outlet of the low-powered channel (CHAN36), averaged

channel (CHAN38) and high-powered channel (CHAN40) are shown in Figures IV-6 through

IV-8, respectively. The channel inlet corresponds to the side-entry orifice, while the outlet

corresponds to the upper tie plate. On the begining the mass flow rate Inlet In all the channels

decrease due to the trip of recirculation pump at 0 s and, by that reason the evolution of the

mass flow rate follows the coastdown of the recirculation pumps. There was a small increase in

the mass flow at 3.0 s, that was due to the close of the main isolation steam. There was a

second surge of flow into the channels caused by the flashing of the liquid into the lower

plenum which started at 8.0 s. The liquid at that time reached saturation and begun to flash into

steam.

The Figure IV-9 shows the calculated pumps speed. In the event of a LOCA, the broken

loop recirculation pump will be reversely driven by two phase fluid flowing back at high speed to

the break pump-side. The coolant In the pressure vessel flashes due to the despressurization

and passes through the jet pump nozzle Into the discarge line forcing the pump to work as a

turbine during the large part of the transient.

10
Generally the pump would rarely work under two-phase flow conditions. As a result, there are

very few data on pump behaviour under these conditions. Built-in code data from Semiscales

experiments has been used in the analysis. The TRAC calculations for the other recirculation

pump, the unbroken loop, show a high reversal velocity in the first instants of the transient

probably due to the considered low pump inertia (Fig. IV-9).

The calculated high pressure coolant Injection (HPCI) and low pressure core sprays (CSs)

flow rates are shown in Figure IV-1 0. The HPCI system was activated by a low-low downcomer

level signal which ocurred at 1.3 s, from this time a delay of 27 s to reach full flow was taken into

account. HPCI was on until 52.08 s at that point the pressure into the vessel had reached the

set point of low pressure HPCI trip, so that the HPCI was turned off. HPCI flow was 186.3 kg/s

and temperature was 338.55 K.

The low pressure core spray (LPCS) system was actived 31 sec. after the initiation signal

took place (at 8,34 m downcomer level) due to the delay of emergency diesel. The evolution of

the mass flow rates of the LPCSs followed the derease of the pressure into the system. LPCS

flow was 200 kg/s per loop and the spray temperature was 340 K. It was found that the injection

water created a two-phase pool across the top of the entire core, while draining into the fuel

bundles.

The void fraction in the two levels of lower plenum is shown in Figure IV-1 1. The void fraction

Increases rapidly at about 8 s, when the lower plenum began flashing. The void fraction

continued to increase until maximum value that was calculated In the level 1 at the second 35

and in the level 2 at the second 50. The lower plenum reflooding began at 54.6 s of the

transient, in the level 1. It was caused primarily by the HPCI water until Instant 52.0 s of the

transient. All this water went into the lower plenum through the downcommer. The contribution

from low pressure core spray began at 33.0 s adding subcooled liquid to the upper plenum

region above the core. One part of this liquid drains into the lower plenum through the core

bypass region and from the fuel bundles through the side entry orifices.

Correct evaluation of dryout or critical heat flux (CHF) Is of importance when predicting

nuclear reactor core behaviour in a loss of coolant accident. The location of dryout within a

bundle determines to a large extent whether rod Integrity will be maintained. The heat transfer

coefficients and the power generated at hottest channel determine the peak cladding

temperature. Figures IV-12 and IV-13 show the heat transfer coefficients for the hottest

channel (CHAN40 or CHAN43) for liquid and vapor. During the first seconds of the transient the

calculated heat transfer coefficient to liquid was relatively large because the rods were generally

calculated to be in nucleate boiling. Large nucleate boiling HTCs are predicted for the whole.

core until the core inlet flow reverse and fluid quality in the bundles Increase. The main heat

transfer mechanism was nucleate boiling and transition boiling along the rods during the first

portion of the transient.

11
Heat transfer liquid coefficient dropped to aproximately 282 kw/(m)2K between the instants

31.0 and about at 70.0 s due to the fact the upper half of the active fuel regions was voided and

therefore, the critical heat flux was reached. The main heat transfer mechanism during that

period was convection to vapor, so that the vapor heat transfer coefficient increased.

The time evolution of the rod temperature before and after spray initiation are represented in

Figure IV-14. That Figure shows three curves of temperatures of the hot rod in the hot channel.

The pointz considered in the rod were the nodes at the top, the middle and the bottom. In

general, the curves may be divided into four stages. In the first stage, the temperature goes

down to saturation during first instants due to the reversal flow through the channel and the

flashing of the lower plenum, mainly affecting the lower nodes of the rods. This phase occurred

during the 25 first instants of the transient.

During the second stage, before the begining of spraying, the wall temperature rises

regulary with the time. At that time of the transient rods are practically dry and all the decay heat

is used in raising the temperature of the rods and the fluid (steam .and droplets). The void

fraction in the hot channel is shown in the Figure IV-15. During this stage the channel has an

averaged void fraction higher than 0.9, except the first nodes that only have a void fraction of

0.5. Therefore, the whole channel was dry during that period. This phase occured from instant

25.0 to 33.0 of the transient.

The third stage starts with spraying initiation. When water is sprayed onto the central

channel, massive heat removal does not occur inmediately over the- entire surface of central

levels. Upper levels are being wetted by the water film and droplets, forcing the temperature to

lower, while the central and lower levels are still dry and not yet effectively cooled. The lowest

level remains near saturation along the transient. In the central levels, cooling Is induced by

steam and falling droplets of water. Consequently, an extremely small heat transfer (Fig. IV-12

and IV-13) exists along this phase. The code calculated the maximun peak of temperature

(PCT) 694.5 K at the instant 77.1 s of the transient. From instants 68 to 71 of the transient a first

cooling of the central nodes occured. Therefore the cooling Is becoming more efective. The

Figure IV-15 shows a small decrease of the void traction at upper levels first, and then at central

levels. That was a consecuence of the phenomena of counter current flow which Is model by

the code. This phehomena Is locate between the upper tie plate and the upper plenum. Figure

IV-16 shows several oscillations of the void fraction that represent the drainage of the channel.

This phenomena becarne more important in the next stage.

Last stage is characterized by a rapid heat removal and rod temperature drop to saturation,

becduse the rods surface are covered by a water film. The evolution of the quench front Is a

primary function ol the counter current flow In each channel. TRAC-BF1I/G1J1 models this

phenomena by the correlation of Kutateladze which is based on experimental data from

General Electric . The CCFL breakdown occurs at about 63 s, hence upper plenum void

fracti•n had a strong oscillation and the top channel void fraction decreased (Fig. IV-15 and IV16). After that, the fluid temperature increase again due to CCFL model. The effect of

12
condensation in the upper plenum is not included in this model, but it is known that the

interfacial heat transfer in the condensation regime affects strongly to the CCFL break down,

after subcooled spray initiation, (Ref. 6)

As a conclusion maximum PCTs of 694.5 K, 598.3 K and 543.2 K were reached at 77 s, 52 s

and 43 seconds respectively in the hot powered channel, averaged powered channel and

peripherial channel (Fig. IV-1 7). The CS flows down through the upper tie plate was sufficient

to turn the cladding temperature around, even though the liquid downflow was limited by the

CCFL phenomena. (Fig. IV-20). The fuel rods were effectively quenched at 122 s when the

cladding temperature dropped to within 5 K around saturation temperature. Fig. IV-1 8 shown

the channels collapsed liquid level.

Fig. IV-19 shows the Rod Temperature evolution during first seconds.

13
V. RUN STATISTICS

TRAC includes a logic that may limit internally the time step, unless the user specifies

maximum time step size. This logic is based on parameters such as the material Courant limit in

the vessel, pressure rates, temperatures and void fractions through the system, and axial

temperature gradients in fuel rods.

Figure V.1 shows the time steps as a function of real time. Since about second 45 the

evolution of time steps show sharp oscillations. Several changes of heat transfer and flow

regimes took place forcing the code to select smaller time steps.

The run statistics are shown in Table V-1. The real time/CPU time ratio was 1/3352.

14
VI. CONCLUSIONS

A considerable margin of 1600C In peak clad temperature respect to the so-called nominal

SAFER calculations has been obtained using TRAC-BFI/GIJ1 code. However, it is

recognized that some of the calculation assumptions could still be considered conservative.

A Critical Heat Flux is not reached within the first phase of the transient and as a result there is

not first peak in PCT.

Identified uncertainty areas will require additional sensitivy analysis in order to fully qualified

the Garoma LOCA model. Two-phase flow through recirculation pumps and leakage flows in

lower plenum, guide tubes, channel and bypass regions are two of these areas.

It Is believed that the current model covers the main LOCA phenomena addressed by

licensing methodologies in a best estimate way.

The Counter Current Flow has been Identified to be the key phenomena in the analyzed

LOCA scenario. The liquid downflow through the upper tie plate Into the channel was the main

contributor to core cooling.

Therefore uncertainties related to this phenomena should be addressed (i.e. by sensitivy

analysis) in order to properly evaluate this particulr" LOCA scenario.

15
VII. REFERENCES

1.- "Acceptance Criteria for Emeroencv Core Cooling Systems for Liaht Water Nuclear Power

Reactor, and Appendix K, "ECCS Evaluation Models", 10 CFR Part 50.46

2.- NUCLENOR, S.A. ",uademos de c•lculo. NUCLE-01". Rev. 0, Febrero de 1989

3.- ENUSA, S.A., "Informe de diseflo nuclear oara Santa Maria de Garot'i. cicto 15". ITEC-172,

Rev. 1. Enero 1989.

4.- Taylor, D.D et al. "RAC-BFI/MODI: An Advanced Best Estimate Comiuter Prooram for

Boiling Water Transient Analysis". NUREGICR3633, EG&G Idaho Inc. 1984.

5.- General Electric Co. "NUCLENOR SAFER/GESTR Analysis". San Josd, 19 Septiembre, 1989.

6.- Nagasaka, Katoh, et al. "nermal Hydraulic Behavoir in Upper Plenum durina Refill-Reflood

Phase of BWR LOCA 1f)411).. Journal Nuclear Science and Technology, 23, 2, February, 1986.

16
TABLE I I - I

GARONA NPP DESIGN CARACTHERISTICS

REACTOR

" Thermal Power (100%)

" Vessel Pressure

" Core Flow

" Steam Flow

* Feedwater Temperature

CORE DIMENSIONS

" Diameter

" Active Lenght

FUEL ELEMENTS

Number of Fuel Elements

* Rod Fuel Layout

* Cladding

* Fuel

" Outter clad diameter

" Clad Thickness

• Channel

1380 Mw

70.3 Kg/cm 2

21.77 x 106 Kg/hr

2.48 x 106 Kg/hr

1832C

0.3683 m.

0.3658 m.

400

8 x 8R and PS x 8R

Zlrcaloy-2

UO 2

1.25 cm/ 1,23 cm

0.086 crrV 0,081 cm

Zrcaloy-4

CONTROL RODS

* Number of Control Rods

* Shape

97

Cruciform

REACTOR VESSEL

Inner Diameter

Inner height

Design pressure

0.4775 m.

18.447 m.

87.90 Kg/cm 2

17
TABLE I I- 1 (Cont.)

RECIRCULATION SYSTEM

Location

Number of loops

Loop diameter

Nominal flow per pump

Number of jet pumps

Jet pump Location

PRIMARY CONTAINMENT

Type

Drywell Design Pressure

Suppresion Pool Design Pressure

Drywell

2

61 cm.

2,019 I/seg

20

inside vessel

Pressure Suppresion

62 psig.

62 psig.

18
TABLA I I I- 1

DESCRIPTION OF THE TRAC.COMPONENTS

NUMBER COMPONENT DESCRIPTION

1 VESSEL Vessel

2 PIPE Central guide tube

3 PIPE Average guide tube

4 PIPE Peripherical guide tube

10 PIPE Suction pipe of the recirculation loop I

11 VALVE Isolation valve of the recirculation loop I

20 PIPE Suction pipe of the recirculation loop II

21 VALVE Isolation valve of the recirculation loop II

22 PUMP Recirculation pump of the loop I

23 VALVE Isolation valve of the recirculation loop II

24 PIPE Discharge pipe of the recirculation loop II

25 JETPUMP Jet pump of the recirculation loop If

26 JETPUMP Jet pump of the recirculation loop I

31 VALVE Isolated valve of the recirculation loop II

32 PUMP Recirculation pump of the loop II

36 CHAN Penpheral bundle

38 CHAN Average bundle

40 CHAN Hot bundle 1

41 PIPE Discharge pipe of the recirculation loop I

43 CHAN Hot bundle 2

50 PIPE Main steam ine from the vessel

52 PIPE Main steam line to the turbine

54 VALVE Main steam isolation valve

58 BREAK Turbine

60 PIPE Feedwater pipe

61 FILL Feedwater

62 FILL High pressure core spray (HPCS)

70 PIPE Pipe of the core spray 1

71 FILL Low pressure core spray 1

72 PIPE Pipe of the core spray 2

73 FILL Low pressure core spray

19
TABLA I V- 1

INITIAL CONDITIONS -.-,

PARAMIETER

Reactor Vessel , .

Total Core Power (104%), (Mw)

Downcomer Water Level, (m)

Steam Dome, (MPa)

Total Core Mass Flow, (kg/s)

Core Bypass (10%), (kg/s)

Recirculatlon Loop

Speed Pump, (rad's)

Feedwater System

Feedwater Mass Flow, (kgts)

Main Steam Line

Steam Mass Flow, (kg/s) :. U.-

MFASUREDQ VALUE -

10.58

7.00-7.03

5,900

- I 1435.00

-10.58

-7.01

6.000

598.40

130.6

714.00

714.00

135

20
TABLE IV-2

CHRONOLOGY OF EVENTS

.E=

Break Initiated

Loss of Offsite Power

Trip of Feedwater System

Recirculation Pump Coastdown Started

Scram Signal Generated

Low-Low Water Level

Control Rods Fully Inserted

Main Steam Isolation Valves Closed

Jet Pump Drive Uncovered

Lower Plenum Flashing

Recirculation Loop Suction Uncovered

HPCI Flow Started

Dryout at the Peak Power Zone

CS Flow Started

Lower Plenum Refill Started

HPCI stopped

Reflood Initiated

TEMe (seconds)

0.0

0.0

0.0

0.0

0.5

1.3

2.9

3.0

5.2

10.0

10.5

28.5

24.2

33.0

54.6

52.0

61.2

21
TABLE IV-2 (Cont.)

Maximum Cladding Temperature reached 77.0

Perlpherial Bundle Quenched 47.8

Average Bundles Quenched 61.5

Hot Bundles Quenched 82.3

22
TABLA V-1

RUN STATISTICS

Real Time

CPU Time

Total number of volumes in the model

Total number of time steps

(CPU x 103)/C x DT - 12.1

RT- 120 seconds

CPU - 402,084 seconds

C- 168

DT. 197,543

23
VENT

ISOLATION CONDENSER

TURBINE

FROM

SYSTEM

LPCI

SYSTEM

PUMPS

TO ESW

DISCHARGE

4 - CORE SPRAY

PUMPS

Fig. Il-1

GA3O& ZMIZNGCY CORX COOLIN SYSTZKS

AMD ISOL&TZON CONDZNSZR SYSTZX

24
STEAM UNE

PQ

CORE SPRAYA Q

@4

2 2-3

INTACT Loop

Figure 111-1 TRAC-BF1 AXIAL NODALIZATION
OUTSIDE OF THE VESSEL

DOWN COMER

PERIPHERIAL CHANNEL

AVERAGED CHANNEL

HOT CHANNELS

Figure 111-2 TRAC-BF1 RADIAL NODALIZATION
WATER ROD WATER ROD

0 FUEL ROD 0 HOT FUEL ROD

AVERAGED FUEL ROD

P8 X 8R COLD FUEL ROD

ACTUAL FUEL ELEMENT

P8 X 8R

FUEL ELEMENT MODEL

Figure 111-3 FUEL ELEMENT MODEL
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

8.0

7.0

6.0 n.

D

(n

0

5.0

4.0

-. PV-010801

120.

3.0

2.0

1.0

.0

0. 30. 60. 90.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fig. IV-1
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

1500.

1200.

TOTPOW-O01

0'

ILO

0

0L

lii

900.

600.

300.

0.

.0 2.0 4.0 6.0 8.0

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

10.0

Fig. IV-2
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

100.

_ TOTPOW-O01

0

o0

bJ

80.

60.

40. L

0. 30. 60. 90. 120.

TIME (s)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fig. IV-3
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

12.

10.

____ LLEV-010001

hi

r-J

0

0

8.

6.

4.

2.

0.

0. 30. 60. 90. 120.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fig. IV-4
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

U)

I-.

Ia.

0.

-J

z

0

8000.

7000.

6000.

5000.

4000.

3000.

2000.

1000.

0.

-1000.

-2000.

-3000.-

- MFLOW- 10001

,& UIF1.CYJ-1-110na

30. 60. 90.

TIME (s)

120.

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Pig. IV-5
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

3500.

UFLOW-.36iOO I

1.41 . A .- . i. i 0 . (I)

N,

V)

Ie0

-j

in

z

z

X2500.

1500.

500.

-500. L

0. 30. 60. 90. 120.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

ll'cj. IV-0,
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

3500.

U)

w

EZL

z

4

IA

0.

____ MFLOW-.S80001a

2500.

1500.

500.

-500.

0. 30. 60. 90.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

120

Fig. IV-7
SANTA MARIA DE GARONA NPP

TRAC-BFI (G1J1). MARCH 90

re

U.

0

C3

IL

X) CA U

Xa

600.

500.

400.

300.

200.

I00.

0.

-100.

-- MFLOW-4000O1

0. 30. 60. 90.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

120.

[,iti . IV-fj
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

(IL

0

0~

0:

150.

100.

50.

0.

-50.

-100.

-150.

-200.

-250.

-300,

-350.

-400.

2

K j * p i p p p p i p p p p i i p p p p

___-OMEGAN-22000

CUEGAN-32ofl(o

I I I I I j * I I I I I ~ ~ ~ ~ ~ I I I I I

0. 30. 60. 90.

-rIME (s)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fig. IV-')

12
SANTA MARIA DE GARONA NPP

rRAC-BF1 (G1J1). MARCH 90.

500.

____ FLEAK*-t6200

A m~l "W . gI?..*

U)

(LA

U)

400.

300.

200.

100.

0. it.

0. 30. 60. 90. 120.

rIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fiq. IV-l()
SANTA MARIA DE GARONA NPP

TRAC-BF I (GlJ1). MARCH 90

1.0

ALP-010101

.8

z

0

U.

W0

0o >

z

(Li

-J

-I

.6

.4

.2

0o

0 .30. bO. 90. 120.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fiq. IV-I1
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

60000.

50000.

____ RODHL-400306

C'14

I.-

CY

ci

z

z

C-)

I--

0

m

40000.

30000.

20000.

10000.

0.

0. 30. 60. 90. 120.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

I.*jtj. IV-I'2 ,3
SANTA MARIA DE GARONA NPP

TRAC--BF1 (G1Ju). MARCH 90

1500.

- ROOHL-400306

N

CD -j

-Jtid

z

z

0

X

1000.

500.

30. 40. b.bO. 70. at0. 90.

T IME (S)

k C k I L A ON P04 r'-P V1 ',0U r 1014 L IiPGF Dii'~AK i OGA

I"1,. IV-I.1 I,
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

200.

- RODHV-400306

N.

U

I.-

0

-LJ

z

z

C-)

I-0

150.

100.

50.

0.

0. 30. 60. 90. 120.

TIME (S)

RECIRCULAriOl-i PUMP SUCTiON LARGE BREAK LOCA

Fy. IV-13
SANTA MARIA DE GARONA NPP

TRAC-BF1 (GIJi). MARCH 90

w

r~0

z

X

t--

0

X

1000.

950.

900.

850.

800.

750.

700.

650.

600.

550.

500.

450.

400.

350.

.____ rSURF-40030b

x ISURF-400302

O. 30. 60. 90.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fig. IV-14

120.
SANTA MARI.A DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

1.0

.8

____ ALI'HA-40000&

L. At PHA-4fl(JA~d

x ALPHA-400008

c) ALPHA-4OO0)10

z

0

IC,

Ii.-

0

.--

ILl

z

z

"-

.6

.4

.2

.0

0. 30. 60. 90.

rIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

120.

Fiq. IV-15
SANTA MARIA DE GARONA NPP

TRAC-BFI (GIJI). MARCH 90

1.0

_ ALPHA-O 0501

z

z

z

0

C.)

0

D

z

-j

(L.

w(-

m.

.6

.4

.2

0. 30. 60. 90. 120.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fi"J. IV-1(
SANTA MARIA DE GARONA NPP

TRAC-BI1 (G(J1). MARCH 90

tn

-1

wJ

z

z

wU

cc)

(Ll

1000.

950.

900.

850.

800.

750.

700.

650.

600.

550.

500.

450.

400.

350.

- PEAKR~ODT-360

A, PEAK mb t -. 1fllO I

X PEAICRODT-40001

0. 30. 60. 90.

T IME (S)

RECIRCULAT ION PUMP SUCTION LARGE BREAK LOCA

Fig. iV-17

120.
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1J1). MARCH 90

3.0

- CLEVL-iti000l

,& CLEVI -:38noni

x CLEVL-400001

(n~

-j

w

Ld

0

2.0

1.0

.01-

0. 30. 60. 90. 120.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

Fig. IV-18
SANTA MARIA DE GARONA NPP

TRAC-BF1 (GlJ1). MARCH 90

z

z

0

a. I

0~

re600.

590.

580.

570.

560.

550.

540.

530.

520.

510.

500.

TSURF-400008

STSURF-400009

X TSURF-400010

(! TSURF-400011

.0 2.0 4.0 6.0 8.0 10.0

TIME (S)

RECIRCULA'I.ION PUMP SUCTrION LARGE BREAK LOCA
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1Ji). MARCH 90

200

____ RHOM-400004

A RHOU-400006

X RHOU-400008

--) (D RHOMJ-400010

z

w&

co Ixl

z

z

X

0

3:

150.

100.

50.

0.

0. 30. 60. 90. 120.

TIME ýS)

Fig. IV-20

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA
SANTA MARIA DE GARONA NPP

TRAC-BF1 (G1Ji). MARCH 90

.04

.03

.02

-DELT-O01

L--

'-0

Id

q0

I.-

.01

0. 30. 60. 90. 120.

TIME (S)

RECIRCULATION PUMP SUCTION LARGE BREAK LOCA

I ti",. V-I

NRC FORM 335 12-89)

NRCM 1102,

3201.3202

U.S. NUCLEAR REGULA1

BIBLIOGRAPHIC DATA SHEET

{See instructions on the reverse)

2. TITLE AND SUBTITLE

Recirculation Suction Large Break LOCA Analysis of the

Santa Maria De Garona Nuclear Power Plant Using TRAC-BF1 (GIJI)

1. REPORT NUMBER

(1EAd by NRC. Add Vol, Supp. Rev.,

and Addacdum Numbers, I any.)

NUJREG/IA-0067

ICSP-GA-LOCA-T

3. DATE REPORT PUBLISHED

MONTH YEAR

.August 1992

4. FIN OR GRANT NUMBER

A4682

5. AUTHORIS) 6. TYPE OF REPORT

J. V. Lopez, J. Blanco, Polytechnical University of Madrid Technical

J. L Crespo, University of Cantabria 7. PERIOD COVERED linclusi.e

R. A. Fernmndez, Nuclenor, S. A.

B. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, pmvld Divdn, Ok* of f iepron, U.S N• ~uckrieastr Commission, and mallingo dd,'ss If contractor. provide name and mailing address.)

Polytechnical University of Madrid

c/Josi Abascal, 2

28006-Madrid

Spain

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NReC, typesin, eese- aho"'ifactr. roioe NRC Office erReieU. ANhwse Regulatory Commission, and mailing addre)

Office of Nuclear Regulatory ResearchU.S. Nuclear Regulatory Commission

Washington, DC 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT t200 words or lekss

A best estimate analysis of a recirculation suction pipe large break loss of coolant

accident analysis for Santa Maria De Garona nuclear power plant using TRAC-BF1 code

is presented.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

Unlimited

Large Break LOCA, TRAC-BF1 14. SECURITY CLSIFI'CTION

(This Pgel

Unclassified

(This Report)

UJnclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 12-89)
THIS DOCUMENT WAS PRINTED USING RECYCLED PAPER

NUREG/IA-W07 RECIRCULATION SUCTION LARGE BREAK LOCA ANALYSIS OF THE SANTA MARIA

DE GARONA NUCLEAR POWER PL.NT USING TRAC-BF1 (G1J1)

AUGUST 1992

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300